Bulletin 79-01B: Supplement No. 2, Environmental Qualification of Class IE Equipment

                                                      SSINS: 6820          
                                                      Accession No.:       
                                                      IEB 79-01B, Supple. 2 

                                UNITED STATES
                           WASHINGTON, D.C. 20555
                             September 30, 1980

                                        1E EQUIPMENT 

Enclosed are the generic questions and answers which resulted from 
NRC/Licensee meetings in NRC Regional Offices during the week of July 14, 
1980 regarding environmental qualification of Class 1E equipment in use at 
power reactor facilities. These answers address specific questions asked 
during the meetings. Due to the generic nature of some of these questions, 
the staff is issuing them as a bulletin supplement. The regional meetings 
highlighted the fact that in some cases, the scope and depth of the 79-01B 
review was not clear to licensees. Therefore, these answers may affect your 
79-01B submittal. These submittals are required by a separate order to be 
completed by November 1, 1980. 

Some answers given in Supplement No. 1 to IEB-79-01B are superseded by these
answers. For example, in Bulletin Supplement No. 1, issued on February 29, 
1980, the answer to question No. 5 specified that TMI lessons learned 
equipment was not included in the review. However, due to the extension of 
the response date from April 14, 1980 to November 1, 1980, this equipment is
now being addressed since its installation is either complete or required 
before the issuance of the February 1, 1981 SER. (See Question No. 21 of 
this Supplement.) 

No specific response is requested by this Supplement; however, all answers 
contained in the enclosure to this Supplement should be carefully reviewed 
and considered for applicability in your response to IEB 79-01B. 

IE Bulletin No. 79-01B was issued under a blanket GAO clearance (B18O225 
(ROO72); clearance expired July 31, 1980) specifically for identified 
generic problems. Supplement No. 2 to Bulletin 79-01B is for information, 
hence no GAO clearance is required. 

1.   Generic Questions and Answers
       to IEB-79-01B and Memorandum
       and Order (CLI-80-21) dated
       May 23, 1980

Q.1      Define the scope of review with respect to the June 1982 deadline. 
          What is required beyond the June 1982 date for qualification? 

          By June 30, 1982, all safety-related electrical equipment 
          potentially exposed to a harsh environment in nuclear generating 
          stations, licensed to operate on or before June 30, 1982, shall be
          qualified to either the DOR guidelines or NUREG-0588 (as 
          applicable). Safety-related electrical equipment are those 
          required in bringing the plant to a cold shutdown condition and to 
          mitigate the consequences of the accident. The qualification of 
          safety-related electrical equipment to function in environmental 
          extremes, not associated with accident conditions, is the 
          responsibility of the licensee to evaluate and document in a form 
          that will be available for the NRC to audit. Qualification to 
          assure functioning in mild environments must be completed by June 
          30, 1982. 

          The qualification schedules for consideration of the dynamic 
          loading of safety-related equipment (electrical and mechanical) 
          and the environmental qualification review of mechanical equipment 
          are being developed. It is the intention of the staff to initiate 
          this effort as soon as possible. 

Q.2       Clarify the required submittal dates for ORs, NTOLs, and CPs. What
          about OLs whose 100% license is not expected by June 1982? 

A.2       The required schedule for submitting information in response to 
          the Commission Order and Memorandum (CLI-80-21) is provided below. 
          Plants who have received an operating license, either for full or 
          limited power operation, are required to meet the schedule for 
          operating reactors. Plants who have committed to the NRC, to meet 
          schedules in advance of those provided below are required to meet 
          that commitment. In all cases, plants are required to have their 
          equipment fully qualified to the applicable standards either by 
          June 30, 1980, or by the time the operating license is granted, 
          whichever comes later. 

