ML19347B840
| ML19347B840 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 09/29/1980 |
| From: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| NUDOCS 8010160076 | |
| Download: ML19347B840 (1) | |
Text
.
. -..=
.-.=
~.
TLC -
>2 Klog 6
./
o UNITED STATES g
[ h egg g NUCLEAR REGULATORY COMMISSION g/ y REGION V g.-
!.14 g
1990 N. CALIFORNIA DOULEVARD i
j
- N.) [
SulTE 202, WALNUT CREEK PLAZA
%,a WALNUT CREEK, CALIFORNIA 94S96 I
j Septemoer 29, 1980 i
i
)
Docket flo. 50-312 i
i Sacramento Municipal Utility District i
P. O. dox 16330 Sacramento, California 95813 Attention:
iir. John J. Mattimoe Assistant General Manager i
Gentlemen:
Enclosed is IE Supplement do. 2 to dulletin No.79-018.
This infonnation is i
presented in the form of generic questions and answers whicn will assist you in responding to cite actions required in 125-79-018 and tne demorandum and Order (CL1-dO-21) dataa Way 23, 1980 wito regard to environmental qualification of Class iE equipment in use ac your power reactor facility (ies) with an 3
operating license.
Should you nave questions regaruing cnis supplement please contact this office.
Sincerely, 7
flYb E,~s..
R. H. Engelken Director 1
Enclosures:
j 1.
IE Supplement rio. 2 to Bulletin j
71-OlB 2.
fi-Issued IE Bullentins cc w/o enclusures:
R. J. Rodriguez, SMUD L. G. Schwieger, SMUD 5
.~
5'0101600N G.
.. u w.
,.s.
--._e,
~
,,--y
-..,y.,o m,
.,mv
.-,,.w--e.s-w,m-w g-
,,,,>,y e.,gv_,ma+-y.,_v p =
,.,3__.y, w-4*
w 4 w e rw er+aw.yw -r-m
-me w *e
SSINS:
6820 Accession No.: 8003220241 IEB 79-018, Supple. 2 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT D 3hn UASHINGTON, D.C.
20555
_u o
_a
_ Jlrd September 29, 1980 IE Supplement No. 2 to Bulletin 79-01B: ENVIRONMENTAL OUALIFICATION OF CLASS 1E EQUIPMENT Enclosed are the generic questions and answers which resulted # rom NRC/ Licensee meetings in NRC Regional Offices during the week of July 14, 1980 regarding environmental qualification of Class 1E eauipment in use at power reactor facilities. These answers address scecific ouestions asked durin9 the meetings.
Due to the generic nature of some of these ouestions, the staft, is 1ssuing them as a bulletin suoplerent. The reoional neetinos highlighted the fact that in some cases, the scoce and oepth of the 79-01B review was not clear to licensees. Therefore, these answers may affect your 79-018 submittal.
These submittals are required by a separate orcer 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 1,1930, 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 GA0 clearance (B180225 (R0072); clearance expired July 31,1930) specifically for identified generic problems.
Supplerent No. 2 to Bulletin 79-01B is for information, hence no GA0 clearance is required.
Enclosures:
1.
Generic Questions and Answers to IEB-79-01B and Memorandum and Order (CLI-30-21) dated May 23, 1980
IEB 79-01B, Supple. 2 September 29, 1980 RECENTLY ISSUED IE BULLETINS Bulletin No.
Subject Date Issued Issued To 80-22 Automation Industries, 9/11/80 All radiography Model 200-520-008 sealed-licensees source connectors 79-26 Boron loss from BWR 8/29/80 All BUR power Revision 1 control blades facilities with an OL 80-20 Failures of Westinghouse 7/31/80 To each nuclear Type U-2 Spring Return power facility in to Neutral Control Switches your regic, having an OL or u CP 80-19 Failures of Mercury-7/31/80 All nuclear power Uetted Matrix Relays in facilities having Peactor Protective Systems either an OL or a CP of Operating Nuclear Power Plants Designed by Combus-tion Engineering 80-18 itaintenance of Adequate 7/24/80 All PWR power reactor "inimum Flow Thru Centrifugal facilities holding OLs Charging Pumps Following and to those PWRs Secondary Side High Energy nearing licensing Line Rupture Supplement 2 Failures Revealed by 7/22/80 All BWR power reactor to 80-17 Testing Subsequent to facilities holding OLs Failure of Control Rods
)
to Insert During a Scram at a BWR Supplement 1 Failure of Control Rods 7/18/80 All BUR power reactor to 80-17 to Insert During a Scram facilities holding OLs at a BWR 80-17 Failure of Control Rods 7/3/80 All BWR power reactor to Insert During a Scram facilities holding OLs a t a BWR 80-16 Potential Misapplication of 6/27/80 All Power Reactor Rosemount Inc., Models 1151 Facilities with an and 1152 Pressure Transmitters OL or a CP with Either "A" or "D" Output Codes
& & vt i Q.1 Define tha scope of review with respect to the June 1982 deadline.
