ML20138J891
| ML20138J891 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 12/12/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20138J888 | List: |
| References | |
| NUDOCS 8512180109 | |
| Download: ML20138J891 (14) | |
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!pnt0 'o g UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION g E WASHINGTON, D. C. 20555 SAFETY EVALUATION PY THE OFFICE OF NUCLEAR REACTOR REGULATION RE: ENVIRONMENTAL Ol'ALIFICATION OF ELECTRIC EQUIPMENT IMPORTANT TO SAFETY YANKEE NUCLEAR POWER STATION 00CKET NO. 50-29
1.0 INTRODUCTION
Equipment which is used to perform a necessary safety function trust be demonstrated te te capable of maintaining functional operability under all
- service conditions postulated to occur during its installed life for the time it is required to operate. This requirement, which is embodied in General Design Criteria (GDC) I and 4 of Appendix A and Sections III, XI, and XVII of Appendix B to 10 CFR Part 50, is applicable to equipment located inside as well as outside containment. More detailed reouirements and guidance relating to the methods and. procedures for demonstrating this capability for electrical equipment have been set forth in 10 CFR 50.49, " Environmental Qualification of Electric Equipment Irrportant to Safety for Nuclear Power Plants," NUREG-0588, "Interin Staff Position on Environmental Qualification of Safety-Related Electrical Equipment" (which supplements IEEE Standard 323 and various NRC Regulatory Guides and industry standards), and " Guidelines for Evaluating Environmental Outlification of Class IE Electrical Equipment in Operating Reactors" (D0R Guidelines).
2.0 BACKGROUND
On February _8, 1979, the NRC Office of Inspection and Enforcement (IE) issued to all licensees of cperating plants (except those included in the systematic evaluation program (SEP)) IE Bulletin (IEB) 79-01, " Environmental Qualification of Class 1E Equipment." This Fulletin, together with IE Circular 78-0F.
(issued on May 31,1978), required the licensees to perform reviews to assess the adequacy of their envircrnental qualification programs.
On January 14, 1980, NRC issued IEB 79-01B which included the 00R Guidelines and NUREG-0588 as attachments 4 and 5, respectively. Subsequently, on May 23, 1980, Comission Memorandur and Order CLI-80-21 was issued and stated that the D0R Guidelines and portions of NUREG-0588 form the reovirements that licensees must meet regardino environmental qualification of safety-related electrical equipment in order to satisfy those aspects of 10 CFR Part 50, Appendix A, 000 4 Supplements to IEB 79-01B were issued for further clarification erd definition of the staff's needs. These supplements were issued on February 29, September 30, and October 24, 1980.
8512180109 851212 PDR ADOCK 05000029 P PDR
In addition, the staff issued orders dated August 29,1980 (amerrded in September 1980)andOctober 24, 1980 to all licensees. The August order required that the licensees provide a report, by November 1, 1980, documenting the qualification of safety-related electrical equipment. The October order required the establishment of a central file location for the maintenance of all equipment qualification records. The central file was mandated to be established by December 1, 1980. The staff subsequently issued a Safety Evaluation Report (SER) on environmental qualification of safety-related electrical equipment to Yankee Atomic Electric Company (the licensee) on April 20, 1981. This SER directed the licensee to "either provide document-ation of the missing qualification information which demonstrates that safety-related equipment meets the 00R Guidelines or NUREG-0588 requirements or commit to a corrective action (requalification, replacement, etc.)." The licensee was required to respond to NRC within 90 days of receipt of the SER,.
In response to the staff SER issued in 1981, the licensee submitted additional information regarding the qualification of safety-related electrical equipment.
This information was evaluated for the staff by Franklin Research Center (FRC) in order to: 1)' identify all cases where the licensee's response did not resolve the significant qualification issues, 2) evaluate the licensee's qualification documentation in accordance with established criteria to determine which equipment had adequate documentation and which did not, and 3) evaluate the licensee's qualification documentation for safety-related electrical equipment located in harsh environments required-for TMI Lessons Learned Implementation. A Technical Evaluation Report (TER) was issued by FRC on May 28, 1982. A Safety Evaluation Report was subsequently issued to Yankee Atomic Electric Company on December 16, 1982 with the FRC TER as an attachment.