          Operating Reactors and NTOL (operating license expected by 
          February 1, 1981) 

          -    Submittal to be received no later than November 1, 1980 

          OLs (operating license expected by June 30, 1982) 

          -    Submittal to be received no later than 4 months prior to 
               issuance of operating license 

          OLs and CPs (operating license expected after June 30, 1982) 

          -    Submittal to be received no later than 6 months prior to 
               issuance of operating license. 

                                    - 2 -

Q.3       Define the requirements and applicable criteria for ORs, NTOLs, 
          and OLs. Specifically address the NTOLs whose CP SER is prior to 
          July. 1974 and after July 1974. Can a CP whose SER is prior to 
          1974 use the DOR guidelines? 

A.3       Table 1 describes the application of each document. All operating 
          reactors as of May 23, 1980, will be evaluated against the DOR 
          guidelines. In cases where the DOR guidelines do not provide 
          sufficient detail, but NUREG-0588 Category II does, NUREG-0588 
          will be used. 

                                  TABLE 1

     ORs                      OLs                           Cps

DOR GUIDELINES      CP SER              CP SER
                    Before 7/1/74       After 7/1/74        

USE NUREG-0588      NUREG-0588(CAT.II)  NUREG-0588(CAT.I)  NUREG-0588(CAT.I)

REPLACEMENT COMPONENTS                                      NEW RULE WHEN
USE NUREG-0588 (CAT.I)                                      IN EFFECT

          All plants licensed after May 23, 1980, shall conform to 
          NUREG-0588. In accordance with Regulatory Guide 1.89, all such 
          operating licenses for facilities whose construction permit SER is
          dated July 1, 1974 or later, are to be reviewed against IEEE Std. 
          323-1974. Thus, for these licensees, the operating license 
          applicant is to qualify equipment to the Category I column in 
          NUREG-0588. For operating licenses issued after May 23, 1980, 
          whose construction permit SER is dated before July 1, 1974, the 
          operating license applicant is to qualify equipment to at least 
          Category II column of NUREG-0588; unless the licensee made 
          commitment in the construction permit record to use the 1974 
          standard, or unless the operating licensee application record 
          indicates that the 1974 standard is to be used, in such cases 
          Column I of NUREG-0588 is to be used. 

          While there are differences between the Category II column of 
          NUREG-0588 and the DOR guidelines, the differences are in details 
          and in the optional part of the documents. The minimum 
          requirements set forth by these documents are general and 
          compatible. Thus, the minimum standards set by either of the two 
          documents are equally applicable to ORs and NTOLs. 

Q.4       Clarify the reporting requirements for LERS with respect to Part 
          50.55e vs 79-01B. 

                                    - 3 -

          Are only those items, known to be unqualified, immediately 
          reportable? Are items, for which there are no data or for which 
          there are insufficient data, open items to be resolved, but are 
          not immediately reportable? 

A.4       The requirement for reporting in IEB 79-01B does not change the 
          reporting requirements defined in the license conditions. In 
          general, CPs should report via 50.55e. Operating plants should use
          the LER. 

          When a determination has been made that reasonable assurance does 
          not exist to ensure that the Class IE electrical equipment 
          component(s) can perform their safety-related function, that is 
          reportable. Inadequate or no data are factors in this 
          determination. The time and technical judgements required to make 
          the determination should be based on the significance of this 
          specific equipment, components, and the discrepancies. 

Q.5       How does the "Q" list review interface with the EQB effort? Can 
          the NRC provide more specific guidance on how to pick out the 
          required safety-related equipment? 

A.5       The "Q" list provides a source from which the required equipment 
          may be selected. The information required to be submitted by 
          November 1, 1980, is for safety-related electrical equipment 
          potentially exposed to a harsh environment resulting from an 
          accident. Safety-related equipment are those required to help 
          bring the plant to cold shutdown and to mitigate the accident 
          (LOCA, HELB inside or outside-containment). "Mitigate" includes 
          safety-related functions such as containment isolation, and 
          prevention of significant release of radioactive material. 