'lhat is required beyond the June 1982 date for qualification?
A., 1 Sy June 10, 1982, all safety-related electrical eauipment potentially exeosed to a harsh environment in nuclear generating stations, licensed to operate on or before June 30, 1982, shall be cualified to either the DDR nuidelines or NUREG-0588 (as anolicable).
Safety-related electrical eouipment are those recuired in bringinq the plant to a cold shutdown condition and to mitigate the consequences of the accident. The qualification of safety-related electrical equipnent to # unction in environmental extremes, not associated with accident conditions, is the resnonsibility 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 environrents must be completed by June 30, 1982.
The cualification scnedules for consideration of the dynamic loadino of safety-related eauipment (electrical anc nechanical) and the env'.,onnental qualification review of mecnanical equipment are ceing developed.
It is the intention of the staff to initiate this effort as soon as possible.
Q.2 Clarify the requirea submittal dates for CRs, flTOLs, and cps.
What about OLs whose 1003 license is not expected by June 1932?
A.2 The recuired schedule for submitting information in response to the Commission Order and denorandum (CLI-80-21) is provided below.
Plants uno have received an cperating license, either for full or limited power operation, are required to c eet the schedule for operating reactors.
Plants who have committed, to the fiRC, to meet schedules in advance of those provided belcw are required to reet that ccmmitnent.
In all cases, plants are required to have their equipment fully qualified to the applicable standards either by June 30,1931 or by the time the operating license is granted, whichever ccmes later.
Operating Peactors and NTCL (operating license expected by February 1, 1981) i Submittal to be received no later than flovember 1, 1980 OLs (operatino 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 af ter June 30,1982)
Submittal to be received no later than 6 months prior to issuance of operating license.
I
. 0.3 Define the requirements and anplicable criteria for ors, NT0Ls, and OLs.
Specifically address the flT0Ls whose CP SER is prior to July 1974 and after July 1974.
Can a CP whose SER is prior to 1974 use the D0R guidelines?
A.3 Table 1 describes the anolication of each document.
All operating reactors as of May 23, 1980, will be evaluated against the D0R 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 REQUIREMEf1TS ors OLs cps 00R GUIDELINES CP SER CP SER Cefore 7/1/74 After 7/1/74 USE f;UREG-0588 flVREG-0588(CAT. II) f!UREG-0588(CAT.I) flVREG-0588(CAT.I)
AS NECESSARY or PEPLACEMEflT COMPONEllTS f!EW RULE WHEN LSE NUREG-0588 (CAT.I)
Iti 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 MUREG-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 flVREG-0588; unless the licensee made commitment in the construction permit record to use the 1974 standard, or unless the operating licensee applica-tion record indicates that the 1974 standard is to be used, in such cases Column I of flVREG-0588 is to be used.
While there are differences between the Category II column of f!UREG-0528 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 flT0Ls.
Q.4 Clarify the reporting requirements for LERs with respect to Part 50.55e vs79-01B.
l l
Are only
- hose items, known to be unoualified, immediately reportable?
Are items, for which there are no data or for which there are insuf-ficient data, open itens to be resolved, but are not immediately recortable?
A.4 The reauirerent for reporting in IEB 79-01B does not chance the recorting reauirerents defined in the license conditions.
In general, cps should report via 50.55e.
Operating plants should use the LER.
'elhen a determination has been made that reasonable assurance does not exist to ensure that the Class IE electrical eouipment component (s) can oerinrn their safety-related function, that is reportable.
Inadecuate or ne data are factors in this determination.
The time and tecnnical ;udaements required to make the determination should be basec ot. cne slanificance of this specific equipment, components, and the discrepancies.
l Q.5 How does the "0" 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 "0" list provides a source from which the required equipment may be selectea. The information required tc be submitted by November 1, 1980, is for safety-related electrical equipment potentially exposed to a harsh environment resulting fron an accident.
Safety-related l
couirment are those required to help bring the plant to cold shutdown and to nitigate the accident (LCCA, HELB inside or outside containment).
"Mitiaate" 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 nenerate 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 ccnnonents 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 systens and equipment identified above should be included regardless of the original classification when the plant received its operatino 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 environnental 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 operator is to use environmentally qualified safety-related display instrumentation to diagnose failure to perform such safety functions.