A final rule on environmental qualification of electric equipment important to safety for nuclear power plants became effective on February 22, 1983.
This rule, Section 50.49 of 10 CFR Part 50, specifies the requirements to be met for demonstrating the environmental qualification of electrical equipment important to safety located in a harsh environment. In accordance with this rule, equipment for Yankee may be qualified to the criteria specified in either the D0R Guidelines or NUREG-0588, except for replacement equipment.
Replacement equipment installed subsequent to February 22, 1983 must be qualified in accordance with the provisions of 10 CFR 50.49, using the guidance of Regulatory Guide 1.89, unless-there are sound reasons to the contrary.
A meeting was held with each licensee of plants for which a TER had been prepared for the staff by FRC in order to discuss all remaining open issues regarding environmental qualification, including acceptability of the environ-mental conditions for equipment qualification purposes, if this issue had not yet been resolved. On January 17, 1984, a meeting was held to discuss Yankee's proposed method to resolve the environmental qualification deficiencies identified in the December 16, 1982 SER and May 28, 1982 FRC TER. Discussions also included Yankee Atomic Electric Company's general methodology for compliance with 10 CFR 50.49, and justification for continued operation for those equipment items for which environmental qualification is not yet completed. The minutes of the meeting and proposed method of resolution for each of the environmental qualification. deficiencies are documented in the March 5, April 5, and July 16, 1984 submittals from the licensee.
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3.0 EVALUATION -
The evaluation of the. acceptability of the licensee's electrical equipment environmental qualificatior rrcgram is based on the results of an audit review performed by the staff of: (1) the licensce's proposed resolutions of the environmental qualification deficiercies identified in the December 16, 1982
-SER and May 28, 1982 FRC TER; (E) ccrpliance with the requirements of 10 CFR 50.49; and (3) justification for continued operatier (JCO) for those equipment items for which the environmental qualification is not yet completed.
3.1 Proposed Resolutfers of Identified Deficiencies The proposed resolutions for the equipment envircrmental qualification deficiencies, identified in the December 16, 1982 SER, and the FPC TER ,
enclosed with it, are described in the licensee's March 5, April 5, and July 16, 1984 submittals. During the Jaruary 17, 1984 meeting with the licensee, the staff discussed the proposed resolution of each deficiency for each ccuipment item identified in the FRC TER and found the licensee's approach for resolving the identified envircreental qualification deficiencies acceptable.
The majority of deficiencies identified were documentation, similarity, aging, qualified life and replacement schedule. All open items identified in the SEP dated December 16,-1982 were also discussed and the resolution of these items has,been found acceptable by the staff with the exception of the inside and outside containment (vaper container) pressure / temperature service conditions.
3.1.1 Pressure / Temperature Profiles Inside Certainment In its_ April 5,1984 letter, the licensee submitted an analysis of the corteirmert temperature response following a LOCA. This analysis is a reconstruction of the worst case LOCA analysis presented in the Integrated Plert Sefety Assessment Report (IPSAF) and was done to show the impact of usirg rcre realistic assumptions on the containr:ent temperature response.
As a result of our review, we noted that the method of calculating the effective cutside heat transfer coefficient (CHTC) was not sufficiently conservative. The licensee submitted a reanalysis in its December 7, 1984 letter taking into account our concern.
The licensee used the CONTEMPT-LT/026 computer code for the analyses presented in its May 5 and December 7,198a letters. By matching the reselts of a calculation ir the IPSAR the licensee denonstrated the acceptability of usirg CONTEMPT-LT/026,-even though it is a slightly different versior of the CONTEPPT-LT code used by the staff, which is CONTEMPT-LT/028. The IPSAF provides_the staff's Safety Evaluation Peport on SEP Topics VI-2.D and VI-3.
The containment temperature profile calculated by the staff is presented in Ficure 1 of the licensee's December 7,1984 letter (attached to this evaluation), and is labeled "LLL".
The licensee used the mass and eneroy release data from the IPSAP and modeled condensing heat transfer to pessive heat sinks inside containment in accordance with the guidance in NUREC-0588. We find the licensee's approach acceptable.