          In order to "pick out" the safety-related equipment, the licensee 
          should generate a list of safety functions typically performed by 
          plant safety systems. Examples are listed in Table II. For each 
          safety function identified in Table II, list the systems, 
          subsystems, or components assumed available in the plant FSAR or 
          emergency procedures to perform that function during a LOCA or any
          HELB inside or outside containment. If a plant specific safety 
          function not listed in Table II is identified, that function and 
          the corresponding systems or equipment to perform the function 
          should be added to the licensee's list. 

          The systems and equipment identified above should be included 
          regardless of the original classification when the plant received 
          its operating license; i.e., some control grade equipment will 
          probably be named in emergency procedures. However, if plant 
          emergency procedures specify a preferred mode of accident 
          mitigation involving equipment recognized by the licensee as 
          unlikely to meet environmental qualification criteria, an 
          alternate mode of performing the safety function and qualifiable 
          equipment may be identified. In such cases, the emergency 
          procedures must clearly indicate how the 

                                    - 4 -

          operator is to use environmentally qualified safety-related 
          display instrumentation to diagnose failure to perform such safety 

          Plant emergency procedures typically include provisions for the 
          operator to sample or monitor radioactivity levels or combustible 
          gas levels, to confirm that valves are in the correct position, to
          monitor flow or temperature, etc. Some of these functions are 
          essential for correct operator action, to mitigate accidents, and 
          prevent radioactive releases. When this is the case, the radiation
          sensors, valve position indicators, pressure transmitters, 
          thermocouples, etc., should be qualified to function in the 
          relevant accident environment. 

          Licensees should, therefore, review their emergency procedures to 
          determine the electrical components needed to perform the 
          functions of Safety-Related Display Information, Post Accident 
          Sampling and Monitoring, and Radiation Monitoring. When equipment 
          implied by the emergency procedures is not listed, justification 
          must be provided that failure of such equipment would not prevent 
          accident mitigation or release of radioactivity. 

          Equipment now indicated in emergency procedures in response to 
          TMI-2 Lessons Learned should be listed. Equipment which is or will
          be installed due to TMI Lessons Learned should be addressed 
          similar to other existing safety-related equipment (e.g., 
          saturation meter, sump level indicators, torus water volume, 

          The licensee should document anticipated service conditions in 
          every portion of the plant where the environment could be 
          influenced by the accident or its consequences. These service 
          conditions should also be correlated with the safety-related 
          systems and subsystems identified above. Whenever an item of 
          safety-related equipment may be located in an environment outside 
          the range of normal conditions, due to the harsh environment 
          resulting from the accident, and the equipment is needed to 
          mitigate the consequences of the accident, place it on the list of
          equipment in a potentially hostile environment. Conclusions which 
          show that equipment is unqualified should include a basis for 
          continued plant operation. 

                                 TABLE II

     Engineered Safeguards Actuation
     Reactor Protection
     Containment Isolation
     Steamline Isolation
     Main Feedwater Shutdown and Isolation
     Emergency Power

                                    - 5 -

     Emergency Core Cooling
     Containment Heat Removal
     Containment Fission Product Removal
     Containment Combustible Gas Control 
     Auxiliary Feedwater
     Containment Ventilation
     Containment Radiation Monitoring
     Control Room Habitability Systems (e.g., HVAC, Radiation Filters)
     Ventilation for Areas Containing Safety Equipment
     Component Cooling
     Service Water
     Emergency Shutdown
     Post Accident Sampling and Monitoring
     Radiation Monitoring
     Safety Related Display Instrumentation

(1)  These systems will differ for PWRs and BWRs and for older and newer 
     plants. In each case, the system features which allow for transfer to 
     recirculation cooling mode and establishment of long-term cooling with 
     boron precipitation control are to be considered as part of the system 
     to be evaluated. 

