ML20030D185
| ML20030D185 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 08/26/1981 |
| From: | Withers B PORTLAND GENERAL ELECTRIC CO. |
| To: | Clark R Office of Nuclear Reactor Regulation |
| References | |
| IEB-79-01B, IEB-79-1B, TAC-42502, NUDOCS 8108310492 | |
| Download: ML20030D185 (78) | |
Text
{{#Wiki_filter:. r - Elll PortlandGeneralElectricCompany !u1 D ei:tes 'Ae Present August 26, 1981 Trojan Nuclear Plant Docket 50-344 License NPF-1 Director of Nuclear Reactor Regulation ATTN: Mr. Robert A. Clark, Chief Operating Reactors Branch No. 3 Division of Licensing U. S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Clark:
Your letter dated May 27, 1981 transmitted the Safety Evaluation Report (SER) for the environmental qualification of safety-related electrical equipment at the Trojan Nuclear Plant. We were requested to provide, within 90 days, the information identified in Sections 3 and 4 of the SER. Accordingly, a detailed response to the SER is provided in the attachment to this letter. This response addresses the outstanding information identified in Section 3.0 of the SER and provides resolutions of the deficiencies identified in Section 4.0. Alternatives to several NRC staff positions are also presented in accordance with the guidance provided in your letter. Sincerely,
- M Bart D. Withers Vice President d\\
Nuclear 3r F ,1 E W.J. ,,b 2 AUG 311981 > 7 ! Attachment 8 d. ~ t]\\ U EN" 7 c: Mr. Lynn Frank, Director , d(S State of Oregon Department of Energy N '/ 4 :. t j Mr. R. H. Engelken, Director U.S. Nuclear Regulatory Commission Region V C108310492 810826 , f, I g PDR ADOCK 05000344 < S W Sfman Owt. Ficand, Oregon 97204
s 4 0 i I TROJAN NUCLEAR PLANT PORTLAND GENERAL ELECTRIC COMPANY'S R"SPONSE TO NRC SAFETY EVALUATION REPORT OF MAY 27, 1981 ON ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT August 26, 1981 This is a detailed response by Portland General Electric Company (PGE) to the NRC's Safety Evaluation Report (SER), dated May 27, 1981, regarding i the environmental qualification of safety-related electrical equipment at the Trojan Nuclear Plant. The following provides the information identi-fled in Sections 3 and 4 of the SER as requested. As part of the ongoing review of the environmental qualification of safety-related nectrical equipment at Trojan, PGE plans to submit an updated master list and equipment qualification work sheets, consistent with previous submittals to IE Bulletin 79-01B, by January 1, 1982. This update will incorporate changes and new information since PGE's last submittal on January 13, 1981. Included will be new qualification summary information covering additions of TMI kssons Learned (NUREG-0737) equipment, recent revisions to specified environmental qualification parameters, updated information from ongoing test report reviews, and results of continuing system analyses and evaluations. Section 3.1 (Completeness of Safety-Related Equipment) The NRC Staff concluded that the information on safety-related systems included in PGE's previous submittals was insufficient to verify that those systems are all the systems required to achieve or support emergency reactor shutdown, Containment isolation, reactor core cooling, Containment heat removal, core residual heat removal, and prevention of significant release of radioactive material to the environment. The SER required a listing of all safety-related systems, both inside and outside mtentinlly harsh environments. Artached Table 1 provides this list and confirms that all pertinent safety-related systems have been addressed in the harsh environment review.. ~ - - - - ~, ~, - -r. ,,y
n b Tha NRC Steff cles rsquictcd thnt a complets.lict of all dispicy instru-mentation mentioned in the LOCA and HELB emergency procedures be provided even though display instrumentation was identified in previous PGE submittals by the functional description " post-accident monitor" or "PAM" in the equipment summary sheets. Table 2 provides a listing of equipment in this category. Except for equipment installed in response to NUREG-0737, equipment qualification information in the form of summary sheets has previously been provided for all components cf the display instrumentation exposed to harsh environments. Instrumentation which is not considered to be safety related but which is mentioned in the emergency procedures also appears in Table 2. Justification.is provided for not considering these instruments safety related. Equipment failure has been reviewed to assure that the operator will not be mislead or subsequent failure would not adversely affect the mitigation of the consequences of the accident. Consideration of post-accident sampling and monitoring and radiation monitoring equipment, which is closely related to the review of TMI Lessons Learned modifications (NUREG-0737), has been deferred from this response in accordance with Sections 3.1 and 5.0 of tha SER. Section 3.2 (Service Conditions). The S:aff requested that PGE verify that the Containment Spray System is not subjected to a disabling single component failure and therefore satisfies the requirements of Section 4.2.1 of the DOR Guidelines. The Trojan Containment Spray System (CSS) is described in detail in Section 6.4 of the FSAR. The analysis providcl in Table 6.4-2 of the FSAR provides sufficient veri 11 cation that the CSS is not subject to disabling single component failures. Section 3.3 (Temperature, Pressure, and Humidity Conditions Outside Containment) l In the SER, the Staff concluded that the minimum temperature profile used in the specifications for equipment qualification purposes should include l l [ c cargin to ccccunt for tha highar-thin-cvsrsgs eszperaturco in tha uppar regions of the Containment that can exist due to stratification, especially following a postulated MSLB. Use of the steam saturation temperature corresponding to the Local building pressure (partial pressure of steam plus partial pressure of air) versus time is mentioned in the SER as providing an acceptable margin for either a postulated LOCA or MSLB, whichever is controlling, as to potential adverse environmental effects on equipment. The SER also stated that "the Licensee's specified temperature (surface condition) of 20(*" does not satisfy the above requirement. A saturation temperature corresponding to the pressure profile (307'F peak temperature at 60 psig) should be used instead. The Licensee should update his equipment summary tables to reflect this change. If there is any equipment that does not meet the Staff position, Licensee must provide either justification that the equipment will perform its intended function under the specified conditions or propose ccrrective action." PGE takes exception to the NRC position delineated above. An alternative approach, consistent with the DOR Guidelines and responsive to the alternative NRC position expressed in the transmittal letter to the SER, is discussed in attached Note 1 to Table 3. Section 3.5 (Submergence) Submergence was considered as an open item in the SER because the maximum flood level inside Containment could not be established by NRC review of previous PGE submittals. PGE was also requested to provide assurance that subsequent failure of submerged equipment exempt from submergence qualification would not adversely affect way other safety function or mislead an operator. Additionally, the NRC requested a discussion of operating time, across the spectrum of events, in relation to the time of submergence. The NRC also requested that submergence of safety-related electrical equipment outside Containment be addressed. The above concerns are addressed in attached N te 3 to Table 3. o Siction 3.6 (Chanical Speny) t The effects of chemical spray were considered unresolved because the FSAR values for the chemical concentration of Containment spray were not mentioned in previous PGE submittals. The resolution of this outstanding item is discussed in attached Note 4 to Table 3. Section 3.7 (Aging) In this section of the SER, PGE was requeste1 to submit supplemental information to verify and identify the degree of conformance to the requirements of Section 7 of the DOR Guidelines. Aging effects have been explicitly considered for all safety-related equipment at the Trojan Nuclear Plant potentially exposed to harsh environments. Attached Note 5 to Table 3 describes PGE's approach for evaluating aging effects on existing equipment and discusses the degree I of conformance to the requirements of Section 7 of 'the DOR Guidelines as requested in the SER. In addition, Table 3 addresses, in summary fashion, the aging evaluation conducted for all equipment identified as required to maintain functional operability in harsh environments. l Section 3.8 (Radiation - Inside and Outside Containment) l The SER stated that the integrated doses required by the Licensee inside Containment (1.1 x 107 rads) do not envelope the DOR Guideline require-ments and are therefore not acceptable. The radiation service condition provided by PGE was judged lower than provided in the DOR Guidelines for i gamma and beta radiation. PGE was requested to either provide justifica-tion for using the lower service condition or use the DOR Guidelines l l for both gamma and beta radiation. For the former option, analysis must ( be provided including the basis, assumptions, and a sample calculation. l l l
O PGE has more recently calculated post-accident gamma and beta total integrated dosea (TIDs) for Trojan utilizing the methodology outlined in Appendix D of NUREG-0588 and NUREG-0578. The TIDs, which were determined for several time periods, locations, and conditions, are used for TMI-related modifications as well as IE Bulletin 79-01B electrical equipment qualification evaluations. The results of this evaluation, which are presented in Table 4, are based on two bounding conditions. The first bounding condition assumed a LOCA wiun 100 percent core noble gases, 50 percent core halogens, and 1 percent otaer core fission products released into the Containment. The second condition assumed the Reactor Coolant System (RCS) remained intact with the same source term as above. The integrated doses presented in Table 4 are an expansion of the Trojan FSAR Section 3.11.3.1 AID values. The Trojan FSAR assumed the fission product release was distributed homogeneously throughout the Containment atmosphere. This evaluation, however, assumed a spatial distribution of fission products in the Containment atmosphere and sump. Due to this distribution, 30-day total integrated gamma doses near the sumb are 7 slightly higher than the Trojan FSAR value of 2.1 x 10 rads, and the dose in the Containment dome is slightly higher. Integrated doses for equipment operating times of 1 hour, 1 day, I week, and 1 year, as well as 30 days as presented in the Trojan FSAR, are provided in Table 4. This allows evaluation of equipment and components whose usefulness is required for times other than 30 days. Beta total integrated doses for material in contact with reactor coolant snd materials within the Containment atmosphere are also provided in Table 4. The maximum total integrated beta dose in air is 1.4 x 108 rads for a 1-year period. Figures 1 and 2 provide beta volume correction factors which were used to determine beta integrated doses for finite volumes in Containment atmosphere. Stetien 4.1 (Equipaznt Rtquiring Immtdiate Corrsctiva Action) No additional information for equipment identified in Appendix A of the SER was requested by the NRC. However, for purposes of completeness and continuity, Table 3 addresses all equipment and NRC-identified deficiencies in this category. Justification for continued operatio'. pending implementation of corrective action by June 30, 1982, althoagh provided in previous submittals, is also included in Table 3. Section 4.2 (Equipment Requiring Additional Information and/or Corrective Action) This section of the SER required PGE to supplement information previously presented by providing resolution of the deficiencies identified in Appendix B. Attached Table 3 provides resolutions of each NRC-identified deficiency in this category. Lacluded are descriptions of any corrective actions considered necessary, schedules for their completion, and justifi-cation for continued operation pending implementation of corrective actions. Section 4.3 (Equipment Considered Acceptable or Conditionally Acceptable) The Staff identified the equipment in Appendix C of the SER as condition-ally acceptable subject to the satisfactory resolution of the Staff concerns identified in Section 3.7 of the SER. The Staff specifically determined that the Licensee did not clearly (1) state that an equipment material evaluation was conducted to ensure that no known materials susceptible to degradation because of aging have been used, (2) establish an ongoing program to review the Plant surveillance and maintenance records in order to identify equipment degradation which may be age related, and/or (3) propose a maintenance program and replacement schedule for equipment susceptible to age-related degrrdation or equipment that is qualified for less than the life of the Pltat. l The supplemental information requested in this section of the SER is provided in attached Table 3 (Sheet 37) and its accompanying Note 5. --
S;ction 5.0 (Dafarrad R;quirrmute) 4 In a letter dated January 31, 1981 to NRC Region V, PGE committed to provide qualification information for NUREG-0737 equipment not yei. submitted by certain specified dates. In another letter, dated June 22, 1981, to NRC Region V, PGE proposed to defer the submittal date for qualification information for Auxiliary Feedwater System flow rate indication from July 1, 1981 to the 90-day SER response. Attached Table 5 and its accompanying Figure 3 provide this information for the subject equipment. See also Table 3, Sheet 21, for additional information. t I I DRS/sh 4-66.63B21 ! l
TABLE 1 Shaat 1 of 2 TROJAN NUCLEAR PLANT LIST OF SAFETY-RELATED SYSTEMS NRC System / Function Identification [a] Disposition [b] Engineered Safeguards Actuation Included (System 53 et al). Reactor Protection Instrumentation included in Reactor Nennuclear Instrumentation System (System 80). Other components in system not exposed to harsh environment. Containment Isolation Instrumentation included in Engi-neered Safety Feature Actuation System, Main Steam System, and Reactor Nonnuclear Instrumentation System, et al. All other electri-cal components not exposed to harsh environment. Main Steam Line Isolation Included in Main Steam System (System 83). ~i Main Feedwater Isolation Instrumentation included in Main Steam and Feedwater Systems. All other electrical components in system not exposed to harsh environment. Emergency Power Electrical components in the system l not exposed to harsh environment. Emergency Core Cooling l High-Pressure Injection Included (Systems 50 and 52). Low-Pressure Injection Included (System 49). Accumulators Included (System 52). Containment Sump Recirculation Included in Residual Heat Removal System (System 49). Containment Heat Removal Containment Air Coolers Included (System 60). Containment Spray System Included (System 61). Containment Fission Product Removal Containment Spray System Included (System 61). Containment Air Purification System not required. and Cleanup
TABLE 1 Shsst 2 of 2 NRC System / Function Identification [a] DispositionID} Containment Combustible Gas Included in Primary Containment Control (Including Hydrogen System (System 59). Vent) Auxiliary Feedwater Included in Feedwater System (System 45). Containment Ventilation and Included in Containment Heating and Cooling Ventilation System (System 60). Control Room Habitsbility Electrical components in system not exposed to harsh environment. Ventilation for Areas Containing Included in Auxiliary Building Safety Equipment Heating and Ventilation System (System 32). Component Cooling Water Included (System 16). 1 Service Water Electrical components in system not exposed to harsh environment. Emergency Shutdown Residual Heat Removal Included (System 49). Power-Operated Relief Valves Included in Reactor Coolant System (System 64). Pressurizer Spray System not required. Chemical and Volume Control included (high-pressure injection and Containment isolation por-tions only, System 50). Steam Dump Electrical components in system not exposed to harsh environment. l Reactor Coolant Included (System 64). Main Steam Included (System 83). [a] Based on Table II of Supplement No. 2 to IE Bulletin 79-01B. Post-accident sampling and monitoring and radiation monitoring systems are not listed since they are closely related to the review of TMI Lessons Learned modifications (NUREG-0737) and will be addressed in conjunction with that review. [b] Systems with equipment exposed to harsh environments have been " included" within the scope of the harsh environment review (ie, IE Bulletin 79-01B) and are so indicated. Systems with equipment in nonharsh environments will be addressed separately later as part of the mild environment review. DRS/sh 4-66.62B24 i u
= TABLE 2 Sheet 1 of 4 TROJAN NUCLEAR PLANT LIST OF DISPLAY INSTRUMENTATION [a] E Component Function Disposition
- LT459, Pressurizer level Included
- LT460, LT461
- PT455, Pressurizer pressure Included
- PT456, I
PT457
- PT2087A, Containment pressure Included [c]
PT2087B
- LT10050, Containment humidity Diagnostic instrument only. Failure will HT10051 not impair accident response.