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The major differences between the licensee's analyses for enviror,rcertzi cualification and that presented ir. the IPSAP are in the internal represer,tation of hect sirks and the value of OHTC used to calculate heat loss to the atmcsphere. The licensee preserted a more detailed listing of the internal
. heat sinks that are actually available in the ccntainment. They include concrete walls, ma.ior structural steel, deck plate and gratirg, reactor cavity liner steel, and crane and support columns. Based on our review of this information, we cercur with the licensee that the heat sink listing is appropriate for use in certainment analysis.
The Yankee containment does not have activc heat removal systems, such as containment sprays and fan coolers, to remove energy from the containtrert.
Heat transfer that occurs across the containment boundary to the atmosphere is the only means available for post-LOCA erercy renoval from the containmer}t.
The licensee has shcwn that radiant heat transfer from the containment structure is an inportant factor in assessing the centcinment temperature
. response. In the IPSAP, radiant heat transfer was neglected for conservatism; however, its contribution to overall heat transfer from the centainment structure can be significant. In its April 5,1984 letter, the licensee o calculated an effective outside heat transfer coefficient of 3.2 Bte/br-ft' *F for an assumed surface temperature of 180*F. In our review, we found that the mettedelogy used to calculate radiant heat tiansfer was not sufficiently conservative. Therefore, the licensee revised its calculation of radiant heat transfer to include radiation emitted from the containtrent surface, direct solar radiation, indirect solar radiation, ard terrestrial radiatinn from the surroundings. From a parametric study involving a range of containment surfect temperature, outside air terrperature and view factor for teprestrial radiation, the licensee calculated an effective OHTC of 2.08 Btu /hr-ft 'F. Using this OHTC value. alerg with the 1979 ANS decay heat rates without the 2-sigma uncertainty, the revised heat sinks in its April 5, 1984 letter, the ress and encrcy release data in the IPSAP, the computer code CONTEMPT-LT/026, the licensee calculated a containment Case 1. The licensee proposes to use this temperature profile for equipment qualification, k'e have evaluated the licensee's overall containment analysis, including assurptinns, rethodology, and input data. Based on our evaluation of the infornation presented in the licensee's April 5 and December 7,19Pa lettcrs, we find the proposed temperature profile acceptable for use in equiprert qualificatier.
3.1.? Pressure / Ten perature Profiles Outside Conttirrert
'The staff has reviewed the licensee's subnittals relating to environmental cualification of ecuiprent cutside of containment. The licensee has evaleeted the effect er.ecuipment of breaks outside of containment in the following systems:
- 1. Charging / letdown system, P. Pain feedwater,
- 3. Pain steam system,.
4.- Steam generator blowdcwn system, and
- 5. Stean hcatire system.
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The plant is required to demonstrate environmental qualification' of electrical
-equipment _ required to attain and maintain hot shutdown. A discussion of the staff's review and conclusions regarding those results are provided below:
3.1.2.1 Charging / Letdown Line Breaks The licensee reported that the charging line contains fluid at a nominal operating temperature of 120 F; the charging flow rate is normally 20 to 30 gpm. The charging pumps are positive displacement pumps so that flow would increase only slightly in-the event of.a charging line break. Further, the licensee stated that the temperature of the fluid in the letdown line is 150'F with flow equivalent to that of the charging pumps.
The licensee stated that a charging line break could adversely affect only -
the hot leg injection valves and the three charging pumps. Therefore, one of the three low pressure and high pressure safety injection (LPSI and HPSI)
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pump trains, and one of the two motor driven emergency feedwater (EFW) pumps, in other areas, could be used to shut down the plant. While proceeding to shutdown, one of the eight full-capacity groups of pressurizer heaters could be connected to the emergency power system in order to provide for control of pressure in the primary system in the event all three charging pumps are disabled. Once the system pressure is reduced from the operating pressure of 2000 psig to below approximately 1560 psig, the HPSI/LPSI ECCS trains can begin to inject water into the primary system and can continue this throughout the cooling process until hot shutdown is achieved.
Therefore, single active failure of a safety injection train, or of a motor-driven EFW pump train or of a full capacity group of pressurizer heaters will not affect the ability of the plant to achieve hot shutdown. We find this to be acceptable.
3.1.2.2 Main Feedwater System Breaks A main feedwater.line break outside of the turbine building and not within another building;will not produce a harsh environment nor affect equipment required to mitigate this accident.