(2)  Emergency shutdown systems include those systems used to bring the 
     plant to a cold shutdown condition following accidents which do not 
     result in a breach of the reactor coolant pressure boundary together 
     with a rapid depressurization of the reactor coolant system. Examples 
     of such systems and equipment are the RHR system, PORVs, RCIC, 
     pressurizer sprays, chemical and volume control system, and steam dump 

(3)  More specific identification of these types of equipment can be found 
     in the plant emergency procedures. 

Q.6       NUREG-0588 was issued for comment. Will any changes impact the 
          requirements established by the Commission memorandum and order? 
          Will the daughter standards referenced be corrected/changed? 

A.6       The requirement established by the Commission memorandum and order
          will not change as a result of comments on NUREG-0588. No 
          substantive changes are anticipated in NUREG-0588 or in referenced
          daughter standards. A revision is anticipated, making corrections.

Q.7       Can IEEE Std. 650 (Standards for Qualification of Class IE static 
          battery chargers and invertors for nuclear power generating 
          stations) be used for qualifying the balance of plant components 
          which are not exposed to harsh environments? 

A.7       The methods and procedures relating to design stress analysis, 
          aging of electrical/electronic components and the stress test 
          identified in this standard are acceptable for qualifying the 
          balance of plant components which are not exposed to harsh 

                                    - 6 -

Q.8       Provide the staff's definition of "central location" for 
          qualification documentation. What documentation is expected to be 
          maintained? Will it be acceptable to maintain summary test reports
          at the utility central file and provide a reference to the NSSS 
          Vendor's file for the actual test reports? Does NRC require test 
          reports to be submitted to support qualification? 

A.8       The central location should be at the utilities corporate 
          headquarters or plant site. Both the DOR guidelines and NUREG-0588
          specify that sufficient information must be available to verify 
          that the safety-related electrical equipment has been qualified in
          accordance with the guidance and requirements. Details for the 
          information and documentation required for type tests, operating 
          experience, analysis, and extrapolation of test data from 
          operating experience are provided in Section 5 of NUREG-0588 and 
          Section 8 of IEEE Std. 323-74. 

          The staff will accept summary test reports maintained at the 
          utility's central file which reference the actual test reports and
          data available in a single location at the NSSS vendor's facility.
          The Licensee/Applicant must make the determination that necessary 
          information and documentation, to support qualification of 
          equipment, is in conformance with DOR guidelines and NUREG-0588. 
          This vendor information file must be maintained current, auditable
          and available throughout the life of the referencing plant. 

          Test reports are not required to be submitted. Test report 
          references must be included in the plant submittals and these 
          reports must be available for staff review on demand. 

Q.9       The staff was directed to codify, by Technical Specification, some
          of the requirements of the Order. Can you give some of the details
          of this requirement, how the staff expects to meet this directive 
          and when? 

A.9       The staff has proposed to the Commission changes to the Technical 
          Specifications (e.g., Appendix A Section 6.10 of the license) 
          which require the establishment and maintenance of a centrally 
          located file which will contain the information necessary to 
          verify the qualification adequacy of all safety-related electrical 

Q.10      With respect to the NRC data base, how will utilities address and 
          obtain information from it? 

A.10      The industry access method for the data base will be addressed in 
          the final stages of system development. This information should be
          available by mid-1981. Licensees will be informed at that time. 

Q.11      How should submittals containing data and qualification 
          information be submitted? What format should we use if we have 
          several facilities at different stages (OR, NTOL, CP)? 

                                    - 7 -

A.11      The qualification information and data should be submitted with 
          the appropriate officer's notarized sworn statements. The format 
          for the data should be in accordance with the format provided in 
          I&E Bulletin 79-01B or the letters provided to the plants in the 
          SEP program. Either format is acceptable. 

Q.12      Is testing required of equipment which completes its 
          safety-related function within the first minute(s) of a LOCA or 
          HELB? (E.g., nuclear instrumentation or other instruments 
          providing RPS inputs, isolation valves, etc.) 