FT513 & 514 Steam flow Included thru l FT543 & 544 l FT510 & 511 Feedwater flow Included thru l FT540 & 541 TE413A&B RCS temperature Included TE423A&B & RCS Subcooling TE433A&B TE443A&B LT517-D519 Steam generator Included thru level LT547-D549 Status Panel Verify reactor trip See Note d Status Panel Verify turbine trip Economic protection only l Status Panel Verify Containment See Note d l Isolation Status Panel Verify ESF alignment See Note d FT970A&B Verify ECCS flow Included l FT971A&B FT918 FT922 FT917 1 I
TABLE 2 Sheet 2 of 4 Component Function Disposition (bl FT3043E Verify AFW Supplying Included FT3043F steam generators FT3043G FT3043H Status Panel Verify Containuent NHE spray pumps operating ZS2216 Verify steam line ZS2277 thru 2280, 2216 & 2236 NHE; others ZS2236 isolation included. ZS2256 ZS2276 ZS2277 thru 2280 LT 5201 Monitor CST level NHE ZS 507A-DC Verify steam dump NHE ZS 500A-J valves closed ZS 2210 Verify atompheric NHE ZS 2230 relief valve ZS 2250 position ZS 2270 l PT514-9516 Steam line pressure Included PT544-9546 LT 1921 RWST I4 vel NHE LT 1899 ZS8716A Verify M08716A or ZS part of motor operator; thus qualified ZS8716B M08716B closed as part of M08716A&B. PT919 Verify RHR supply to Included PT923 SIS PRM-1 & Containment ARM-6 included [c]; provides primary indi-ARM-6 radiation cation. Failure of PRM-1, therefore, will not impair accident response. Included [c] LT 4208Al Containment sump LT 4208A2 level LT 4208B1 LT 4208B2 LT 102 Boric acid storage NHE LT 106 tank level
1 TABLE 2 Sheet 3 of 4 Component Funetion Disposition [b} ZS455A PORV and block valve Included ZS456 position ZS8000A&B FT121 Normal charging flow Diagnostic instrument only. Failure will not impair accident response. Sell 89A Pressurizer safety Included SE1189B valve flow monitor Sell 89C LT2069A&B NA0H tank level Included FT600 Hot leg injection NHE flow PRM-6 Condenser air dis-hE charge radiation PRM-10 S/G blowdown NHE radiation ZS455C Pressurizer spray Diagnostic use only. Failure will not ZS455B position impair accident response. ZS8145 TE10101 Containment Tempature Diagnostic instrument only. Failure will thru not impair accident response. TE10139 PT3023 Feedwater Pump Diagnostic instrument only. Information Discharge Pressure can be obtained from qualified feedwater flow. V-A1/2 ESF Bus Voltage NHE V-A2/2 PRM-2 Auxiliary Building hE Ventilation Exhaust Radiation PT403 RCS Wide Range Included PT405 Pressure C285A Hydrogen Sample Recently included in accordance with C233B System NUREG-0737. Qualification to be addressed per Note c. 1
TABLE 2 Sheet 4 of 4 Component Function Disposition [b] ZS8716A&B RHR Train Cross Included Connect Indication TE604&605 RHR Heat Exchanger Included Outlet Temperature [a] Includes all display instrumentation mentioned in the LOCA and HELB emergency procedures, except those instruments identified for establishing normal charging and letdown af ter safety injection. [b] Components of the display instrumentation exposed to harsh environments have been " included" within the scope of the harsh enviornment review (ie, IE Bulletin 79-01B) and are so indicated. Components in non-harsh environments (NHE) will be addressed separately later as part of the mild environment review, and are so indicated. [c] Equipment recently installed in response to NUREG-0737. Qualification will be addressed in the next update of the master list and qualification work sheets, to be submitted by January 1,1982. p DRS/sh 4-66.62B4
- !OTE d TO TABLE 2 Criteria for Including Stem-Mounted Limit Switches on the IE Bulletin 79-01B Master List Stem-mounted limit switches, subject to harsh environments, are included on the Master List if:
a. Their failure could impair the operation of a system needed to mitigate a LOCA or HELB, or b. They provide status panel indication and meet the cri-teria discussed below. Limitorque cam operated limit switches are considered part of the motor operator since qualification of the operator included qualification of the limit switch. Therefore, these were not specifically identified as limit switches on the Master List. Status Panel Indication a. Reactor Trip Status All inputs to the reactor trip status panel are from Solid State Protection System inputs located in the control room. Since this is not a harsh environment, none of the equipment providing input to the reactor trip status panel is included on the Master List. b. Safeguard Yip Status The safes sard trip status panel functions from inputs from safeguard trip comparators in the control room similar to the reactor trip status panel. c. Safety Injection and Containment Isolation Status r Inputs to these status panel are included on the Master List if: l
- The input device is in a harsh environment, and
- The valve must change state following the accident, and The function is not implemented redundantly and i
automatically for the train in question, and The status of the valve cannot be determined by other already qualified means. i KM/jr 4-66.62B8 1 r-<- .n,
TABLE __3 TROJ AN NLCLEAR PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND GENERAL ELECTRIC COMPANY'S RESCLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 1 of 43 Equipment Description Manufacturer Model No. Component No. Location Function Differential Transmitter Barton 384 FT 970A&B, 971A&B A RHRS flow indication FT 918, 922 A SIS flow indication LT 547 C Steam generator level indication and RPS/ESFAS input i FT 542 C Main steam line flow indica-tion and RPS/ESFAS input FT 917 A CCP flow indication i. NRC-Identified Resolution of Deficiency Deficiency QT, T. P, R, A, QM, RPS (for FT 970A&B, 971 A&B, 918, 922) Replace with qualified units by 6/30/82. Testing is currently underway to qualify Rosemont 1153D and Foxboro N-E10 series transmitters. QT, T. P CS, R A. QM, RPS (for Ur 547) Replace with Barton 764 by 6/30/82. T, P, CS, R, A, QM, RPS (for FT 542) Replaced with Barton 764 Juring 1981 refueling outage, which is con-sidered qualified, except for aging (see Sheet 13). QT, T, P, H. R. A, QM, QI, RPS (for FT 917) Replace with qualifted unit by 6/30/82. Testing is currently under way to qualify Rosemont 1153D and Foxboro N-E10 series transmitters. Justification for Continued Operation Pending Corrective Action For RHRS, SIS and CCP flow, other qualified means are available to confirm core injection flow, such as RCS temperature and pressure. 1. 2. For steam generator level (LT 547), recandant leve tadication and RPS/ESFAS inputs are provided by Barton 764 differential transmitters (LT 517, 518, 519, 527, 528, 529, 537, 538, 539, 548, 549) which are considered qualified except for aging (see Sheet 13). 3. Corrective action has been implemented for steaa line flaw (FT 542)). 4. It should be noted that qualification test reports show post-accident operability for suf ficient time to perform trip function (i.e., RPS/ ESFAS input). Moreover, qualification test reporte demonstrate adequate qualification for outside Containment applications; however, vendor quality assurance records are inadeqaute to completely verify applicability of test report to Trojan transmitters. x
TABLE 1 TROJQ[3 NUCLEAR PLANT ENVIRONMENTAL QUALIFICA)l0N OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND CENERAL ELECTRIC COMPANY'S RESOLUTION OF LDENTIFit.D DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 2 of 43 Equipment De sc r i pt ion Manu f ac t urer Model No. Component No. Location Function Pressure Transmitter Barton 145 PT 919, 923 A SIS pump pressure indication PT 947 A CCP/Blf outlet pressure ind ic at ion PT 515, 525, 535, 545 M Main steam line pressure indication and ESFAS input NRC-Ident i fied De f ic iency Resolution of Deficiency CT.T.P H.R,A,QM,RPS Replace with qualified unite by 6/30/82. Testing is currently underway to quali fy Rosemont il53D and Foxbora N-ERO series transmitters. Justification for Continued Operation Pending Corrective Action 1. For SIS pump pressure and CCP/8IT outlet pressure, other qualified means are available to confine injection flow, such as RCS pressure and t empe rat ure. Moreover, direct flow indication via FT 918/922 and FT 917 is preferred parameter to monitor and will be qualified by 6/30/82. 2. Eteam line pressure transmitters are qualified for an MSLB inside Contsimment. For an MSLB outside Containment, manual initiation of ESFAS and/or other qualified autumatic trips are available (e.g., high steam flow coincident with low-low Tavg.). Moreover, redundant pressure t r ansmi t t e r s ( PT 514, 516, 524, 526, 534, 536, 544, 546) are located in a non-harati environment. 3. It should be noted thas qualification test report s show post-accident operability for sufficient time to per form t rip funct ion (i.e., RPS/ ESFAS input). Moreover, quali fic at ion test reports demonst rate adequate qualification for outside Cont ainment applic at ions; however, vendor quality assur ance recor as are inadeqaute to completely verify applicability of test re po r t to Trojen transmitters.
r-- Tant.E 3 TROJAN NUCLEAR FRANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND GENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 3 of 43 Equipment De sc r ipt ion Manufacturer Model No. Component No. Loca t ion Function Fressure Transmitter Earton 389 rr 403 C RCS wide-range pressure indication, RHR isolation valve Interlock, and input to RCS subcooling margin monitore. NRC-Identified Def tetency Resolution of Deficiency T. F. H CS, A. S, QH RPS Replace with Earton 763 by 6/30/82. Justification for Continued Operation Fending Corrective A_ tion 1. Redundant function provided by a Berton 763 pressure transmitter (PT 405) which le conaldered qualified except for aging (see Sheet 13). 2. For RCS pressure indication, redundant narrow-range pressure indication provided by Ratton 763 pressure transmitters (PT 455, 456, 457), which are considered qualif ted except for aging (see Sheet 13). 3. The RHR 1.olation valve interlock performa no safety function in a harsh environment.
(_ TABt1 3 TROJAN NUCLEAR FIANT ENVIROM1 ENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND CENERAL ELECTRIC COMPANY'S ~ RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 4 of 43 Equipment De sc r ipt ion Ma nu f ac t ure r Model No. Component No. Location Function Pressure Transmitter Barton 393 PT 2080, 2081, P Containment pressure indt-2082, 2083 cation and ESFAS input PT 458 C Pressuriser pressure indt-ration and RPS/ESFAS input NRC-Identified Deftetency Resolution of Deficiency QT, T P, H. R, A, QH, RPS (for PT 2000, 2081, 2082, 2083) Replace with qualtf ted unite by 6/30/82. Testing to currently underway to qualify Rosemont 1153D and Fomboro N-E10 series transmitters. QT T. P, H, CS, R. A. QH, RFS (for PT 458) Replace with Barton 763 by 6/30/82. Justification for Continued Operation Pending Corrective Action 1. For Containment pressure, manual initiation of ESFAS available to actuate CSS and steam line toolation. However, automatic initiation of ESFAS function will likely occur before transmitters receive significant radiation dose, and the tranaeltters are acceptably quellfled for other environmental conditions. Redundant wide-range pressure indication previded by Rosemount Il53As (PT 2087A&B) which are qualtfled to IEEE 323-1971 (NUREG-0588, Category 11). See Table 5 for quallitcation summary intossatlnn applicable to this equipment. 2. For pressuriser pressure, redundant pressuriser pressure indication and RPS/ESFAS inpute are provided by Barton 763 pressure transettters (PT 455, 456, 457) which are considered qualif ted except for aging (see Sheet 13). E'
TARI.F. 3 TROJA] NUCLEAR PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTMICAL EQUIPMENT PORTLAND CENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Shees 5 of 43 Equipment De sc r ipt ion Manu f act urer Model No. Cumponent No. Loc at ion Func t ion RfD Burne POR-302 TE 604, 605 A RHRS heat eschenger outlet temperature indication NRC-Identified De fic ienc y Resolution of Deficiency R*P ace uith qualified unite by 6/30/82. l qT, T. P H R. A. QN, RFS Justification for Continued Operation Pending Corrective Action Redundant RCS t emperature indication is available via Rosemont 176KF RTDs uhich are considered qualified (see Sheet 14). CCWS temperature and flou indication is also available via inst rument e located in a non-harsh environment (TE 3273Al&A2 and 3283Al&A2 and FIS 3324A&R). 4
P TARI.E 3 TROJAN NUCLEAR PLANT ENVIRONMENTAL QUALIF! CATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND CENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 6 of 43 Equipment De sc r ipt ion Ma nu f ac t ur er Model No. Cumponent No. Location Function Limit Switch NANCO D2400X ZS 1782, 1783 A RHR sample isolat ion valve posit son idication ZS 8880 P Containment isolation valve position indication and seat in ZS 8888 P Containment isolation valve position indication and seat in ZS 2256, 2276, M Main steen line isolation 2295, 2297 valve position indication PRC-Ident i fied Resolution of Deficiency Deficiency Limit suitches ZS 1782,1783, 8880 and 8888 have been replaced with QT, T, P. H. R, A. QN, RPS (all) NANCO EA 170s. Limit switches ZS 2356, 2276, 2295 and 2297 to be l replaced with NAMCO EA 180s by 6/30/82. Justification for Continued Operation Pending Corrective Action 1. The NAMCO EA 170 series used for RHR semple and Cont ainment isolation valve position are considered qualified (see Sheet II). For an MSLB ou. side Cont ainment, alter-Steam lit a isolation valve position suitcLes are qualified for a LOCA or MSL5 inside Containment. 2. nate qualified means are available to ascertain isolation valve position, such as main steam line flow and pressure indication.