A main feedwater line break inside the turbine building (TB) may affect EFW flow indication equipment. However, flow can.be determined, indirectly, by observation and maintenance of level in the steam generators. Further, a main feedwater line break, either inside or outside of the TB may be mitigated by means of one of the two motor-driven EFW pumps which are located in a separate area, the primary auxiliary building (PAB). The feedwater provided is taken from the demineralized water storage tank with an alternate supply from the primary water storage tank. The EFW may be supplied via the normal feedlines, or alternatively, via the steam generator blowdown lines. The charging pumps may be connected to emergency power by remote manual operation of
circuit breakers. Therefore, even with a sirg!r active failure at least two of the three charging purp trains may be employed to maintain the levei of the water in the primary system. The FFSI/LPSI pump trains may
_ be used for this purpose once tha primary system pressure has been reduced from the operating value (2000 psig) to a value belcw 1560 psig. Therefore, a single active failure, corsisting either of a loss of a charging pump train or safety in,iection pump train or EFP pump train will not affect the ability of the plant to achieve hot shutdown. We find this to be acceptable.
3.1.2.3 Main Steam Lire Preaks A steam line break has the same results es a feedwater line break and is
.. mitigated in similar fashion. We find this to be acceptable. .
3.1.2.4 Steam Generator Blowdown Line / Breaks A steam generator blowdown line break will result in slow blowdown of one steam generator into the ncrsbielded portion of the PAB. The resultant harsh environment may adversely affect the two motor-driven EFW pump trains and two of the three charpino pumps, leaving one charging' pump, the three.HPSI/LPSI trains and the emergency boiler feed pump (EBFP) available to sFut the plant dewr. ' A sinole failure may eliminate either the remaining charging purp, the EBFP or one HPSI/LPSI train. Loss of the charging pump in accordance with the single failure criterien would reouire use of the eight groups of pressurfrer heaters and a HPSI/LPSI train fc r pressure control and maintenarce of primary system water level. Loss of the SBFP in accordance vith the single failure criterior wculd require EFW supply by a HPSI/LPSI train, while loss of one HPSI/LPSI train would have no effect whatsoever, since two HPSI/LPSI trains would still be available. We find this to be acceptable.
3.1.?.5. Steam Heating Line Breaks (A) The licensee stated that steam heating line breaks in the PfP er in the switchgear room (SWGR) in the TB would not cause the reactor or the turbine to trip nor cause actuation of the reactor protection syster.
Therefore, in accordance with the provisions of Paragraph B.3.b'(1) of PTP ASB 3-1 of Standard Peview Plan (SRP) Section 3.6.1, " Plant DesMr for Pretrction Against Postulated Piping Failures in Fluid Systems Outside Containment" we conclude that offsite power is not assumed te be urcvailable for shutdown in the event of steam hebting line failures in these areas. In such cases, shutdown may be effected by use of normal equipment such as main feedwater system, condensate system, turbine generator bypass and other nonsafety-related systems. However, for breaks ir steam heating lines in the PAB and SWGP, the following analyses show tFet het shutdown may be attained and maintained with safety-related acuipment with the exception of the pressurizer heaters which have beer fcurd acceptable for use in plant shutdown.
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0-In the Primary Auxiliary Building
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A steam heating,line failure in the shielded portion of the PAB could disable the three charging pumps -leaving the two emergency motor-driven feedwater pumps, the EBFP and three HPSI/LPSI trains available for use in. bringing the plant to hot shutdown.
Primary system pressure control may be effected by maintaining one full capacity group of pressurizer heaters on the offsite or onsite power source, as desired when the primary system is at a pressure in excess of the shutoff head of a HPSI/LPSI train. One HPSI/LPSI train can be
. employed to maintain primary system water level and pressure once the pressure has been reduced to below 1560 psig. EFW may be provided by one of the three EFW pumps. Single active failure of either one of the three HPSI/LPSI trains, or one of the three EFW trains or even of one group of pressurizer heaters would not compromise the ability of the plant to attain hot shutdown (there are a total of eight groups of pressurizer heaters available for pressure control).
A steam heating line failure in the nonshielded portion is bounded by the effect of a steam generator blowdown line break cited above in section 3.1.2.4.