A.12      The staff does not require that the nuclear instrumentation and 
          its associated components be environmentally qualified for a LOCA 
          or HELB. The nuclear instrumentation system is used for transient 
          conditions but is not required for a LOCA or HELB. 

          The staff does require that equipment designed to perform its 
          safety-related function within a short time into an event be 
          qualified for a period of at least 1 hour in excess of the time 
          assumed in the  accident analysis. The staff has in indicated that
          time is the most significant factor in terms of the margins 
          required to provide an acceptable confidence level that a 
          safety-related function will be completed. Our judgment of at 
          least 1 hour is based on the acceptance of a type test for a 
          single unit and the spectrum of accidents (small and large breaks) 
          bounded by the single test: Also see answer to question 21. 

Q.13      Testing is currently being performed on some equipment, and 
          contracts have been issued for testing additional equipment 
          specifying conformance to IEEE Std 323-1971. For sequential 
          testing, how do we factor in aging? If early test failure occurs 
          due to "non E-Q" mechanisms, can the test be extrapolated using 
          analytical methods? 

A.13      Sequential testing requirements are specified in NUREG-0588 and 
          the DOR guidelines. Licensees must follow the test requirements of 
          the applicable document. 

          1.   If the test has been completed without aging in sequence, 
               justification for such a deviation must be submitted. 

          2.   If testing of a given component has been scheduled but not 
               initiated, the test sequence/program should be modified to 
               include aging. 

          3.   Test programs in progress should be evaluated regarding the 
               ability to comply by incorporating aging in the proper 
               sequence.  These would then fall in the first or second 

          When a failure occurs due to a non-EQ related mechanism, 
          acceptability of analysis to extrapolate the test data would be 
          dependent on several considerations (e.g., the specific function 
          being demonstrated, the 

                                    - 8 -

          failure mechanism, when the failure occurred, etc,) may be very 
          difficult to achieve. If such a failure occurs it may be more 
          prudent to correct the failure and continue with the test. 

Q.14      What is the definition of harsh environment? How are the 
          environmental profiles defined outside containment? 

A.14      Harsh environment is defined by the limiting conditions, as 
          specified in IE Bulletin 79-01B, resulting from the entire 
          spectrum of LOCAs HELBs. Specifically, the harsh environment from 
          a LOCA considers the worst parameters resulting over the spectrum 
          of postulated break sizes, break locations and single failures. 
          Similarly, the HELBs inside and outside of containment consider 
          the spectrum of breaks including main steam and feedwater line 
          breaks. The parameters to be considered are: temperature, 
          pressure, humidity, caustic spray, radiation, duration of 
          exposure, aging and submergence. Mechanical and flow-induced 
          vibrations and seismic effects will be considered separately. 

          Environmental profiles for HELB outside of containment have not 
          been generically established due to the uniqueness of each 
          facility. Service conditions for areas outside containment exposed
          to a HELB  must be evaluated on a plant-by-plant basis. Each of 
          the parameters listed above must be considered. Acceptable 
          engineering methods should be used for this calculation. 
          Temperature and pressure history may be available from earlier 
          HELB evalations. The radiation source terms are discussed under 
          Question 18 below. Further guidance for selecting the piping 
          systems and conducting the review are delineated in Regulatory 
          Guide 1.46 and Standard Review Plans 3.6.1 and 3.6.2. 

Q.15      The DOR Guidelines and NUREG-0588 give time and temperature 
          parameters. Can we use different values of these parameters? Will 
          plant-specific profiles still be with the guidance provided? 

Q.15      For minimum high temperature conditions in pressure-suppression-
          type containments, we do not require that 340 F for 6 hours be 
          used for BWR drywells or that 340 F for 3 hours be used for PWR 
          ice condenser lower compartments. These values are a screening 
          device, per the Guidelines, and can be used in lieu of a 
          plant-specific profile, provided that expected pressure and 
          humidity conditions as a function of time are accounted for. 