(- TABl.E 1 TIDJAN NUCLEAR PIAIT ENViniMet:NTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT ~ PORTLAND CENERAL ELECTRIC COMPANY'S RESOLUTION OF IDEKTIFIED DEFICIENCIES IN NBC SAFETY EVALUATION REPORT Sheet 7 of 43 Equipment De sc r i pt ion Manu f acturer Model No. Component No. Location Function Splices Raych em WCSF-N C Electrical insulation NRC-Ident i fied De ficiency Resolution of Deficiency CS This equipment is qualified by test for a maximum pH of II.0, which exceeds the specified qualification requirement of pH 10.5. Also see Note 4. A Analysis of test data using the Arrhenius methodology demonstrates that a 7-day accelerated aging test at 302*F is equivale r ta at least 40 years operation at maximum normal Containment temperature. This le in addition to a 30-day test at Containment post-LOCA condi: lone (see Note 5). Justi fication for Continued Operation Pending Correct ive Action This equipment is considered qualified.
r TABl,E 1 TROJAN NUCLEAR PLA' T ENVIRONNENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND GENERAL ELECTRIC COMPANY'S ~ - RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 8 of 41 Equipment De sc r i pt ion Manu f ac t urer Model No. Cumponent No. Location Function Feed Through CONAX Mll001-31 C Environmental seal NRC-Identified De ficienc y Resolution of Deficiency Temperature and pressure test curves were inadvertently omitted in T. P previous submitt ele to NRC. Both test parameters envelope specified requiremente. H This equipment is qualified for a steam / water environnent, which is con-sidered acceptable as discussed in Note 2. CS This equipment is qualified by test for a masimum pH of 10.5, which meets the specified requirement. Also see Note 4. A Analysis of test data using the Arrhenius methodology demonstrates that a 30-day test at elevated temperatures is equivalent to at least 40 years opera' ' me at maximum normal cont ainment temperature, plus 30 days opers. ,n at Containment post-LOCA conditions (see Note 5). Justification for Continued Operation Pending Corrective Action This equipment is considered qualified.
TABIE 1 TRtUCD NUCLEAR FRANT ENVikONMENTAL QUALIFICATION OF SAFETY-kELATED ELECTRICAL Equ!PMENT FORTI.AND CENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 9 of 4) Equipment Description Manufacturer Model No. Compenent No. Incation Function Cable Okontte IC, 3MIMCM, A13 A High-voltage cable %V, EF NRC-Identified Deficiency Resolution of Def tetency QT, F, it 1his equipment to located in the Austliary Sutiding and is not subject to abnormal environmental conditione during post-accident operation, except for radiation esposure during long-term recirculation. Adequate operating time folloutng an accident is demonstrated by enveloping the spectited radiation requirement. Analyste of test data using the Arrhentue methodology demonstratas a A 21.5-day test at elevated temperatures le equivalent to at least 40 years operation at mas.eus normal Auxiliary Butiding temperature (see Note S). Justification for Continued Operation rending Corrective Action e This equipment to considered qualified.
g 1 l t 4 t a p f t ec e e o e a l i oc h sp l v t r t x t ee 0 t l ut e r c 1 ua n oc t g us oc ol e c, n t u t i i a kl e n e. i as e rr t c ne jo p r e et ai a bi ao et h tc l r rl ut ul po S ne ot A saqe mn al sc e6 r ev e h e i e v5 t ed n t e c l i0 opA e r s e t e t 2 no d a er n n n i n y e uo e) o d0 o t t .b t n n t t i m3i dMi n nn ao o aat n6t a( t eeod vi i erc i 6 c c mdi e et t h ee a5 e yee nit t l c c pn t n av n i caa eu n R mn nOn rl n acl r r u R eo oMo pao t aut t t F Rt c C( c S vc n-cs as ot rn n C sio t o ocm nc epe e e d rd ef i g po s nme l n t i rl as o ur e gl i out t ea T t d - n i T R e A P P F s ge sr N O c aend i e E P a enoi. h t n M E L r ol ct t a o P R ai cn m. i I t gae f t N nin m osn c T. O idi ne eo A E I nr ar dt i T d ou i eat e L A F e cJ g u t c a v A U c t aq cid i C L n al el e ud a t I A e car ur d nt c R V i ot uo nis e T E c l nsl n o e r C S o i e ol o csd r E T N f empoi i o L 'T T e l nsf t ssl C E NE t D oe a aya AF n t r ei wl m g D PA e f ninmd ar n E mS n 9 0 5 0 o e v oia t ne i T t o 0 4 2 6 mnit r sah dn TA CC p 4 4 4 4 n pet e ,t NL l m B B B B o i agd t e lAE CM o T T T T i ul ine r o P R I C A A B A t qadii yet P - RN u ematf av n T TI l r rai d oe o 1 RT C o so rc - el i AE ES s i nr ee 6 rb t E EF LE e hb opp ot a LA EI R T af os AMt r I CS C e RA U LN p T NF AE O O RI N EC d AN NA e JO EF o u OI CE N n i RT I TA D l t C ND e 5 n AE d o I F t I u B C I TF H E L RI r A OT o U PN f Q ED no L I i F t "F O ac Fe N i t O r f Of U i e t R r s I U u u V L t J N O c E S a E f R un d e a E i M C f i lauq dere n d i o s i n t o p c i rc d e l s e e i y t D k f c c i n n t o t e e n l ni m e B ec p i m di u p l I f i a - e q u n CD e q i R H E m N s r i e T h T q A T (
I TABl.E 1 TROJAN NUCLEAR FRANT ENVIRONteNTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND CENERAL ELECTRIC COMPANT'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EYALUATION REPORT Sheet it of 4) Equi pme nt De sc r ipt ion Manu f ac t ure r Moded No. Component No. Location Function Limit suitch NAMCO EA 17011302 ZS 8811, 8814 A Valve position indication and permissive interlock for MO 8004A&B ZS 8883 A Valve position indication and seat in ZS 8964, 8152 P Containment isolation valve 8028 position indication and aesi in ZS 8811A&B P Valve position indication and permissive interlock for MG 8004A&B NRC-Identified De ficiency Resolution of Deficiency H, P This equipment is located in areas outside Containment not subject to abnormal environmental conditione during post-accident operation, except for radiation esposure during long-term recirculation. QN, A Analysis of test data seing the Arrhenius methodology demonstrates that for all materials, except for the contact block and cont act carrier, a 200-hour test at 200*F is equivalent to at least 40 years operation at maximum room t emperatus es. The cont ac t block and cont act carrier used in the EA 170 are ident ical to the ones used in the EA 180. Analysis of the EA 180 test date using the Archenius methodology demonstrates that the EA ISO aging and LOCA teste subjected the contact block and contact carrier assemblies to at least 40 yeare of equivalent thermal aging at seminum room temperatures. Just ification for Continued Operation Pending Correct twe Act ion This equipment is considered qualified.
TABI.E 1 TROJA3 P*iCLEAR PLANT ENVIROlelENTAL QUALIFICATION OF 1AFETY-RElATED ELECTRICAL t.QUIPMENT PORTIAND ElERAt. ELECTRIC LXDtPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN hRC SADETY EVALUATION REPORT Sheet 12 of 41 Equipment Description Manufacturer Model No. Component No. location Function Instrument Cable Rockbestos 2C, #16 E02 A Conductor 600V, ILPE letC-Ident if led Deficiency Resolution of Deficiency QT,P H T Wie equipment is loc ated in the Austliary Building and is not subject to abnormal environmentail conditione during post-accident operatloa, escept for radiation exposus e during long-tere recirculation. Adequate operating time folloutng an accident is demonstrated by enveloping the specified radiation requirement. Analyste of test dataa using the Arrhenius methodology demonstrates that QN, A an 850-hour accelerated aging test to equivalent to 40 years operation at meuteum normal Aust!!ary But!!ng temperature (see Note 5). R Wla equipment le quellfled by test for a redletion dose in escess of she 8 5 specifled qualifIcat ton requirement (2 m 10 rede vs 4.0 m 10 rade). Justification for Continued Operation Pending Corrective Action This equipment to considered qualifled.
0 = TABLE 3 TROJAN C3 CLEAR PLANT ENVIRONMENTAL QUALIFICATION C7 SAFETY-RELATED ELECTRICAL EQUIPMEYf PORTIAND CENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 13 of 43 Equipment Description Manufacturer Model No. Component No. Location Function Pressure Transmitter Ba rton 763 PT 405 C RCS wide-range pressure indict-lon, RRR isolation valve interlock, and input to RCS subcooltag margin monitor 763 PT 455, 456, C RCS pressurizer pressure 457 indication and RPS/ESFAS in put Differential Transmitter Barton 764 LT 517, 518, 519, C Steam generator level 527, 528, 529 indication and RPS/ESFAS 537, 538, 539, input 548, 549 764 FT 512, 513, C Steam line flow and 522, 523 RPS/ESFAS input 532, 533 54 3 NRC-Identified Resolution of Deficiency Deficiency This equipment is qualif ted by test for a maalaus pH of 8.5 Although CS this does not envelope the maximum specified qualification require-ment of pH 10.5, these transmitters have sealed steel housings and are qualified for water submergence and a steam environment. Therefore, they are not subject to failure from a wider pH range. See also Note 4. QI, QN, A (QI for LT 527 and FT 512, 513, 522, 523, 532, Previous testing diet not include accelerated aging. Fu,ther testing by Westinghouse and an.nlysis are under way to establish a qualified lif e. 533, 543 only) This equipment is q2alified by test for water submergence and a steam H environment. Therefore, it is considered qualified for a 100 percent relative humidity esvironment. T. P (for LT 527 and FT 512, 513, 522, 523, 532 This equipment is qaalifted by test for a maximum pressure and temperature of 75 psig and 320*F respectively, which exceeds tre maximum specified 533, 543 onlF) conditions of 60 peig and 288*F. Moreover, the 16-day test profiles envelop the FSAR Containment analysis pressure-temperature profiles (see Note 1). Justification for Continued Operation Pending Corrective Action This equipment is considered qualifted, except for aging. Operating experience to date has revealed no identifiable age-related failures. Existing Plant surveillance and maintenance procedures are suf ficient to correct age-related (or other) fattures by repair or replacement of equipment. Further testing and analysis is underway to establish qualified life.
TABlf 3 TROJAN NUCLEAR PLANT ENVIRONENTAL QUALIFICAtl0N OF SAFETY-RELATED ELECTRICAL EQUIPENT PORTLAND CENERAL ELECTRIC 00MPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NBC SAFETY l' VALUATION REPOR1 Sheet la el 43 Equipment Description Manu f ac t ur e r Model No. Component No. Loc at ion Function RTD Rosemont libEF TE 410A&E C RCS temperature indication ellA&B and RFS input 420 A68 42BA&B 430A&B 431A&B 440A&B 44 t A& B 176KS TE &l3A&B C RCS loop temperature 423 A& R indication and input to 433A&R oubcooling margin monitors 443A&B NRC-Identified Resolution of Deficiency De fic i enc y This equipment is qualified by test f or a steam environment which is H acceptable as described in Note 2. Moreover, RfD is enclosed in a metal thimble and therefore le not subject to humidity degradation. This equipment was tested in a boric acid environment but not a caustic CS spray environment. Novever, the RTD is enclosed in a metal thimble and therefore is not subject to chemical spray degradation. Analysis indicates materiale of construction are not susceptible to A thernel degradation. Justification for Cont inued Operation Pending Cc trect ive Action This equipment is considered quali fied. I i I
TABl.F. 1 = TROJAN NUCLEAR P! ANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUtrnENT FORTIAND CEDERAL ELECTRIC COMPANT'S ~ = RESOLUTION OF IDENTIFIED DEFit"ENCIES IN 80tC bAFETY EVALUATION REPORT Sheet 15 of 43 Equipment Description Manufacturer Model No. Component No. Incattun Function Motor Allio-Chalusto Type CV MP 204A&B A CSS pump motor driver NRC-Identifled DefSciency Resolutton of Deficiency QT, T r, H 1hte equipment le located in the Ausittery Building and is not sub jec t to abnormal environmental conditions during post-accident operettoa escept for radiation esposure Juring long-term recirculation. Adequate operat-i ing time following an accident le demonstrated by enveloping the specified radiat ton requirement. (pt, A Ongoing vendor performance testing so far equivalent to 27 years con-tinuous duty operation. Espected duty f rom 40 years of planned periodic operation and I year post-accident operation would be much less than 27 years continuous operation. Since motor aging to significantly accelerated during operation due to temperature rise, standby aging to insignt f Scant. Justification for Continued Operation Fending Corrective Action This equipment to considered qualt fled.