We find the results of a steam heating line break either in the shielded or unshielded portion of the PAB to be acceptable.
(2) In the Turbine Building (TB)
(a) Switchgear Room The only location in the TB with steam heating lines in which a mild environment has to be maintained is the SWGR. The SWGR contains two unit heaters, each supplied by a 1 inch steam line. The licensee noted that only a slight rise in temperature and humidity would occur in the SWGR, even if one steam line did rupture. This conclusion was based upon the assumption that the independent SWGR ventilation system, consisting of a 30,000 CFM fan located in an adjacent equipment room, would continue to operate. In addition, the licensee installed an excess flow check.
valve in each steam supply line, in order to prevent a break of this sort from adversely affecting the SWGR environment. Moreover, in accordance with the provisions of paragraph 3 of appendix B to BTP ASB 3-1, breaks ir. lines 1 inch in diameter (nominal size) and smaller do-
,not need to be considered in the spectra of possible~line breaks.
In view of the foregoing, we find the effects of a steam heating line break in the SWGR to be acceptable.
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e (b) Other Areas in the Turbine Building Other areas of the TB contain steam heating lines; these areas also contain main steam and/or main feedwater lines. Breaks in main steam and main feedwater lines bound those in steam beating lines and, therefore, the analyses of the effects of the steam heating lines in those areas are not required.
We find the results of a steam heating line break in other areas (i.e.,
areas other than the SWGR which was discussed previously) of the TB to be. acceptable.
(B) In the Diesel Generator Building In a September 26, 1985 letter, the licensee stated that a steam heating line break in the Diesel Generator Building (DGB) could lead to an automatic plant trip. In accordance with the provisions of SRP section 3.6.1, the staff assumes in its analysis that a loss of offsite power occurs in conjunction with a single active failure. These conservative assumptions for this event could result in loss of the three 480 Volt emergency buses (including safety injection),-the non-emergency 480 Volt and 2400 Volt non-emergency buses, due to the assumed loss of offsite power, and the steam-driven auxiliary feedwater pump due to the assumed single active failure.
To mitigate the consequences of this postulated scenario, Yankee committed in its September 26, 1985 letter to install excess flow check valves in the steam heating lines to the DGB. This action does not in itself provide the assurance required to provide protection for this postulated event, since the valves to be installed, and the steam heating lines themselves are not safety-related equipment.
To provide assurance of plant safety for this postulated event, two additional features of the Yankee plant provide additional defense in depth. The first is the stability of the grid around the Yankee site.
In its 25 years of operation, all offsite power has been lost only once.
This was November 8, 1965, during the Northeast blackout. This history provides an extremely high reliability figure for the offsite power system.
A second feature is the Safe Shutdown System (SSS) that provides a dedicated shutdown system for Appendix R concerns (fire protection) and alternate shutdown capability for seismic and tornado concerns under
'the SEP. NUREG-0825, " Integrated Plant Safety Assessment, Systematic Evaluation Program, Yankee Nuclear Power Station" dated June 1983, provides the following description of the SSS:
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"The function of the Hot Shutdcwn System (HSS) is to remove decay heat and maintain primary inventory following any event that disables 611 etter means of performing these functions. In order to remove decay heat, the HSS must have the capability to add feedwater to at least one steam generefor, and to vent steam to the atmosphere from the same_ steam generator. The HSS will have the capability to remove deca even thouch coly cne (1) is yrequired.
heat fromInallorder four to (t-) steam primary maintain generators inventory, the HSS must have the capability to add water tc. the primary system. j
" Prior to placing the HSS into' operation, the follering events are assumed to have occurred:
- 1. initiating event, i.e., seismic or tornado,
?. reector shutdown or scram, and all control rods inserted,
- 3. reactor coolant pressure boundary and secondary heat sink boundary have been established,
- 1. a loss of all off-site and on-site AC power has occurred or been caused,
- 5. all _other means of adding water to the steam generators and primary system have been rendered inoperable.
"Fellering this series of events, the HSS is manually placed into operation.
" DESIGN BASES ASSUMPTIONS
- 1. System is designed to remain functional and operational after a seismic event.
- 2. System is adequately' protected against tornado wind loadings and postulated missiles.