          In general, the containment temperature and pressure conditions as
          a function of time should be based on analyses in the FSAR. 
          However, these conditions should bound those expected for coolant 
          and steam line breaks inside the containment with due 
          consideration of analytical uncertainties. The steam line break 
          condition should include superheated conditions: the peak 
          temperature, and subsequent temperature/pressure profile as a 
          function of time. If containment spray is to be used, the impact 
          of the spray on required equipment should be accounted for. 

                                    - 9 -

          The adequacy of a plant-specific profile is dependent on the 
          assumption and design considerations at the time the profiles were
          developed. The DOR guidelines and NUREG-0588 provide guidance and 
          considerations required to determine if the plant-specific 
          profiles encompass the LOCA and HELB inside containment. 

Q.16      Could you elaborate on what the staff expects with regard to 
          quality assurance? 

          If parts or subcomponents are purchased from a vendor who does not
          have a quality assurance program, can it be qualified to meet IEEE
          Std. 323-74 requirements? 

A.16      The QA programs should accommodate any increased scope due to the 
          new environmental qualification documentation requirements. 
          Procedures incorporated by the licensee for data acquisition 
          should be documented and available for staff review upon request. 
          Requirements for QA programs are provided in Part 50, Appendix B, 
          of the Code of Federal Regulations. 

          Part 50, Appendix B of the Code of Federal Regulations states that
          the applicant/licensee shall be responsible for the establishment 
          and execution of quality assurance programs. Specifically in 
          purchasing parts or components, it is the responsibility of the 
          licensee/applicant to ensure that the applicable quality assurance
          procedures for their plant are met. 

          In determining the qualification status of existing equipment 
          purchased from a vendor, where a QA program did not exist, the 
          utility should consider the following: 

          1.   The complexity of design, complexity of manufacturing 
               process, and end use. 

          2.   Past performance of vendor. 

          3.   Past operating history of products, especially similar 
               products, made by vendor. 

          4.   Procedures, equipment, and results of environmental 
               qualification testing relative to those for other equipment 
               for which a QA program was applied. 

Q.17      Define the requirements for "replacement parts." Are they the same 
          for "spare" parts? Clearly discuss the alternatives for existing 
          inventories of parts/components. If equipment is ordered to meet 
          IEEE Std. 323-1974 standard but lead time exceeds June 1982, can 
          we use IEEE Std. 323-1971 qualified components in the interim? 

A.17      The requirements for "replacement" and "spare" parts are the same 
          for the purposes of complying with the Commission order and 

                                   - 10 -

          memorandum. After May 1980, all parts used to replace presently 
          installed parts shall be qualified to Category I of NUREG-0588 
          "unless there are sound reasons to the contrary." Nonavailability 
          and/or the fact that the part to be used as a replacement is a 
          spare part purchased prior to May 23, 1980, and is in stock are 
          among the factors to be considered in weighing whether there are 
          "sound reasons to the contrary." All replacement parts shall as a 
          minimum conform to the requirements described in the answer to 
          question 3. Justification for deviation from Category I or 
          NUREG-0588 shall be documented by the licensee and records shall 
          be available for audit, upon request by the NRC. 

Q.18      DOR Guidelines, NUREG-0588 and NUREG-0578, define or give guidance
          for calculating radiation source terms. However, since one is more
          restrictive than the other, which do we use? 

A.18      Both the DOR guidelines and NUREG-0588 are similar in that they 
          provide the methods for determining the radiation source term when
          considering LOCA events inside containment (100% noble gases/50% 
          iodine/1% particulates). These methods consider the radiation 
          source term resulting from an event which completely depressurizes
          the primary system and releases the source term inventory to the 

          NUREG-0578 provides the radiation source term to be used for 
          determining the qualification doses for equipment in close 
          proximity to recirculating fluid systems inside and outside of 
          containment as a result of LOCA. This method considers a LOCA 
          event in which the primary system may not depressurize and the 
          source term inventory remains in the coolant. 