TABLE 1 TROJAN NUCLECR PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPS.NT PORTLAND CENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFILD DEFIClLNCIES IN NRC SAf ETY EVALUATION REPORT Sheet 46of45 Equipment De sc r i pt ion Manu f ac t urer NoJet No. Compos eent No. Location Function Solenoid Valves ASCO NP8)l654E SV 8149A&At C Containment isolation (NP series) SV 81495&B1 valve actuation SV 8149C&cl SV 8026 SV 455AA&At C RCS pressut iner power-456A&R operated relief valve ac t uat ' >n NRC-Identified De f ic ienc y Resolution of Defic iency P Original specified reguirement for pressure was not enveloped by tes? curve (i.e., 20 peig required wo 15 peig test at Il-to 24-hour time interval). However, test conditione do envelope FSAR Cont ainment analysis (see Note 1). Moreover, qualification testing at higher pree-eures early in the test sequence (110 peig) demonstrates component has no significant failuse modes duw to preneure. H This equipment is qualified by test for a steam environnee which is acceptable as descrit ed in Note 2. CS This equipment to qualified by test for a manimum pH of .5, which does not envelope the manimum specified quali fistir. requer<. ant of 10.5. This is considered acceptable, however, because the component is enclosed in a NEMA 4 steel cover which is not susceptible to corrosive ef fects at the higher level. Saee also Note 4. A Analysis of vendor aging data indicates a qualified life for this equip-ment of 9 years in the naminum normal Cont ainment temperature at Trojan. This is in addition to 30 days operation in a post-LOCA Containment environment (see Note 5). Justification for Continued Operation Pending Corre.:tive Action This equipment is considered qualified for an installed life of 9 years. Periodic replacemeat schedules will be implemented to assure equipment change-out by end of qualified li fe.
TABLE 3 TROJAN NUCLEAR PLANT ENVIRONNENTAL QUALIFICATION OF SAFETT-REIATE') ELECTRICAL EQUIPMENT a PORTLAND CE NERAL ELECTRIC COMPANT'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETT EVALUATION REPORT Sheet 17 uf 4) Equipment De sc r i pt ion Manu f ac t ure r Model No. Component No. Loc at ion Func t ion Power Cable General Cable 2C,f l 4,600V.E P N02 M. P Conduc t or SC.J14,600V,EP N05 C 7C,fle,6DOV,EP N07 C 9C,f l 4,600V,EP N09 C IC,f4/0,600V,EP PO5 C IC,ft/0,600V EP P07 C 3C,f8,600V.EP PBS C 3C,fl0,600V,EP P16 C 3C,fl2,60GV,EP Pl7 C IC f 6,600V,EP Pl8 C 2C,f t 2,600V,EP P20 N NRC-Identified De ficiency Resolution of Deficiency R. T, QT, P H CS, A QH S (for P07 and P20 only) NRC review van based on 9/5/80 submittel, which inadvestrw4ty omitted all qualification information for these componente. R (N02 only) This equipment is qualified by test to a radiation dow a 1-scess of the J 8 red vs. r-w+ t background specified qualification requirement (2 x 10 levels). For M02 and P20, ohich are outside Containment, there le no specified CS qualification requirement for chemical spray. The remaining componente inside Containment are qualified by test for a pH of 9.0, which does not envelope a specified pH of 10.5 (see Note 4). Houever, this le considered acceptable because! EPR insulation material is not susceptible to acid ur caustic degradatisen. This equipment is qualified by test for submergence (see Note 3). 3 Analysis of test data using the Arrhenius methodology demonstrates a QH, 4 7-day test at elevated temperature le equivalent to 24 years operation et easieue normal Containment temperature, plus 120 daye et Containment post-LOCA conditione (see Note 5). Justification for Continued Operation Pending Corrective Action This equipment is considered qualified for an installed life of 24 years. Evaluation le continuing to entend qualified tile to 40 years.
TABl.E 3 TROJAN NUCLEAR PLANT ENVIRONElffAL QUALIFICATION OF SAFETY-REl.ATED ELECTRICAL EQUIPENT PORTLAND CENERAL ELECTRIC COMPANY'S ~ ~ RESOt.UTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 13 of 43 Equipment Desc ript ion Manu f act urer Model No. Component No. location Function Instrument Cable American Wire 2C/S,f!6,300V EP IO2 C Conduc tor & Cable 4C/S,fl6,300V,EP 104 C 2C/S,fl6,300V,EP S02 C 2C/S,fl6,300V,EP 202 C 4C/S,#16.300V,EP 204 C NRC-Ident i fied De fic i enc y Resolution of Deficiency P Equipment to quali fled by teet for specified containment pressure. CS Equipment is quali.fied by test for a masinue pH of 9.0, uhich does not envelope the speci.fied pH of 10.5 (see Note 4). However, this is considered acceptable because EPR insulation material is not susceptible to acid or caustis; degradation. M, QM, A Analysis of test stata using the Arrhenius methoJology demonstrates a 2-day test at elevated temperature le equiealent to 24 years operat.on at maximum normal Containment temperature, plus 120 days at Containment post-LOCA conditienne (see Note 5). S Equi m nt le quaitfied by test for subenergence (see Note 3). Justification for Continued Operation Pending Corrective Action This equipment is considered qualified for an installed life of 24 years. Evaluation is cont nuing to estend qualified life to 40 years. 1
TABLE 3 TROJAN NUCLEAR FIANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND GENERAL ELECTRIC CONFANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 19 of 43 Equipment Description Manufacturer Model No. Component No. Location Function Limit Switch NANCO EA 180 ZS 8701, 8702 C walve status indic6tton and 4.rmissive interlock for HQ 8004A&B. ZS 8026, 8871 C ' Containment teolattoa valve status indication ZS 455A, 456 C LCS pressurtser power-operated rettet valve status Indication NRC-Identified Def tetency Resolution of Deficiesty X Equipment to qualif ted by test for a stese environment, which to accept-able as described in D'ote 2. CS Equipment te qualtfled by test for pH 11.0, which envelopes maximum spectiled requirement of pk 10.5 (see also Note 4). A Analyste of test data using the Arrhentuo methodology demonstrates e 7-day test at elevated temperature is equivalent to 18 years operation at masteue normal Containment temperature, plus 30 days at Containment post-LOCA conditions (see Note 5). Justification for Continued Operation Pending Corrective Action =- This equipment to conaldered quallfled for an installed life of 18 years. Periodic replacement schedules wt!! be implemented to assure equipment change-out by end of qualtfled life.
F e* TABLE 3 TDJA] NUCLEAR PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUlPMENT PORTLAND GENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Bheet 20 of 43 Equipment De sc ri pt ion Na mu lac t ur e r Madel No. Component No. Location Function Solenoid Valve ASCO Va r i ou s SV 8870A&B A BIT isolation valve actuation (Non-NP series) 8883 SV 8871 C Containment isolation valve 4000, 4006 P ac t uat ion 4181, 4303 P 8013, 8028 P 8I52 P 10004, 10014, 10015 P 8880, 8888 A 10001, 8964 M SV 1782, 1783 A RHR ensple isolation valve actuation SV 8875A,5,C.D C Accumulator tank isolation 8877A,B,C,D valve actuation 8878A,B.C.D 8879A.B.C.D 8881 SV 2295A&B M Steam line isolation 2297A&B valve actuation NRC-Ident i fied De f ic i enc y Resolutiog of Deficiency CS, QN, et al (all except SV 2295A&B and SV 2297A&B) These ASCO s on-NP series solenoid valves control air-operated valves that attain their 1esired safety position upon loss of air. Westinghouse analysis demonatrates that for design basis accident conditions the only credible solenoid f ailure mode is in a position which vents the air operator, thus placing the associated vales in its safe state. R QM (for SV 2295A&B and SV 2297A&B) Failure of these solenoide in an MSLS environment outsidi Containment would cause their associated air-operated isolation valves to f ail in the open position. This is an acceptable f ailure position since neither steam line isolation nor auxiliary feeduster flow lajection would be inhibited due to the small eine of these lines (1-inch). Moreover, line isolation could be ef fected manually by closing downetream valves. A (all) Operating experience (IE Bulletin 78-14) has shown age degradation of diaphram can cause f ailores in other than the vent posit ion for service life beyond 7 years. Justification for Continued Operation ren'dina Corrective Action ~ This equipment is considered qualified for an installed life of 7 years. Periodic replacement schedules will be implemented to accuse equipment change-out by vivt of qualified life. I
TABl.E 3 TROJAN NUCLEAR FRANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND GENERAL ELECTRIC COMPANT'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 21 of 43 Equipment heocription Manufacturer Model No. Component No. Location Function Differential Transmitter Flecher Forter 13D2493 LT 2069A&B A CSS spray additive tank level indication 1052495 FT 3043C&D M AFW flow indication NRC-Identified Deficiency Resolution of Deficiency A, QT, T. F. M, QM, Ql, RFS (for LT 2069A&B) This equipment la located in the Auxiltary Building and to not subject to abnormal environmental conditione during post-accident operatice, except for radiation exposure during long-term recirculation. Howeve., because this equipment performa its safety function during the short-tere injection phase, no radiation qualification is required. A, H, QT, QM, RF (for FT 3043C&D) Additional qualified AFW flow transettters were installed during the 1981 refueling outage. These units (FT 3043C&3) are Rosemont 1153As which are qualtfled to IEEE 323-1971 (NUREC-0588 Category II). See Table 5 for qualification summary information applecable to thle equipment. Justification for Continued Operation rending Correttive Action 1. The spray additive tank level transmitters are except from qualification on the basis that the equipment perforse its function before esposure to a harsh environment. 2. For AFW flow indication, redundant level indication is provided by Rosemont Il53A dif ferential transettters (FT 3043C&M) which are qualified to IEEE 323-1971.
F TABIE 3 TROJAN NUCLEAR F1 ANT ENVIRONMENTAL QUALIFICATION OF SAFETT-RELATED ELECTRICAL EQUIPMENT PORTLAND CENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICLEP.CIES IN NRC SAFETY EVALUAf tDN REPORT Sheet 22,,f 41 Equipment De sc ri pt ion Manu f ac t ur e r Model No. Component No. Location Function Pene t r at i ons Amphe nol A2 03 C conduc t or A2 03 A2 05 A2 07 52 01 B2 01 52 05 B2 07 C2 07 D2 07 MRC-Identified Resolution of Deficiency De fic ienc y Analysis of test data using the Arrheniue methodology demonstrates that a QN, QT, A 24-hour test at elevated temperature is equivalent to 102 days operation post-LOCA conditions (see Note 5). Radiation testing and analysis of atmateriale of construction indicate equipment is probably not susceptible to radiation or thermal degradation. Further analysis is underway to establish qualified life. T. F Equipment is qualified by test at a maximum Containment condition of }00*F and 60 peig. Test envelope esceeds specification encept for preneure during a short time interval (20 peig we 20-30 pois at 900 to 2000 seconde). This is considered acceptable because pressure testing at higher pressures early in the test sequence demonstrates no significant equipment failure mode due to pressure. N Equipment is qualified by test for a etcom environment, which is con-sidered acceptable as discussed in Note 2. CS Equipment le qualified by test for a maximum pH of 10.0, which does not envelope a specified pH uf 10.5. However, this is considered acceptable se discussed in Note 4. See discussio.n under justification below. Ql, RFN Justification for Continued Operation Fending Corrective Action This equipment is considered qualified, except for aging. Operating esperience to date has revealed no identif eeble age-related f ailures. Existing Plant surveillance and maintenance procedures are suf f scient to correct age-related (or other) f ailures by repair or replacement of equipment. Further analysis is underway to establish qualified life.
TAB 12 3 TROJAN NUCLF.AR FRANT ENVitOletENTAL QUALIFICATION OF SAFETY-REIATED ELECTRICAL EQUlFNENT FORTIAND CENERAL ELECTRIC CONFANY'S ~ RESOLUTIP. OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 23 of 43 i Equipment Description k nufacturer Edel No. Component No. Location - Function Level Switch Flecher Porter 1338365 LS 2074, 2075 C Containment rectreulation suep level indication T IstC-Identif ted Def tetency Resolution of Dettelency QT, T F. H, a '. %, 4, 5, QH. RFS 1hte equipment was replaced with Ratton 764 level transetttere during the 1981 refueling outage, which are considered quattfied, encept for aging (see Sheet 13). 1 l l Justification for Continued Operetton Pending Corrective Action Corrective action has been taptemented for thte equipment.