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 nonitor 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 l
sensors, valve position indicators, pressure transmitters, thermo-couples, etc., should be qualified to function in the relevant accident environment.
Licensees should, therefore, review their emergency procedures to determine the electrical ccmoonents needed to perform the functions of Safety-Related Display Information, Post Accident Sampling and Monitoring, and Radiation l'onitoring. When equipment implied by the i
emercency procedures is not listed, justificiation must be provided that failure of sucn 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, etc.).
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 j
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 environ-ment.
Conclusions which show that equipment is unqualified should include a basis for continued plant operation.
TABLE II TYPICAL EQUIPMEllT/FUf!CTI0ilS flEEDED FOR MITIGATIO!! 0F A LOCA OR MSLB ACCIDENT Engineered Safeguards Actuation Rea; tor Protection Concainment Isolation Steamline Isolation Main Feedwater Shutdown and Isolation Emergency Power i
\\
i i l Emergency Core Cooling Containment Heat Removal Containrent Fission Product Renoval Containment Combustible Gas Control Auxiliary Feedwater Containment Ventilation l
Containment Radiation Monitorinq ContrM nuom Habitability Systems (e.g., HVAC, Radiation Filters) l l
Ventilation for Areas Ccntaining Safety Equipment l
Component Cooling l
Emergency Shutdown l
Post Accident Samolina and Monitoring l
Radiation Ponitorina Safety Related Display Instrumentation (1) These systems will differ for PWRs and BWRs and for older and newer l
plants.
In eacn case, the system features wnicn allow for transfer to j
recirculation cooling mode and establishment of long-term cooling with boron precipitation controi are to be consiaered as part of the system to be evaluatea.
(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 br'ach of the reactor coolant pressure boundary together with a rapid l
depressurization of the reactor coolant system.
Examples of such systems i
and equipment are the RilR system, FORVs, RCIC, pressurizer sprays, chemical and volume control system, and steam dump systems.
(3) More specific identification of these types of equipment can be found in the plant emergency procedures.
Q.6 flUREG-05E3 was issued for conr:ent.
'!ill any changes impact the l
requirements established by the Commission memorandum and order?
l Will the dauqhter standards referenced be corrected / changed?
A.6 The requirement established by the Cctmission memorandum and order l
will not change as a result of comments on f!UREG-0588.
!!o substan-l tive changes are anticipated in flVREG-0583 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?
l A.7 The methods and procedures relating to design stress analysis, aging i
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 environments.
1 i
j l 4
I Q.8 Provide the staff's definition of " central location" for qualifica-l tion documentation. What documentation is expected to be maintained?
Will it be acceptable to maintain summary test reports at the utility j
central file and provide a reference to the flSSS Vendor's file for the actual test reports? Does ?!RC require test reports to be sub-mitted to support qualification?
i A.8 The central location should be at the utilities corporate head-4 quarters or plant site.
Both the D0R guidelines and flVREG-0588 specify that sufficient information must be available to verify that l
the safety-related electrical equipment has been qualified in
}
accordance with the guidance and requirements.
Details for the j
information and documentation required for type tests, operating experience, analysis, and extrapolation of test data from operating j
experience are provided in Section 5 of fiUREG-0588 and Section 8 of j
IEEE Std. 323-74.
1 i
The staff will accept summary test reports maintained at the utility's central file which reference the actual test reports and j
data available in a single location at the flSSS vendor's facility.
i The Licensee / Applicant must make the determination that necessary 1
information and documentation, to support qualification of equipment, is in conformance with DOR guidelines and tiUREG-0588.
This vendor 1
information file must be maintained current, auditable and available i
throughout the life of the referencing plant.
I 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.
j Q.9 The staff was directed to codify, by Technical Specification, some i
of the requirenents of the Order.
Can you give some of the details i
of t'is requirement, how the staff expects to meet this directive
]
and when?
A.9 The staff has proposed to the Commission changes to the Technical l
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 equipment.
a 0.10 With respect to the flRC data base, how will utilities address and obtain information from it?
f 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.
i Q.11 How should submittals containing data and qualification information i
be submitted? What format should we use if we have several facili-l' ties at different stages (OR, NT0L, CP)?
l
\\
1
A.11 The qualification information and data should be subnitted with the approcriate 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 olants in the SEP p rog ran.
Either format is acceotable.
Q.12 Is testina required of eauipment which comcletes 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 reouire that the nuclear instrumentation and its associated components be environmentally oualified for a LOCA or HELB. The nuclear instrumentation system is used for transient conditions but is not recuired 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 I hour in excess of the time assumed in the accident analysis.
The staff has indicated that time is the most significant factor in terns of tne marains required to provide an acceptable conridence level that a safety-relateo function will be completed.