          NUREG-0588 also provides the radiation source term to be used for 
          qualifying equipment following non-LOCA events both inside and 
          outside containment (10% noble gases/10% iodine/O% particulates). 

          When developing radiation source terms for equipment 
          qualification, the licensee must ensure consideration is given to 
          those events which provide the most bounding conditions. The 
          following table summarizes these considerations: 

                                        LOCA                NON-LOCA HELB

Outside Containment                     NUREG-0578          NUREG-0588
                                        (100/50/1           (10/10/0
                                         in RCS)             in RCS)

                                   - 11 -

Inside Containment                      Larger of

                                        NUREG-0588          NUREG-0588
                                        (100/50/1           (10/10/0
                                         in containment)    in RCS)


                                         in RCS)

Q.19      Can gamma equivalents be used rather than beta exposure for 
          radiation qualification? 

A.19      Yes. Gamma equivalents may be used when consideration of the 
          contributions of beta exposure have been included in accordance 
          with the guidance given in the DOR guidelines and NUREG-0588. 
          Cobalt 60 is one acceptable gamma radiation source for 
          environmental qualification of safety-related equipment. Cesium 
          137 may also be used. 

Q.20      If a piece of equipment will become submerged after completing its
          required action, must it be qualified for submergence? 

A.20      If the equipment (1) meets the guiadance and requirements of the 
          DOR guidelines or NUREG-0588 for the LOCA and HELB (small and 
          large breaks) accidents and (2) licensees demonstrate that its 
          failure will not adversely affect any safety-related function or 
          mislead the operator after submergence, the equipment could be 
          considered exempt from that portion (submergence) of 

Q.21      What qualification is required of Reactor Pressure Vessel internal
          instrumentation (e.g., thermocouples) and new instruments required
          as the result of TMI Lessons Learned? 

A.21      TMI Lessons Learned instrumentation will be considered in the 
          February 1, 1981 SER. This equipment is subject to the same 
          requirements as other safety-related electrical equipment. The 
          guidance and requirements of NUREG-0588 referenced daughter 
          standards, and Reg Guides will be used by the staff in assessing 
          the adequacy of the qualification information. The in-core 
          environment should consider the worst source term for radiation 
          effects, the worst humidity for the corresponding temperature, and
          high temperatures consistent with that of a damaged core. 

Q.22      Is qualification "by use" an acceptable method (e.g., CRDM's in 

A.22      Qualification by use has limited application. Often the equipment 
          has never seen the harsh environment and no conclusions can be 
          drawn as to its operability in a harsh environment. Some 

                                   - 12 -

          based on operating experience is a recognized method subject to 
          the requirements of NUREG-0588 and the Guidelines. Credit can be 
          taken for the natural aging of the equipment and for the location 
          of the equipment or other portions of the overall qualification 

Q.23      How long should "long term" equipment be qualified for 
          environmental qualification? 

A.23      "Long term" for the purpose of qualifying equipment for a harsh 
          environment is variable. A determination of "long term" for 
          qualification of equipment should be based on the considerations 
          listed below for each postulated accident scenario. Justification 
          for the value used should be provided with the equipment 
          qualification documentation. 

          1.   The time period over which the equipment is required to bring
               the plant to cold shutdown and to mitigate the consequences 
               of the accident. 

          2.   The ability to change, modify or add equipment during the 
               course of the accident or in mitigating its effects which 
               will provide the same safety-related function. 

Q.24      Why do we want component surface temperature rather than the bulk 
          environment temperature? 

A.24      Temperature measurements are required during the qualification 
          testing to establish that the component was subjected to the most 
          severe temperature environment postulated to occur. These 
          temperature measurements are required to be made as close to the 
          component surface as practicable to ensure that they are 
          representative of the environment in which the component is 
          tested. The surface temperature of the component, although not 
          specifically required, is considered to be a conservative 
          measurement of the test temperature environment. 

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