TAB 12 1 TROJAN NUCLEAR FtANT ENVIROOMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT FORTIAND CENERAL ELECTRIC CONFANY'S RESOLUTION OF IDENTIFIED DEF ACIENCIES IN NEC SAFETY EVALUATION REPORT Sheet 24 of 41 Equipment Description Manufacturer Model No. Component No. Location Function I/F Converter Flecher 546 FY 606 A RRRS flow contro! 601 NRC-Ident if ied Resolution of Deficiency De f ic iency QT, T. F. W. QN This equipment is located in the Aust!! sty Building and le not subject to abnormal environmental conditione during post-accident opera-tion, except for radiation esposure during long-term recirculation. Adequate operating time following en accident is demonetiated by enveloping the specified radiation requirement. Dets recently received f rom vendor indicates equipment la qualified for QI, R specified radiation levela during long-term recirculation phase (1 m 107 rade we 4.5 a 105 rade). Aging analyste of materlate of construction using the Arrhentue QN, A methodology shows equipment to have a qualtfled lif e of 15 years (realistic estimate). See discuselon under justification below. R PS Justification for Continued Operation Pending Corrective Action This equiseent is considered qualified for an installed !!fe of 15 years. Evaluation is contlautng to better define qualified life. Periodic replacement schedulee util be implemented, based on results of further analysis, to aneure equipment change-out by end of qualified Af fe. l
j' ~ 3 4 w y o e f g hl sv n o r cl ei a est o i e it j r 5 neaf gr r a o o 2 et h na sa t r f - a e E i mn. T t h cd rl s h me) d e gl aua e t usyo e e ia esc. r set t t sid u r ri n ah h l h r e s es ot pd y e S o er yt a f s eit m r t vph ae er mi e n nco ppm et ul l o i t i l shi p c asl nd cl b cn b m t ul enia aioea i n w noaiat n l ri vc o o ormb no i t t ii e i l c smeai acitl b t f e ar t v ed ap c th a uti al nl p d n W ot t sod. Eoea l u F n aaondt n cr u F A ,i p an el r o srvmx s e eaot o c ee aeif m b rt nf ovae i p e e n d eet f e i snd ca o wr soreu aeert i t oa unq hG t ea t aH onl i e s pd a n h et oim ad e r o t n i ar g t et 0 s e i . i d t f en an 0i p ,t di n1 h o T t mnl ea i T R a M oo s r et ae t N O c oir outde t t ed l E P n rsepd n t sbb ne n a M E I rt x eos eo at o u P R aeaet mt e t s a i n I mw r p yaF u t a U N nbd yoiyt l l h* l c m Q O i uel h uac at 2 a A E I sef sqwe t nt 1 v l T d f e ert nen2 e e a L A y eo i e eo ece( v c A U c t t d rhtd r mem o i o C L n a nbt nnp nr pSt t l I A e cea au o ,iS c R V i on eod r r uS y e T E c l ombt nee i r qMa r C S o i ra urh vee w r n E Y N f sped ed at nv er o o L 'Y T 1212l2i2 e i tl ue r eerh e C i D rsu wat d t E NE t AA5BCCDD D t s ai AF n 44444444 t o o ef ool n g c nnnv.fi nHi nu n n D rA e 00000000 f e i od o mi i u E MS n 00000000 o T O o 33333333 msadB ut e d f .i TA CC p n pi mnL nt e sE r n NL R m S o i aE osd rt L a e y AE CN o I i ud e Hi e oneE F t LR I C F t qnh g t gt f et H s e P - RN u eat onann oma t n f Y TI l t aria t pcnr o a 3 RT C o seh n arr eiiao i s AE ES s int egper r ud f t E EF LE e hiipneea hqnff a l L LA EI R 1l woisnw I eioE r a B CS C e i A U LN p t T NF AE O n O RI e N C d s AN EA e s J0 EF o u e O1 GE N n i s RT I TA D l t t C ND e A n i AE d 8 o I 8 C m F I I P 2 r I TF r o L RI A OT o f U F N f r Q E e D n p o L I i l AT F t l N O a i E c w M N i f t N O r i n O I e t e R T r I U u s m u p V L t J i N O c E S a n uq E f o e R u t n r s a a i M B h t t a S h P t R s t l s i n Q s. o h el i c t t N o p i Q er ct i w nn r S c d A ao s g e rc u e n i y D i f c R sw t i n so t a t e al f n t ni H e c ec e l m m i di p d I f F b r i n - e ae nt u I CD q e T o-t sg E w l an o l T eo F Q Rl
TABLE 3 TCDJAN NUCLEAR FtJCT ENVIRONMENTAL QUALIFICATION OF SAFETT-RELATED ELECT! ;AL EQUlPMENT N)RTLAND GENERAL ELECTRIC Cl1NPAFf *S Sheet 26 of 43 RESOLUTION OF IDENTIFIED DEFICIENCIES IN NMC SAFETT EVALUATION REPORT Equipment De sc ript i on Manu f acturer Model No. Compement No. Location Function Presure Transmitter Fosboro EP3DN LT 459 C Pressuriser level indi-460 cation and RPS/ESFAS input 4LI NRC-Identified Resolution of Deficiency De ficiency This equipment is qualified by test at temperature and pressure condi-tiene that exceed the specified Containment envelope (see Note 1). l T, P Equipment is qualified by test for a steam environment, which is con-N sidered acceptable as discussed in pote 2. l Equipment is qualified by test for a naminum pH of 10.0, which does not CS envleope a specified pH of 10.5. However, this is cor.sidered acceptable because the transmitter is sealed in a metal housing. See also pote 4. This equipment is bnown to cont ain wateriale susceptible to thermal age QI, QN, A degredation. Further analysis and testing is currently underway to esteSlish qualified life, l Equi pment is qualified by test for an inte rated dose exceeding the i R spec, fied level (2 x 108 rede vs. 3.8 m 10 rade). l See diacussion under justification below. RPS Fosboro l*tter of 3/12/81 (Attachment I to SER) not applicable to this e equipment. Justification for Continued Operation Pending Corrective Action This equipment is considered qualified except for aging. Operating esperience to date has revealed no identifiable age-related f ailures. Esisting Plant surveillance and maintenance procedures are suf ficient to correct ege-related (ur other) failures by repair or replacement of eq ui pme nt. Further analysis and testing is underway to establish qualified life. (\\
TABLE 3 M RJA2 NUCLEAM PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED E11.CTRICAL EQUIPMENT . 2eTIAND GENERAL ELECTRIC CUMPANY'S RESOLUTION OF IDENTIF13 DEFICIENCIES IN MLC SAFETY EVALUATION REPORT Sheet 27 of 43 Equipment Description k nufacturer Ndel No. Component No. tocation Function RTD Rurne FOR-320 TE 463 C RCS pressurizar safety and t 464 power-operate 1 relief 465 valve tailpipe temperature 466 Indication I NRC-Identifled De f icienc y Resolution of Deficiency QT, T. P H. CS, R. A. QH, QI, RPS This equipment only provides indirect means of monitoring the ope -close status of the preneuriser safety and power-operated relief values. Other, more direct means are available for monitoring valve pos! tion (via, ZS 455A & 456 and SE 11894,R&C). Moreover, monitoring tatlpipe temperature is not an essential safety function. L I i i [ Justification for Continued Operetten Pending Corrective Action This equipment le exempt from qualification on the baste that it does not perfore an essential safety function in a harsh environment.
TARLE 3 TMOJAN NUCLEAR PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND CENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 28 of 43 Equipment Description Manu f act urer Nid.fl No. Component No. Location Function Acoustic Monitor TEC $00 SE Il89A C RCS pressuriser safety 1889R walve position indication ll89C NRC-Ident i fied De ficienc y Resolution of Deficiency QT, T. F. M CS, R A. QN, QI Qualification information for this equipment was not previously avail-able. Qualification testing has been recently cumpleted but test resulte sh=w this equiprant to be unqualified. Development of qualified equip-ment to reple,e the failed componente is underway by the equipment ve nd o r. Justification for Continued Operation Pending Corrective Action This equipment is considered unqualified. Nowever, neither these componente, nor their associated valves, perform an essential safety fue4 tion in the harsh environment resulting f rom a large-break LOCA or NSLS inside Containment. 4
m TAB L E 1 T[3JAN NUCLEAR FIANT ENVIRONNENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUlPNENT-p0RTLAND CENERAL ELECTRIC COMPANY'S RESOLUTIOel 0F IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 29 of 43 Equipment De sc ri pt ion Manuf acturer Model No. Component No. Location Function Radiation Monitor Victorcen 847-1 RE 6101 A Radiat ion monitor (identi-6102 fication of ECCS post-accident leakage outside containment) NRC-Identified Resolution of Deficiency De ficiency I QT, T. P, H This equipment is located in the Auxiliary Building and is not subject to abnormal environment al conditione during poet-accident opera-tion, except for radiation esposure during long-term recirculation. Additional analysis of materials of construction is underuey to QI, QN, R A RPS ascert ain radiation and aging qualification. l l Justification for Continued Operation Pending Corrective A. tion This equipment is considered unqualified. RedunJant means of detecting ECCS leakage is available via Auxiliary Building HVAC radiation monitors located in a non-harsh environment.
TABIE 3 TROJAN NUCLEAR PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND GENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Shee t 30 of 43 Equipment Description Manufacturer Model No. Component No. Location Func tion Hydrogen Recombiner Westinghouse Electric RE 318A&B C Mydrogen control NRC-Identified Deficiency Resolution of Deficiency QT Analysis of additional test data using the Arrhenius methodology demon-strates that the test profile is equivalent to at least 30 days at post-LOCA conditions. Therefore, the equipment is qualiffed for the specified post-LOCA operati g duration. P Recently obtained test data shows equipment was qualified to pressures exceeding the maximum specified accident condition (60 peig) during 2 of the ? LOCA test cycles for a total time of 8 hours. There f ore, this is acceptable since it envelopes the worst-case specified condition. H This equipment is qualified by test for a steam environment, which is considered acceptable as discussed in Note 2. CS This equipment is qualified by test for a maalmum pH of 10.0, which does not envelope the maximum specified pH of 10.5. This is considered acceptable, however, as discussed in Note 4. A Analysis of additional test data using the Arrhenius methodology demon-strates an accelerated aging test of the heaters is equivalent to >40 years at maximum normal Containment temperature. Qi, RPS Previous submittals to the NRC indicated the electrical interface terminations and internal cable were of unknown qualification. Test data recently obtained from vendor indicates internal cabling is fully qualified for all specified environmental parameters and has a demon-strated qualified life, via an Arrhenius analysis of accelerated aging test data, of at least 40 years. The electrical interf ace terminations will be replaced with materials of known qualification by 6/30/82. Justification for Continued Operation rending Corrective Action This equipment is considered qualified except for the electrical interface terminations. However, failure is unlikely because terminations are enclosed in sealed metal housing, are mechanically separated, and typical termination materials are not generally susceptible to post-accident environmental conditions.
TABLE 3 TE3JAN NUCLEAR FIANT ENVIRONNENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTIAND CENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NNC SAFETY EVALUATION REPORT Sheet 31 of 43 Equipment Description Manu f ac t ur e r Ndel No. Component No. location Function Fan Motor Westinghouse Type SBDP NV 25tA&B A CCP room air cooler 252A&B A CSS pump room air cooler 253A&B A RNR pump room air cooler 254A&B A SIS pump room air cooler NRC-Identified Deficiency Resolution of Deficiency QT. T. P. H. RPS This equipment is located in the Auxiliary Building and is not subject to abnormal environmental conditions during post-accident operation, except for radiatiu exposure during long-tere recirculation. Adequat e operating time follouing an accident is demonstrated by enveloping the specified radiation requirement. QlQNA Plant motors are operated only during testing and post-accident condi-tions. Since motor aging is significantly accelerated during operation due to temperature rise, standby aging is insignificant. R Nre detailed analysis has recently shown the radiation qualification requirement to be within the capability of the equipment. Justification for Continued Operation Pending Corrective Action This equipment is considered qualified.
s
- ~
TABI.E 1 TROJA3 NUCLEAR PLANT ENVIROMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND GENERAL ELECTRIC CuMPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 32 of 43 .ipment Description Manufacturer Model No. Component No. Location Function ,otenoid valve R. C. Laurence 12541AW SV 2216A B.C.D M Main steam line isolation SV 2236A B.C.D valve actuation SV 2256A.R.C.D SV 2276A B.C.D a NRC-Identified Deficiency Resolution of Deficiency QT, T. F. H R. A, QM, Ql, RPS This equipment is located in a room that does not contain a high-energy line and is not prone to submersion. However, the room communicates with the main steam and feedwater line area via a small overhead hatch opening d and would be briefly exposed to stene at ambient pressure following an HELB. Due to the short duration of exposure and physical separation of redundant equipment, f ailure is not anticipated. Engineering studies are underway to determine if additional measures are warranted to further protect existing equipment. Environmental qualificatJon test data is unavailable for this equipment. Justification for Continued Operation Pending Corrective Action Reasonable assurance exists that this equipment will perforu its essential safety function. Lacal manual operation could be taptemented for trip function.