- 3. System must be operable within 30 minutes folleeine-tbe event which disables all other cooling systems.
- 4. System must be capable of operating for three (3) days without bringing in additional makeup water or diesel fuel.
- 5. System designed to supply 200 gpm at 1200 psto (see Note 1);
150 gpm to the steam generators-and 50 gpm to the primary system.
- 6. System is independent of all other systems except the fire system (fire tank water supply), the reactor coolant pressure
-boundcry end the secondary heat sink pressure boundary.
- 7. Single failure criteria does not apply since multiple failures have already been assumed."
The~HSS has a dedicated diesel generator which provides electrical poker to dedicated primary and secondary makeup pumps. Water is taken frcm the fire water storage tank, and is pumped through irderordent, seismically cualified piping. Primary makeup ties into the normal charging piping for loop 4. Secondary makeup is piped to the bicwdown piping to provir'e makeup for any or all steam generators.
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In the staff's judgement, the use of excess flow check valves, the proven stability of the Yankee grid, and the SSS provide adequate assurance that the Yankee plant can be safely shut down for a postulated steam heating line break in the DGB.
Therefore, we find acceptable the effect of steam heating line breaks in the PAB, DGB, and TB.
3.2 Discussion and Conclusions
-In this review, the licensee is not required to protect cold shutdown equipment nor to attain cold shutdown. The refore, only equipment needed to attain and maintain hot shutdown has been considered. In addition, it is noted that Yankee Atomic Electric Company had the option of either
- qualifying the equipment required for hot shutdown to conform with the environment resulting from a high energy line break in the area surrounding the equipment or to show that the equipment affected by such a line break is not required in bringing the plant to hot shutdown. The licensee chose the latter course.
We have reviewed the licensce's submittals and concluded that the licensee is able to attain and maintain hot shutdown for high energy line breaks outside
. containment, as noted in the discussion in the sections above. The licensee's approach to identifying the equipment required to be environmentally qualified is acceptable.
3.3 -Other Identified Deficiencies The' approach described by the licensee for addressing and resolving the other deficiencies identified in the December 16, 1982 SER includes replacing equipment, performing additional analyses, utilizing additional qualification documentation beyond that reviewed by FRC, obtaining additional qualification documentation and exempting some equipment from qualification, e.g., located in the mild environment. We discussed the proposed resolutions in detail on an item by item basis with the licensee during the January 17, 1984 meeting. Replacing or exempting equipment, for an acceptable reason, are clearly acceptable methods for resolving environmental qualification deficiencies. The more lengthy discussions with the licensee concerned
~t he use of additional analyses and' documentation. Although we did not review the additional analyses or documentation, we discussed how-the analysis was being used to resolve deficiencies identified in the FRC TER and the content of the additional documentation in order to determine the acceptability of these methods. The licensee's equipment envirormental qualification files will be audited by the staff during follow-up inspections to be performed by Region I, with assistance from IE Headquarters and NRR staff as necessary. Since a significant amount of documentacion has already been reviewed by the staff and FRC, the primary objective of the file audit will be to verify that it contains the appropriate analyses and other necessary documentation to support the licensee's conclusion that the equipment is
.aualified. The inspections will verify that that licensce's program for surveillance and reintenance of environmentally qualified equipment is adequate to assure that this equipment is naintcired in the as-analyzed or tested condition. The methed used for tracking periodic replacement parts, and implementation of the licensee's ccrritments and actions, e.g., restrdit.g replacement of eceipment, will also be verified. .
- Based on our discussions with-the licensee and our review of its submittal, we find the licensee's apprcach for resolving the identified environmental qualification deficiencies acceptable.
3.4 Compliance With' 10 CFR 50.49 In its Verch E. and July 16, 1984 submittals, the licensee has described the epproach used.to identify equipment within the scope of paragraph (b)(1) cf
- 10 CFR 50.49, safety-related equipment relied upon to remain functionel during the following design basis events. The licensee states that floodirp and environmental effects resulting frcm all postulated desicr-hasis events documented in the Yankee Firal Hazards Summary Report including less of coolant accidents and main steam line breaks inside certainment (vapor container) and high energy line breaks outside containment were reviewed in the identification of safety-related electrical eouipment which was to be environmentally qualified. -The flooding and environmental effects resulting from high-energy
- line breaks (PELPI outside containment were also considered in the identification of this equipment. Therefore, all design basis events including acciderts et Yankee were considered in the identification of electrical equipment within the scope of paragraph (b)(1) of 10 CFR 50.49.