Cur judgment of at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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 eouipment specifying confor-mance to IEEE Std 323-1971.
For sequential testing, how do we factor in aging?
If early test failure occurs due to "non E-Q" mechanisns, can the test be extrapolated using analytical methods?
1 A.13 Sequential testing requirements are specified in i;UREG-0588 and the DDR guidelines.
Licensees must follow the test requirements of the 1
applicable document.
1.
If the test has been conpleted 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 category.
When a failure occurs due to a non-E0 related mechanism, acceptability of analysis to extrapolate the test data would be dependent on several considerations (e.g., the specific function being demonstrated, the
_g_
failure nechanir.m. when the failure occurred, etc.), may be very difficult to acnieve.
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 environ-mental 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 separa tely.
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 D0R Guidelines and fiUREG-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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> be used for BWR drywells or that 340 F for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 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.
i i
l
-9 l
The adecuacy of a plant-specific crofile is dependent on the assump-l tion and design considerations at the time the profiles were developed. The DCR guidelines and MUREG-0588 provide guidance and considerations recuired to deternine if the plant-specific profiles encomcass the LOCA and HELB inside containment.
1 j
Q.16 Could you elaborate on what the staff expects with regard to quality l
assurance?
If parts or subcomponents are purchased from a vendor who does not have a cuality assurance procram, can it be cualified to meet IEEE Std. 323-74 reauirenents?
A.16 The OA orocrans should acccmmodate any increased scope due to the new environmental qualification documentation recuirements.
Proce-l dures incorporated by the licensee for data accuisition should be l
documented and available for staff review upon request.
Requirements for 0A proarams are provided in Part 50, Appendix B, of the Code of Federal Regulations.
Part 59, 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 j
licensee /aoplicant to ensure that the applicable quality assurance j
procedures for their plant are met.
l In deternining the qualification status of existing equipment purchased frcn a vendor, where a OA program did not exist, the utility should consider the following:
1.
The complexity of design, complexity of manufacturing process, and end use.
2.
Past nerformance of vendor.
3.
Past operating history of products, especially similar products, made by vendor.
l 4.
Procedures, equipment, and results of environmental qualifica-tion testing relative to those for other equipment for which a l
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
, memorandun.
After May 1980, all carts used to replace presently installed parts shall be qualified to Category I of NUREG-0588 "unless there are sound reasons to the centrary." Nonavailabili ty 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 weiahina 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.
Justifica-tion for deviation from Category I or MUREG-0588 shall be documented by the licensee and records shall be available for audit, upon request by the NRC.
Q.18 D0R Guidelines, NUREG-0588 and MUREG-0578, define or give guidance for calculating radiation source terns.
However, since one is more restrictive than the other, wnich do we use?
A.18 Both the 00R 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 /15 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 containment.
NUREG-0573 provides the radiation source term to be used for deter-mining the qualification doses for equipment in close proximity to recirculting 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-0538 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 /0% particulates).
When developing radiation source terns 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)
I
. Inside Containment Larner of l
NUREG-0508 NUREG-0588 (100/50/1 (10/10/0 in containrent) in RCS) or NUREG-0578 (100/50/1 in RCS)
Q.19 Can carma eouivalents be used rather than beta exposure for radiation qualification?
A.19 Yes.
Gar.na eouivalents may be used wnen consideration of the contri-butions of beta exoosure have been included in accorcance with the guidance given in the D0R 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 af ter completing its required action, must it be qualified for submergence?
A.20 If the eauipment (1) meets the guiadance and requirements of the DOR guidelines or NUREG-0588 for tne 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 frca that portion (submergence) of qualification.
Q.21 What qualification is required of Reactor Pressure Vessel internal instrumentation (e.g., thermocouples) and new instruments requircd 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 require-ments as other safety-related electrical equipment.
The guidance and requirements of NUREG-0588 referenced daughter standards, and Req Guides will be used by the staff in assessing the adequacy of the qualificat on information.
The in-core enviror. ment should consider the worst source ter.n for radiation effects, the worst humidity for tLa corresponding temperature, and high temperatures consistent with that of a damaged core.
Q.22 Is qualification "by use" an acceptable method (e.g., CRCM's in BURS)?
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 qualification
. 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 cualification information.
Q.23 How long should "long tern" equipment be cualified for environmental qualification?
A.23 "Lona tern" for the ournose of nualifying equionent for a harsh environment is variable.
A determination of "long term" for qualifi-cation of equipment should be based on the considerations listed below for each postulated accident scenario.
Justification for the value used should te provided with the equipment qualification documentation.
1.
The time period over which the eouipment is reouired 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 requirer. 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.
J s