TABLE 3 TROJAN NUCLEOR PlJCT ENVIROMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAI. EQUIPMENT PORTLAND CENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 33 of 43 Equipment Description Manu f ac t ur e r Ndel No. Component k. Location Function Terminal Board Square D 828 ATB 203-210 C RCS temperature electrical STB 203-210 connection cts 203-206 Drs 203-206 NRC-Identified Deficiency Resolution of Deficiency QT, T. P. M. CS, R, A. QM, Ql, RPS Qualification testing has not been performed for this equipment. Nouever, analysis of material and seametrical similarity to other quali-fied unita (CE EB-5, nochanan 2n100 and Marathon 1600) and analysis of material susceptibility to radiation and thermal degradation sapporte qualification for plant life and post-accident operating time. Justification for Continued Operation Pending Corrective Action This equipment is considered qualified. m
~-. TABLE 3 11t0JAN NUCLEAR PLANT I ENVIRONMENTAL QUALIFICATION (O SAFETY-RELATED ELECTRICAL. EQUIPMENT I PORTLAND CENERAL ELECTRIC. COMPANY'S RESOLUTION OF IDEfff!FIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 36 of 63 i Equipment Description Manufacturer Model No. Component No. Incation Func t ion Motor West inghouse Type HSDP KP 202A&B A RHR pump motor driver 203A&B A SIS pump motor driver 205A&B A CCP motor driver i u NRC-Identified Deficiency Resolution of Deficiency QT, T, P, H This equipment is located in the Auxiliary Building and is not subject to abnormal environmental conditions during post-accident operation. 1 except for radiation exposure during long-term recirculation. Adequat e l operating time following an accident is demonstrated by enveloping the specified radiation requirement. QI, R Equipment is qualified by test for specified radiation levels during 8 rads vs 2 x IOI rads). long-term recirculation phase (2 x 10 Aging analysis based on Westinghouse insulation testing shows life much QM, A longer than expected operating time of motor. Insulation materials and lubricants have been determined not to be susceptible to degradation f rom specified radiation levels. Justification for Continued Operation Pending Corrective Action This equipment is considered qualified.
l TABLE 3 TROJAN NUCLEAR PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND CENERAL ELECTRIC COMPANY'S Sheet 35 of 43 RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Equipment Description Manufactuter Model No. Component No. Location Func tion Reliance Series 2000 h 201A&B C Containment air cooling Motor 202A&B 203A&B 204A&B M 220A&B C Containment hydrogen alming NRC-Identified Resolution of Deficiency De ficiency Equipment qualified by test at temperature and pressure conditions that T. P exceed the specified Containment envelope (see Note 1). However, test envelopes were cyclical rather than conforming to specified profiles. This is considered acceptable because a series of cyclic tests impose more severe stresses than specified Containment profile. Equipment qualif ted by test for a steam environment, which is considered N acceptable as discussed in Note 2. Equipment qualified by test for a maximum pH of 10.5, which meets CS specified pH condition. See also Note 4. Analysis of test data using the Arrhenius methodolog.y demonstrates an Ql, QM, QT 8-day test at elevated temperature conditions is equivalent to 30 days operation at Containment post-LOCA conditions (see Note 5). l l Analysis of test data using the Arrhenius methodology demonstrates a l 100-hour acclerated aging test at 418'F is equivalent to 40 years l A This is in addt-I l operation at maximum normal Containment temperature. tion to the post-LOCA test noted above (see Note 5). Equipment, including insulation, qualified by test for specified l Containment environmental conditions. Lubricant has been determined 8 by analysis not to be susceptible to degradation f rom specified Containment environmental conditions. I Justification for continued Operation Pending Corrective Action l This equipment is considered qualified. I l l l
TABLE 3 TROJAN NUCLEAR PLANT ENVIROIMElffAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND CFDERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN Mtc SAFETY EVALUATION REPORT Sheet 36 of 43 Equipment Description Manufacturer Model No. Component No. Imcation Function Scotch 23 A Insulation for S-kV entor Tape 70 terminations NRC-Identified Deficiency Resolution of Deficiency QT, T. F. H this equipment is located in the Aus111ary Rut! ding and la mot subject to abnormal environmental conditions during post-accident operation, except for radiation esposure during long-ters recirculation. AJequat e operating time following an accident is demonstrated by enveloping the specified radiation requirement. Environments. qualification test data for this equipment was previously Ql, QN, m A unavailable. Recent information indicates that testing has acceptably qualified sletter equipment for aging and specified radiation requirement. W re detailed confirmatory evaluettoa is ongoing. Justification for Continued Operation Pending Corrective Action Reasonable aneurance estets th.at equipment is qualified. Wreover, teref eations are mechanically separated so that fatture of insulating material vould not necessarily result in motor fatture.
TABI.E 1 TROJAN NUCLEAR PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND GENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 37 of 43 Equipment Description Manufacturer Model No. Component No. Location Function Motor-Operated Velve Limitorque SMB-O HD 8801A&B P BIT outlet isolation 8802A&B P SIS hot leg isolation 8803A&R P BIT inlet isolation SMB-00 No 3290,3291 P CCW isolation to RCP&CAC 3292,3346 P CCW isolation from RCP&CAC ll2B&C P JCP isolation from VCT 8100 P BCP seal water isolation 8105,8106 P RCS clerging isolation 8110,8111 A CCP sintflow line isolation 8813,8814 A SIS 6intflow !!ne isolation 8821A&B P SIS cold leg isolation 88078 A SIS isolation from CCP SMB-000 Ho 2056A&B A Sodium hydroxide tank outlet isolation 3rc-fientified niency Resolution of Deficiency QT + *F? #s91A&B aus 8862A&B only) Analysis of test data using the Arrhenius methodology demonstrates that a 16-day test at elevated temperature conditions is equivalent to 6242 hours at the maximum Auxiliary But! ding or Penetration Area accident tem pe rat ure. This exceeds the 30-day (720-hour) operatina time requirement. Haterial analysis combined with analysis of test data using the Arrhenius A methodology demonstrates a 40-year qualified life (see Note 5). T P, OH (for Ho 88078 only) This equipment is located in the Auxiliary Building and is not subject to abnormal environmental conditions during post-accident operation, except for radiation exposure during long-term recirculation.
- (for HD 8807B only)
Operator brakes have not been qualified for specified radiation environ-ment. However, operator performs essential safety function beTore exposure to severe radiation environment. Justification for Continued Operation Pending Corrective Action These operators are considered qualified. L.
TABLE 3 TROJAN NUCLEAR PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND CENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NkC SAFETY EVAL 9ATION REPORT Sheet 38 of 43 Equipment Description Manufactures Hudel No. Component No. Location Function Motor-Operated Valve Limit orque SMB-000 HG 3210A&B A CCW isolation to RHR heat exchangers 5663,34/3 P Containment radiation monitor isolation 32 9BA, B,C,D C CCW isolation se RCP motor bearings 5675,5676, P Containment hydregen analysis 5677, 5678 isolation SMB-3 No 8835 P SIS cold leg isolation NRC-Identified De fic ienc y Resolution of Deficiency M QT (except for NO 3298A,B,C,ti Analysis of test data using the Arrhenius methodology demonstrates that a 16-day test at elevated temperature conditions is equivalent to 6242 houra at the maximum Auxiliary Butiding or Penetration Area accident temperat ure. This exceeds the 30-day (720-hour) operating time requirement. A (except for MO 32984,B,C.D) Heterial analysis combined with analysis of test data using the Arrhenius methodology demonstrates a 40-year qualified life (see Note 5). T P, R. M A, Ql, RPS (for MO 3298A,B,C D only) This equipment is not qualified for post-accident environmental conditions inside Containment. However, the equipment is exempt f rom qualification because it does not perform essential safety functions in the harsh environment (valves are normally open, and would f all in the open position, and are not required to be closed for post-accident shutdown or recovery). Justification for Continued Operation Pending Corrective Action These operators are considered qualified.
TABLF 1 TROJAM NUCLEAR PLANT t.NVIRNNNTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPME!G PORTLAND CENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Shee t 39 of 43 Equipment Description Manufacturer Model No. Component No. Location Func t ion Motor-Operated Valve Limitorque SMB-000 MO 5674 C Containment hydrogen analysis teolation SMB-2 MD 8702 C RHR from RCS teclation SMB-3 MO 2069A&B C Containment recirculation sump isolation NRC-Identified Deficiency Resolution of Deficiency QT T. P. H. A, QH, QI (for MO 2069A&B) As noted in earlier submittele to NRC, qualification of these operators P. CS, QH, Q1 (for MO 5674) was in question because uncertainty about installed component serial numbers did not allow identification of applicable test reports. The serial numbers were checked and corrected during the Spring 1981 cutage. Confirmation of valve type has been requested from the vendor. Qualification testing has been completed for each type. Therefore, final. resolution will be dependent on identification of the applicable test re po r t. T, CS, A. RPS, Q1 (for MO 8702) Identification of the correct eertal number has allowed determination t ha t the operator is qualified via the test report applicable to operators addressed on Sheet 43 (see latter for discussion of deficiencies). Justification for Continued Operation reading Corrective Action All other similarly applied operatore are appropriately qualified. Mureover, valves would fall in normal safe position (No 5674 closed and MO 2069A&B open). For fatture of MO $674 in its normal closed position (which le safe position for Containment teolation), a redundant hydrogen analysis flow path would be avalleble via a qualified operator (No 56' j.
TMM 3 I TWOJA3 NUCLEAR PLANT ~ ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND GENERAL ELECTRIC COMPANY'S i RESOLUTION OF IDENTIFIED DEF1CIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 40 of 4) Equipment De sc ript ion Manufacturer Model No. Component No. Location Func tion Motor-Operated Valve Limi to rque SMB-O MO 112D,112E P CCP supply from RWST teolation i 8806 A s15 supply from RWST isolation SMB-00 NO 8924,8807A A SIS isolation from CCP 892 3 A& B A SIS supply from RWST teolation 10005,10006 P Hydrogen vent supply isolation 10011,10012 P Hydrogen vent exhaust isolation SMB-1 MO 2053A&R P CSS Containment spray header isola t ion SMB-2 8716A&B A Rhk t rain isolation 8804A&B A RHR isolation to CCP 8809A&B P RHR cold leg tootation 2052A&B P CSS pump suction isolation from Containment recirculation sump SMB-3 MO 8812 A RHR supply f rom kWST 8700A&R isolation SMB-000 No 8112 C RCP seat water isolation NRC-Identified De fic ienc y Besolution of Deficiency QM, T. P, QT (except for MO 8112; no QT for MO 8806, 8807A, This equipment le located in the Aust11ery Butiding or Penetration Area 8923A&B,10005,10006 and 88121 no T for MO 8812) and to not subject to abnormal environmental conditione during post-accident operation, except for radiation esposure during long-term recirculation. Adequate operating time following an accident le demonstrated by enveloping the specified radiation requirement. CS (for MO 8812 only) This equipment le located in the Austilary Building and is not subject to a chemical spray environment during post-accident operation. A (escept for MD 8112) Material analysis combined with analyste of test data using the Arrhenius methodology demonstrates a 40 year quellfled life (see Note 5).
- (except for MO 10005,10006, 10011, 10012 & 8112)
Operator brakes have not been quantited for specified radiation environ-ment. Hauever, operatore perform essentist safety function before exposure to severe radiation environment.