The licensce's approach for identifying equipment within the scope af paragraph (b)(1) is in accordance with the requirements of that paragraph, and therefore acceptable.
The method used by the licensee for identification cf electrical equipment within the scope of peregraph (b)(2) of 10 CFR 50.49, nonsafety-related electric equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions, is summarized below:
- 1. A list was generated cf safety-related electric equipment as defined in paragraph (b)(1) of 10 CFR 50.49 required to retain functional during or following design-basis events such as loss of coolant accidents and main steam line brcaks inside containment and HELB outside containment. A system analysis was performed to identify the set of electrical eouipnent which the system required in order to perform its design basis safety function. The list was based on reviews of Technical Specifications. Energency Operating Procedures, Plant Piping and Instrumentation Drawings (P&ID), the Final Hazards Summary Pepcrt for Yankee, and schematics and electrical one line diagrams and control logic' diagrams; and f
- 2. The elementary wiring diagrams of the safety-related electrical equipmer.t identified in Step I were reviewed to identify any auxiliary devices electrically connected directly into the contrcl or power circuitry of the safety-related eouiprent (e.a., automatic trips) whose failure due to postulated environmental conditions could prevent the required operetien of the safety-related equiprtent; and
- 3. The operation of the safety-related systems and equipment were reviewed to identify any directly frechanically connected auxiliary systems 2
with electrical components which are necesu ry for the required operation of the safety-related ecuipirent (e.g., cooling water or lubricatingsystems). This irvolved the review of P&ID, component technical manuals, and/or systems descriptions in the Final 14azards Summary Report; and ,
- 4. Nonsafety-related electrical circuits indirectly associated with the electrical equipnent identified in Step 1 by contron prwer supply or physical proxinity were considered by a review of the original electrical design including the use of applicable industry standards (e.g., IEEE, NEMA, ANSI, UL, and NEC) and the use of properly coordinated protective relays, circuit breakers, and fuses for electrical fault protection.
The licensee states that the results of the above review indicated that no additional electrical equipment was identified which was not previously included on the " Master List." Therefore, the list of electrical equipment provided in its March 5, 1984 submittal is iudged by the licensee to address all electrical equipment within the scope ~of paragraph (b)(2) of 10 CFR 50.49.
We find the rethodology used by the licensee is acceptable since it provides reasonable assurance that equipment within the scope of parapraph (b)(2) of 10 CFR 50.49 has been identified.
With regard to paragraph (b)(3) of 10 CFR 50.49, the licensee evaluated existing system arrangements and identified equipment for the five types of variables defined in Regulatory Guide (RG) 1.97, Rev. 3. A report outlining the results of the review, schedules for modifications where necessery, and justification of deviations not reouiring modifications has been submitted to the NRC for approval. Since the report dealing with RG 1.97 instrumentation is still under review by the staff, sorre of the equipment identified in the report has not been added to the 10 CFR 50.49 scope. However, some of the
- equipment items icintly within the scope of NUREG-0737 and RG 1.97 have been included in the 10 CFR 50.49 scope. When the RG 1.97 report and equipment lists contained therein have been finalized and accepted by the staff, appropriate eovipment not already in the 10 CFR 50.49 scope will be added in accordance with the RG 1.97 implementation schedule.
We find the licensee's approach to identifying equiprent within the scope of paragraph (b)(3) of 10 CFP 50.49 acceptable since it is in acccrdance with the requirements of that paragraph.
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4.0 CONCLUSION
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Based on the above evaluation, we conclude the following with regard to the qualification of electric equipment important to safety within the scope of 10.CFR 50.49.
Yankee Atomic Electric Company's electrical equipment environmental qualification program complies with the requirements of 10 CFR 50.49.
.The proposed resolutions for each of the environmental qualification deficiencies identified in the December 16, 1982.SER and FRC TER-are acceptable.
5.0 ACKNOWLEDGEMENTS ,
- This SE was prepared by P. Shemanski, C. Y. Li, N. Wagner and J. Clifford.
Dated:
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