- (for MO 10005, 10006, 20011 & 10012 only)
Operator brakes have not been qualified for speelf fed radiation environ-ment. Brake failure will cause valve to remain in pre-iatture position. m
TABLE 3 TROJAN NUC1JEA] P1 ANT ENVIRONNENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND GENERAL ELECTRIC COMFAh4'S ~ RESOLUTION OF IDEMilFIED DEFICIENCIES IN NRC SAFETY EVALUATION REPORT Sheet 40a of 43 NRC-Ident i fied Deficiency Resolution of Deficiency T. P, CS, R H. A Ql, RPS (for HO 8112 only) This equfsment is not qualified for post-accident environmental condi-tions inside Containment. However, the equipment is esempt from qualif t-cation because it performs its function before esposure to a harsh environment. Adequate time margin to perform safety-related f unction is assured by valve design and partial qualification in a steam and radia-tion environment (i.e., valve closes within 10 seconde upon receipt of an autreatic Containment isolation etsnal and motor operator is qualified by test to 212*F and 2.4 a 107 rad). Justification for Continued Operation Pending Corrective Action 4 These operators are considered qualified, except for MO 10005, 10006, 10011 & 10012. The hydrogen vent valves are partially open during normal o pe ra tion. Hewever, the equipment would perform its short-term safety function (Containment isolation) before exposure to a severe radiation environment since these valves are designed to close within 5 sec upon receipt of an automatic Containment isolation signal. Therefore, brake failure will not inhibit Containment isolatisn function. Redur. dant hydrogen control is available via qualified hydrogeo recombiners. 1
TABLE 3 TRajAD NUCLEAR FlJLNT ENVIR0fetENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT PORTLAND CENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICIEtJCIES IN NRC SAFETY EVALUATION REPORT Sheet 41 of 43 Equipment Description Manufacturer Model No. Component Nu. Location Function Motor-Operated Valve Limitorque SMB-00 H0 2050A&B P CSS supply from RWST isolation 8000A&B C Power-operated relief valve isolation SM8-000 MO 5672,5673 C Containment hydrogen analysis isolation 5651A,B,C.D Accumulator tanks sample isolation 5653,5654 Not leg sample isolation 5656 Pressurizer 11guld sample isolation 5658 Pressurizer vapor sample isolation 4005 RCDT isolation 5660 RCS hydrogen sample (f rom RCDT) isolation 4180 Containment sump discharge isolation 4300 Cas collection header isolation HD 2218,2228, M AFW pump turbine driver 2238,2248 steam supply isolation HO 10013,10016 C Con t a inment chilled water isolation NRC-Identified De fic iency Resolution of Deficiency A, QM (no QH for MO 5653) Material analysis combined with analysis of test data using the Arrhenius methodology demonstrates a 40-year qualified lif e (see Note 5). CS (except for MO 2050A&B) Equipment qualified by test for a maximum pH of 10.0, which does not envelope the maximum specified pH of 10.5. However, this is concidered acceptable as discussed in Note 4. F (except for HD 2050A&B, 2218, 2228, 2238 and 2248) Original specified requirement for pressure was not enveloped by test curve (i.e., 20 psig required vs 15 pais test at 11-to 24-hour time interval). However, test conditions do envelope FSAR Containment analysis (see Note 1). Moreover, qualification testing at higher pressures early in the test sequence (75 to 105 psis) demonstrates component has no significant failure modes due to pressure. Y, R. M, ql, RPS (for MO 5653 only) As noted in previous submittal to NRC, qualification of these operators uns in question because of discrepancy between recorded serial number and valve application inside Containment. Identification of correct serial number has allowed determination that the operator is qualified via the test re port app!! cable to operators addressed on this sheet. Justification for Continued Operation Pending Corrective Action These operators are considered qualif ted. u
TABLE 3 TROJAN NUCtEAR PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ElICTRICAL EQUIPHE,ff e PORTLAND CENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED DEFICIENCIES IN NRC SAFETY EVALU4 TION REPORT Sheet 42 of 41 Equipment Description Manufacturer Model No. Com ponent No. Location Function Motor-Operated Valve Limitorque SMB-00 CV 3004A1 M AFW flow control 3004A2 300481 3004 B2 3004C1 3004C2 3004DI 3004D2 NRC-Identified De f icienc y Resolution of Deficiency QT T. P, H. R A QH, RPS Environmental test data is unavailable for this equipment. Equipment is located in a room that does not contain a high-energy line and is not prone to submersion. However, the room communicates with the main steam and feeduster line area via a small overhead hatch opening and would be briefly exposed to steam at sabient pressure following an RELB. Due to the short duration of exposure and physical separation of redundant equipment, failure is not anticipated. Engineering studies are underway to determine if additional measures are warranted to farther protect existing equipment. Justification for Continted Operation rending Corrective Action Reasonable assurance estata that this equipment will perform its essential safety function. Local manual operation could be implemented for long-term flow control.
TABLE 3 1ROJAN NUCLEAR PiANT ENVIRONMEKTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT ~ PQRfLAND GENERAL ELECTRIC COMPANY'S RESOLUTION OF IDENTIFIED 1EFICIENCIES IN NkC SAFETY EVALUATION REPORT sheet 43 of 43 Equipment De sc ript ion Manuf ac t ure r Hudel No. Com ponen t No. Location Function Motor-Operated Valve Limitorque SMB-2 HQ 8703 P RhR hot leg isolation SM B-3 No 8811A&B P EllR pump suction isolation from Containment recirculation sump SMB-00 HQ 3294,3296,3300 C CCW isolation to/from RCP 3301 A& B,3302A& B, CCW isolation to/from CAC 3305 A& B,3 306A&B, 3309A&B,3310A&B 3313A&B,3314A&B 10007,10008 Ilydrogen vent supply isolation 10009,10010 Hydrogen vent exhaust isolation SMB-t *) Ho 3299 C CCW isolation to Excess Letdown test ex c ha nge r-3320 CCW isolation from RCP 3347,3293 CCW flow control from CAC SMB-2 NO 8701,8702 C RHR from RCS isolation SMB-3 MO h000) C Purge exhaust isolation 10002 Purge supply isolation SMB-4 NO 8808A,B,C D C Accumulat or tank outlet isolation NRC-Identified Deficiency Resolution of Deficiency QT (for MO 8811A&B, 10007, 10008, 10009, 10010, 8701, 8703) Analysis of test data using the Arrhenius methodology demonse. rates that M (for MO 3293, 8701, 870), 10002) a 7-day test at elevated temperature conditions is equivale at to at least QM (for MO 3347, 8808A,B.C.D 8111A&B) 30 days operat ion at Containment post-I SCA conditions (see Note 5). A Material analysis combined with analysis of test data using the Arrhenius methodology demonstrates a 40 year qualified life (see Note 5). P (for MO 8811A&B, 10002, 1000), 10007, 10008, 10009 Test pressure envelopes the first 6 days of both the original specified 10010, 3293, 3347, A808A,B C.D only). Containment pressure and FSAR Containment pressure profiles (see Note 1). u
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IABLE 3 ABBREVIATIONS Abbreviation System or Analysis Designations AFWS Auxiliary Feedwater System BIT Boron Injection Tank CAC Containment Air Cooler CCP Centrifugal Charging Pump CCWS Component Cooling Water System CSS Containment Spray System ECCS Emergency Core Cooling System ESFAS Engineered Safety Features Actuation System HELB High-Energy Line Break HVAC Heating, Ventilating and Air Conditioning LOCA Loss-of-Coolant Accident MSLB Main Steam Line Break RCDT Reactor Coolant Drain Tank RCP Reactor Coolant Pump RCS Reactor Coolant System RdRS Residual Heat Removal System RPS Reactor Protection System RWST Refueling Water Storage Tank SIS Safety Injection System VCT Volume Control Tank Plant Location Designations A Auxiliary Building C Containment M Main Steam Support Structure (also MSSS) P Containment Penetration Area NRC Deficiency Designations A Material Aging Evaluation, Replacement Schedule, Ongoing Equipment Surveillance CS Chemical Spray H Humidity M Margin P Pressure i QI Qualification Information Being Developed QM Qualification Method QT Qualification Time R Radiation RPN Equipment Relocation or Replacement, Adequate Schedule not Provided RPS Equipment Relocation or Replacement Schedule Provided S Submergence T Temperature
- : Foxboro letter (3/12/81), " Potential Deficiency Af fecting Foxboro Transmitters".
Replace lub-icant and insulation with qualified ones by June 1982. Replace brakes with Limitorque SB conversion kit by June 1982. DRS/sh 4-66.62B27 ]
) i NOTE 1 TO TABLE 3 Sheet 1 of 2 DISCUSSION OF TROJAN EQUIPMENT QUALIFICATION WITH RESPECT TO TEMPERATURE AND PRESSURE CONDITIONS INSIDE CONTAINMENT Heretofore, in reviewing the environmental qualification of safety-related l electrical equipment at Trojan, specified pressure and temperature conditions for equipment inside Containment were based on Figures 3.11-1 and 3.11-2 of the FSAR, respectively (maximum pressure and temperature of 60 psig and 286*F, respectively). These curves, however, are test envelopes defined by the original Westinghouse qualification test program in WCAP-7744. The environmental conditions for which equiiment inside Containment should i be qualified is better represented by the Containment transient analyses described in Section 6.2 of the FSAR. Figures 6.2-62 and 6.2-76 of the FSAR (attached) more appropriately define the worst-case LOCA environment to be specified for qualification of safety-related electrical equipment inside Containment (maximum pressure and temperature of 59.3 psig and 288'F, respectively). This approach is consistent with Section 4.1 of the DOR Guidelines, which recommends that Containment temperature and pressure conditions as a function of time be based on the analyses in the FSAR. The Trojan FSAR analysis, which uses the Bechtel C0FATTA computer program, a derivative of the CONIEMPT code, is also consistent with Section 1.1 of NUREG-0588 and has been previously accepted by the NRC.(1) With respect to the environmental conditions resulting from an MSLB inside Containment, Section 4.2.1 of the DOR Guidelines states that " equipment qualified for a LOCA environment is considered qualified for an MSLB accident environment in plants with automatic spray systems not subject to disabling single component failures. This position is based on the 'Best Estimate' calculation of a typical plant peak temperature and pressure and a thermal analysis of typical components inside Containment.(2) The final accept-ability of this approach, i.e., use of the 'Best Estimate', is pending the completion of Task Action Plan A-21, " Main Steam Line Break Inside Containment". Because Trojan is equipped with an automatic Containment spray system not subject to a disabling single component failure, the LOCA environmental con-dicions discussed above are considered to effectively bound the MSLB condi-tions. The Trojan Containment Spray System is described in detail in Section 6.4 of the FSAR. The analysis provided in Table 6.4-2 of the FSAR provides sufficient verification that the Containment Spray System is not subject to disabling mingle component failures. i l It is recognized that final acceptability of the approach stated in Sec-tion 4.2.1 of the DOR Guidelines (fe, use of the Best Estimate) depends upon 5 (1) Saf ety Evaluation Report Trojan Nuclear IIant, Docket No. 50-344, U. S. l Atomic Energy Commission, October 1974. l (2) See NUREG-0458, Short-Term Safety Assessment on the Environmental Quali-fication of Safety-Related Electrical Equipment of SEP Operating Reactors, for a more detailed discussion of the best estimate calculation. I
NOTE 1 TO TABLE 3 Sheet 2 of 2 the resolution of Task Action Plan A-21. However, other analyses conducted for PGE have evaluated the thermal capabilit of typical equipment inside Containment exposed to an MSLB environment (3{>.This analysis shows that a superheat condition (with corresponding high Containment atmosphere temperature) should have no significant effect on electrical equipment temperatures because the expected equipment surface temperature would follow the Containment saturation temperature, which is substantially lower than the peak vapor temperature during the superheat phase of the accident. The reason for this is that energy transfer from the Containment atmosphere to heat sinks is significant only when the sink surface is cooler than the saturation temperature so that condensation can occur. If the equipment surface temperature were to become higher than saturation, then the low energy transfer mechanism of convection would govern heat transfer. Since the Containment peak pressure is at a ma. imum following the design basis LOCA, the Containment saturation temperaturc for an MSLB accident is no higher than would be the case for a LOCA. It is on these bases that the worst-case environment for Containment equipment will result from the design basis LOCA. Therefore, the position as stated above (Section 4.2.1 of the DOR Guidelines) should remain a valid conclusion for the Trojan design. With regard to temperature stratification affects following a LOCA or MSLB, any such eff ects will be short-lived, ie, less than several ninutes, due to Containment Spray System operation. Moreover, any stratification effects will be ccnfined to equipment located in the highest elevations inside Concainment. Based on the above discussion, it is PGE's position that the previously used pressure and temperature profiles (FSAR Figures 3.11-1 and 3.11-2) represent a con:ervative basis for evaluation of the environmental qualification of safety-related electrical equipment inside Containment for both a postulated LOCA or MSLB event. (3) R. W. Braddy, H. M. Scheenhoff, J. W. Thiesing, " Surface Temperature Response of Equipment Inside Containment Following Pipe Break", Trans-action s of the American Nuclear Society, 24, (1976). DRS/cw 4-66.62B11 ]
m l, 1 A I IO a 4 6 I I I t il 6 i e iIlli i i i a iIll i i i i iill l l t i ligt 58.3 psig -(116 sec) SO o / R l Gouge Pressure //
- t. t f.,
M .F Steam Pressure E .L' / \\, h. a-l \\ ,3 / f-* Sump Wster Recirculation ./ .M..I (1500 sec) l l 20 l l 10 .g eI ' I l lill t t 1 811't I I I t i ll' I ' ' l 111' 100 tot in2 ig3 ig4 ins Time, sec Figure 6.2-62 Containment Pressure versus Time - DEPS Guillotine Break Maximum Engineered Safety Features Amendment 9 (February 1974)
1 340 3288F (52 sc) 300 [ / epor Region V f - 252 8F C*, sec.) 260 g g l-- Sump Water Recirculation i 1 n500 .) i / ump Region I 7." S 5 220 Ee I .s i E I \\ \\ l l 140 \\ W l 't' f f f ! f !f 100 100 yg1 102 103 104 105 Time After Rupture, see l l Figure 6.2 - 76 Containment Temperature Response for the DBA Amendment 9 l (February 1974) .]
NOTE 2 TO TABLE 3 Sheet 1 of 1 DISCUSSION OF TROJAN EQUIPMENT QUALIFICATION WITH RESPECT TO CONTAINMENI HUMIDITY CONDITIONS I In some cases, electrical equipment qualification testing was not conducted in a manner that assured attainment of 100 percent relative humidity nor in a manner that allowed calculation of actual test humidity conditions. However, the testing did expose com>osents to saturated steam conditions for an extended period of time. This is judged to provide rea: nable assurance of equipment operability under high humidity conditi .s because: 1. Continued injection of steam into a closed test chamber eventually resulted in an extremely high humidity conditin - 2. Injection of steam into a closed test chamber is analogous to high-energy water / steam pipe ruptures inside Containment. 3. Sufficient moisture exists in the high-temperature saturated steam conditions to identify significant component failure modes that might result from condensatjon or water absorption. DRS/cw 4-66.62B13
NOTE 3 TO TABLE 3 Sheet 1 of 1 DISCUSSION OF TROJAN EQUIPMENI QUALIFICATION WITH RESPECT TO SUBMERGENCE The maximum water level inside Containment resulting f rom a worst-case large break LOCA event is 52 f t, 11.5 in. as indicated in the master list key supplied with previous PGE submittals. The bases used for determining this level are discussed in Section 15.4.1.6 of the FSAR. All equipment required for post-LOCA/MSLB long-term operation has either been relocated above flood level or replaced by equipment qualified for submergnnce, or is scheduled for replacement by equipment qualified for submergence. Equipment required only for short-term post-LOCA/MSLB operation is automatically actuated by an ESFAS signal. Since the flood level calcu-lation assumes the RWST, sodium hydroxide tank and all accumulator tanks are emptied into Containment, ESFAS actuation is prerequisite to signifi-cant Containment flooding. Therefore, the safety function of equipment in this category will be completed before flooding occurs. Flooding of this equipment after completion of its safety function will not result in unacceptable failure of the flooded equipment, nor will it result in failure of other safety-related equipment that must be operable. The submergence of safety-related electrical equipment outside of Contain-ment was not specifically addressed in previous submittals to IE Bulle-tin 79-OlB. This was not because floodir.g of cafety-related equipment outside Containment was never considered, but because no safety-1 related electrical equipment outside Containment has been identified which is prone to submergence following a high-energy line break. Topical Report PGE-1004, " Trojan Nuclear Plant Analyses of Pipe System Breaks Outside Containment", describes in detail the analyses performed, and specific features incorporated into Plant design, to assure that no safety-related equipment, electrical or otherwise, would be affected by flooding as a result of the high-energy line break outside Containment. DRS/cv 4-66.62B14
e NOTE 4 TO TABLE 3 Sheet 1 of 2 DISCUSSION 07 TROJAN EQUIPMENT QUALIFICATION WITH RESPECT TO CHEMICAL SPRAY The worst-case chemical spray conditions used for environmental qualifi-cation of electrical equipment are boric acid, pH 4.2, and sodium hydroxide, pH 10.5. These bounds were established to envelope conditions resulting from either a reactor coolant system pipe rupture or Contain-ment spray. An acidic condition would result only from a ruptured reactor coolant system pipe during and immediately af ter boron injection by the centri-fugal charging pumps. In this case, some degree of protection against the acidic spray environment is afforded by the physical separation of the RCS loops. For equipment within the scope of the IE Bulletin 79-01B review, the followiag have not been tested under boric acid conditions 4 but could be subjected to an acidic pH environment: 1. Terminal blocks for RC3 hot and cold leg RTDs 2. Pressurizer safety valve position indication i l 3. Pressurizer relief valve solenoids and position indication 4. Pressurizer block valves and position indication. The RTD terminal blocks are located near their respective reactor coolant pumps. Failure of a component associated with the ruptured RCS loop will not adversely affect the ability of the Plant to respond to its design basis accident. Equipment associated with the intact loops would be afforded protection due to the separation between loops; therefore, canponents are not expected to experience a strong acid environment. l Additionally, the equipment in question is housed in metal enclosures j that protect the vital parts from direct effects of the acidic spray. i l The remaining items are located in the pressurizer enclosure. They all function to relieve the pressurizer to the pressurizer relief tank or l to detect that relief is taking place. The pressurizer safety, relief and block valves are all normally closed and are not relied upon to mitigate the effects of a large break loss-of-coolant accident. Qualifi-l cation of Items 2 through 4 for acid spray, therefore, is not required. I Equipment outside these areas should only be subjected to the basic Containment spray with pH between 8 to 10.5 as discussed in Sec-tion 6.4.3 of the FSAR. Equipment in this category has either been tested in a basic spray environment or has been analyzed for the spray effects. In some cases, spray testing was not done at a pH level as severe as the most basic Containment spray possible. Neve r-theless, this equipment is considered to be qualified for its applica-tion since materials known to be reactive with the chemical spray have not been used in Contaicment, and no failures resulted from the 4 w-+, .m. y- -. ,,._,,_,,_.,_p .,__,,-_..,-._,,.y._,
- mm_,
,%,,,,,y.
NOTE 4 TO TABLE 3 Sheet 2 of 2 exposure of the equipment to the very high pH (usually pH 8 or above) conditions that these items were subjected to in their respective test programs. Moreover, Containment spray with a pH of 10.5 is a worst-case condition for a short time duration (ie, less than 30 minutes), based on bounding values of parameters (see FSAR Figure 6.4-7). Actual pH values would be in the 9.0 to 10.0 range. During the long-term recircu-lation phace, which is more representative of the environmental conditions equipment should be qualified to, pH is maintained betweeg 8.0 to 9.5 (FSAR Section 6.4.3). NCD/cw 4-66.62B15 ]
NOTE 5 TO TABLE 3 Sheet 1 of 4 DISCUSSION OF TROJAN EQUIPMENT QUALIFICATION WITH RESPECT TO AGING Basis and Approach The DOR Guidelines and NUREG-0588 require censideration of aging effects on safety-related electrical equipment, but by different methods. As stated in Sections 3.7 and 4.3 of the SER, Section 7 of the DOR Guidelines considers the following to provide an acceptable basis for addressing equipment aging if c specific qualified life cannot be demon-strated as required by IEEE Standard 323-1974 (Category 3 of NUREG-0588): 1. Conduct an equipment material evaluation to ensure that no known materials susceptible to degradation becaase of aging have been used, ~ i 2. Establish an ongoing program to review ' plant surveillance and maintenance records in order to identify equipment degradation which may be age related, and/or 3. Propose a uaintenance program and replacement schedule for equipment identified in Item 1 or equipment that is qualified for less than the life of the plant. In evaluating the qualification of most safety-related electrical equip-ment at Trojan, the foregoing DOR Guideline method was employed only when a specified qualified life could not be demonstrated for equipment by an accelerated aging test or excess margin during the accident simulation. Because NUREG-0588 considers the Arrhenius methodology an acceptable method of addressing accelerated aging, this method was used to establish a qualified life for equipment with available pre-aging test data. The Arrhenius method was also used to determine an " equivalent" qualified life when eveeeaive test margin existed between the test and requirement temperature profiles. This proves to be a very conservative approach I since Figure 3.11-2 of the FSAR has been used in these evaluations as the required temperature profile and represents a much greater potential for thermal degradation than the more realistic (and appropriate) l Figure 6.2-76 profile, which is based on Plant-specific analyses (see Note 1). In cases where insufficient accelerated aging was applied as l part of an equipment test program, material analysis has been conducted l in accordance with Item 1 above. For example, aging effects for Limitorque motor operators were evaluated l by a combined approach using component material analysis and the Arrhenius methodology. Each component was evaluated as to its potential for i thermal degradation, and if found to be susceptible, its qualified life j was determined by Arrhenius calculations. Insulation, switches, seals, terminal blocks, and lubricants were included in the material evaluation. The terminal blocks used in the motor operators are an example of qualifi-l cation by material analysis. These blocks are molded phenolic, which has no potential for significant thermal degradation over a 40-year period. )
NOTE 5 TO TABLE 3 Sheet 2 cf 4 On the other hand, an RH insulated motor is an example of an ites suscep-tible to thermal degradation. Based on analysis of test data using the Arrhenius methodology, Limitorque gives the life of an RH insulated motor as 9 x 109 hours at 120*F. This is much greater than the required 40-year life (3.5 x 106 hours). Application of Arrhenius Methodology Aging theories are based on material degradation due to stress factors. The Arrhenius model simulates accelerated thermal aging by increasing the reaction rate of certain stress factors. Enduring an environment of high stress reaction rates for short periods of time is equivalent to enduring an environment of lower stresses for longer periods of time. The Arrhenius model also relates rate of reaction to temperature by the function: r = A exp(-4/kT) where r = rate at which the degradation reaction proceeds A = constant for the material (frequency factor) 9 = activation energy (eV) k = Boltzman's constant (0.8617 x 10~4eV/*K) T = absolute temperature,
- K.
I J
NOTE 5 TO TABLE 3 Sheet 3 of 4 The following relationship is used to determine equivalent aging: 9 (l) In (t,/t,) = \\*
- )
where e, = simulated or service age T, = operating service temperature t,= accelerated aging time required T,= accelerated temperature. To illustrate by example, consider the following case involving the Limitorque operators listed in Appendix B of the SER. The test conditions, although shorter in duration than the specified accident conditions, can be shown to be more severe. Using the Arrhenius method, the thermal degradation resulting from testing and accident conditions can be compared. The operating time requirement in an accident environment is 30 days (720 hours) at 104*F (313*K). A conservative representation of the test profile is: 1. 0.5 hours at 250*F (394*K) 2. 1.4 hours at 120*F (322*K) 3. 24 hours at 250*F (394*K) 4. 350 hours at 200*F (366*K) Using a conservative activation energy * = .5, the above time and tempera-ture profile can be transformed to a second time and temperature profile, where temperature is confined at 104*F-(313*K), that represents equivalent thermal degradation. Manipulating Equation (1) yields: e / T-1 i l tequiv* " Ca eXP Y 'g equiv ~T T aj_ I 1)~ 1 => cequiv " C exp 5804. g 394 i a i-This yields the following equivalent aging times: t,qi = 22.6 hours at 104*F t,q2 = 2.4 hours at 104*F t,q3 = 1086.0 hours at 104*F t,q4 = 5131.4 hours at 104*F i Total = 6242.4 hours at 104*F l
NOTE 5 TO TABLE 3 Sheet 4 of 4 Therefore, the test was equivalent to 6242 hours at the maximum specified post-accident temperature of 104*F. This far exceeds the required 720 hours at 104*F, and thus the 30-day operating requirement is con-sidered satisfied. Similar calculations are used to determine qualified plant life of equip-ment, either on the basis of excess test margin as the above illustrates, or from available test data for accelerated aging. Maintenance and Surveillance Program Considerations The above discussed techniques have been used to establish periodic maintenance or replacement intervals for equipment that has a calculated qualified life of less than 40 years. Further analysis is underway to establish qualified lives for equipment where aging qualification is presently indeterminate. These analyses will also form the basis of periodic maintenance or replacement schedules. In view of the inherent uncertainties associated with material analysis or extensive extrapolation of aging test data, PGE concurs with the need to systematically evaluate cperating experience in order to further validate existing aging qualification. There exist, however, a broad spectrum of approaches to performing this evaluation. Unfortunately, the accuracy and completeness of the results of such programs as are under consideration for Trojan appear to be directly proportional to the dif ficulty of implementation. Therefore, PGE is participating in an industry-wide examination of this issue through the EPRI Equipment Qualification Advisory Group in order to strike an appropriate balance. In the meantime, equipment on the 79-OlB master list is generally subject to periodic functional testing and calibration. Since Trojan is a relatively new plant and most of the currently installed equipment is of approximately the same age, grouped failures of similar equipment would likely be identified by this program even in the absence of a specific tracking program. l l l l DRS/cw 4-66.62B17 \\
TABLE 4 TROJAN NUCLEAR PLANT CALCULATED TOTAL INTEGRATED DOSES
- IN RADS (GAMMA AIS BETA) 1st Hr 1st Day 1st Wk ist Mo ist Yr LOCA y, Inside Containment 1.4E6 7.lE6 2.0E7 2.7E7 3.3E7 Below Elev. 93' y, Inside Containment 1.7E6 5.5E6 1.0E7 1.3E7 1.3E7 Above Elav.93' 8 Air, Inside Containment 8.lE6 4.6E7 9.6E7 1.4E8 1.4E8 (Large Volume) y, Outside Containment 1.6ES 1.6ES 6.2E6 8.4E6 1.lE7 Areas Containing RCS Liquids Dose to Components in Contact with Reactor Coolant Contact 8 2.7ES 2.7E6 1.0E7 1.3E7 2.lE7 Contact y 3.2E5 3.2E6 1.2E7
- 1. 7E'/
2.2E7 Intact RCS y, Inside Containment 7.7ES 4.7E6 1.2E7 1.7E7 2.3E7 All Areas Below Elev. 93' y, Outside Containment 1.4E6 8.2E6 2.1E7 2.9E7 4.0E7 Areas Containing RCS Liquids Dose to Components in Contact with Reactor Coolant Contact S 3.4E6 2.0E7 5.3E7 6.8E7 1.lE8 i Contact Y 2.7E6 1.6E7 4.2E7 5.8E7 8.0E7 [a] These TIDs are maximums for the areas specified. Location-specific evaluations if performed may provide lower TIDs for that location. JLT/mm 4-66.62B26 i
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e e TABLE 5 11t0JAN NUCLEAR PLANT ENVIRONMENTAL QUALIFICATION INFORMATION ROSEMOUNT PRESSURE TRANSHITTERS H0 DEL 1153 SERIES A (AUXILIARY FEEIWATER FLOW TRANSHITTERS) Environment: Documentation: Qualification: Parameter Specification Qualification Specification Qualification Method Operating Time <1 hr 64 hr PCE 1004 (8/75) Rosemount Report No. 3788 Simultaneous Test Temperature (F) 225 Figure 3 PCE 1004 (8/75) Rosemount Report No. 3788 Simultaneous Test Pressure (paig) 4.3 Figure 3 PCE 1004 (8/75) Rosemount Report No. 3788 Simultaneous Test Re1 Humidity 100% 100% PCE 1004 (8/75) Rosemount Report No. 3788 Simultaneous Test Chemical Spray None pH 10.5 Rosemount Report No. 3788 Simultaneous Test HB03 + NADH Radiation Normal 4.4 x 107 Rads Rosemount Report No. 3788 Simultaneous Test Aging 40 yrs FSAR Table 5.1-1 Analysis Submergence None N/A Location N/A Analysis DRS/jr 4-66.62B23 a-
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