ML19347B832
| ML19347B832 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 09/29/1980 |
| From: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Goodwin C PORTLAND GENERAL ELECTRIC CO. |
| References | |
| TAC-42502, NUDOCS 8010160058 | |
| Download: ML19347B832 (1) | |
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NUCLE.iB 9EGULATORY COMMISSION g
y REGION V o
j 1990 N. CALIFORNIA BOULEVARD
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- e SulTE 2t,2 WALNUT CREEK PLAZA l
4,,,e WALNUT CREEK, CALIFORNIA 94596 September 29, 1960 i
Docket ilo. 50-344 Portland General Electric Company 121 S. W. Salmon Street i
Portland, Oregon 97204 1
l Attention:
Mr. Charles Gooawin i
Assistant vice President Gentlemen.
Enclosed is IE Supple,nent ilo. 2 to 3ulletin ilo.79-018.
This information is presented in the form of generic questions and answers which will assist you in responding to the actions required in IEB-79-01B and tne Hemoranauta ana Order (CLI-oU-21) dated May 23, 1980 wita regard to environmental qualification of Class IE equipment in use at your power reacto.r facility (ies) with an operating license.
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Should you nave questions regarding this Supplement please contact this office.
q Sincerely, hTh~;&lN-R. H. Engelken 1
Director
Enclosures:
1.
IE Supplement ilo. 2 to Bulletin flo.79-01B 2.
Recently Issued IE Bullentins cc w/o enclosures:
i C. P. Yundt, PGE' F. C. Gaidos, PGE J. W. Lentsch i
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6820 Accession fio.: 3003220241 IEB 79-01B, Supple. 2 UilITED STATES MUCLEAP OEGULATORY CCMMISSION OFFICE OF INSPECTION Af!D Et;FORCEMEflT UASHINGT0!!, D.C.
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lfn et M 32 Sectember 29, 1980 IE Supplerrent tio. 2 to Bulletin 79-01E: ENVIR0ilMEtlTAL QUALIFICATION OF CLASS 1E EQUIPMENT Enclosed are the generic questions and answers which resulted from fiRC/ Licensee meetings in NRC Regional Offices during the week of July 14, 1980 regarding environmental qualification of Class IE eouicment in use at power reactor facilities. These answers address soecific cuestions asked during the meetings.
Due to the generic nature of some of these cuestions, the staff is issuing them as a bulletin supplerent.
The reoional meetinos highlighted the fact that in some cases, the scope and depth of the 79-01B review was not clear to licensees.
Therefore, tnese answers may ariect your 79-018 submittal. Thesa submittals are requirea by a separate order to be ccmoleted by November 1, 1980.
Some answers given in Supplem.'nt No. 1 to IES-79-01B are superseced by these answers.
For exancie, in Bulle in 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, cuo to the extension of the response date from April 14,1930 +o liovember 1,1930s "lic equipment is now being addressed since its instaliation is either complete or required before the issuance of the February 1, ZS1 SER.
(See Ouestion 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 IED 79 01B.
IE Bulletin I!o.79-01B was issued under a blanket GA0 clearance (B180225 (R0072); clearance expired July 31,1930) specifically for identified generic problems.
Supplerent Ib. 2 to Bulletin 79-01B is for information, hence no GA0 clearance is required.
Enclosures:
1.
Generic Ouestions and Answers to IES-79-01B and l'enorandum and Order (CLI-30-21) dated May 23, 1930
9 IEB 79-01B, Supple. 2 September 29, 1980 RECENTLY ISSUED IE BULLETINS Bulletin No.
Subject Date Issued Issued To 80-22 Automation Industrit_.
9/11/80 All radiography Model 200-520-003 sealed-licensees source connectors 79-26 Boron loss from BWR 8/29/80 All BWR power Revision 1 control blades facilities with an OL 80-20 Failures of Westinghouse 7/31/S0-To each nuclear Type U-2 Sprino Return power facility in to Meutral Control Switches your region having an OL or a CP 80-19 Failures of Mercury-7/31/30 All nuclear power Wetted Matrix Relays in facilities having Reactor Protective Systems either an OL or a CP of Operating Nuclear Power Plants Designed by Combus-tion Engineering 80-13 Maintenance of Adequate 7/24/80 All PWR power reactor Minicum 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 holdirig 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 at a BWR 80-16 Potential Misapplication of 6/27/80 All Power Reactor Posemount Inc., Models 1151 Facilities with an and 1152 Pressure Transmitters OL or a CP with Either "A" or "D" Output Codes
O. aux l Q.1 Define the scope of review with respect to the June 1982 deadline.
Uhat is reauired beycnd the June 1982 date for qualification?
A.I By June 30, 1932, all safety-related electrical equipment potentially excosed to a harsh environment in nuclear generatina stations, licensed to ocerate en or before June 30, 1982, shall be cualified to either the 00R guidelines or fluREG-0588 (as aoplicable).
Safety-related electrical ecuipment 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 equipnent to function 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 aucit.
Qualification to assure functiening in mild environments must be concleted by June 30, 1982.
The cualification schedules for consideration of the dynamic loading of safety-related eauipment (electrical and nechanical) 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.
I Q.2 Clarify the required submittal dates for ors, itTOLs, and cps. What about OLs wnose 100% license is not expected by June 1982?
l A.2 The reauired schecule for submitting information in response to the Comission Order and i'enorandum (CLI-CC-21) is provided below.
Plants who have received an operating license, either for full or limited oower operation, are required to meet the schedule for operating reactors.
Plants who have committed, to the I;RC, to meet schedules in advance of those provided below are required tc reet that comi tment.
In all cases, plants are required to have their cauipment 'ully qua'ified to the applicable standards either bv i
Jur-30, 1932, or by the time the operating license is granted, whichever ccmes latec i
i Operatino Peactors anc ilTOL (operating license expected by February 1,
)
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Subnittal to be received no later than i:ovember 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 af ter June 30,1982)
Submittal to be received no later than 6 months prior to issuance of operating license.
. 0.3 Define the requirements and applicable criteria for ors, flT0Ls, 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 DDR guidelines?
A.3 Table 1 describes the anplication of each document.
All aperating reactors as of May 23, 1980, will be evaluated against the D0R guidelines.
In cases where the D0R guidelines do not provide sufficient detail, but f!UREG-0588 Category II does, NUREG-0588 will be used.
TABLE I REQUIREMEhTS ors OLs cps D0R GUIDELINES CP SER CP SER Cefore 7/1/74 After 7/1/74 USE NUREG-0588 flVREG-0588(CAT.II) flUREG-0588(CAT.I) flVREG-0588(CAT. I)
AS flECESSARY or REPLACEME!lT COMPONEilTS I!EU RULE WHEN USE NUREG-0588 (CAT.I)
Ifl EFFECT All plants licensed after May 23, 1980, shall conform to flVREG-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 flVREG-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 constructicn 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 11 column of flUREG-0583 and the D0R 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 l
minimum standards set by either of the two documents are equally applicable to ors and flTOLs.
(
Q.4 Clarify the reporting requirements for LERs with respect to Part 50.55e vs79-01B.
1 9
y
. 1 Are only those items, known to be unoualified, immediately reportable?
Are items, for which there are no data or for which there are insuf-3 ficient data, coen itens to be resolved, but are not ir. mediately reportable?
A.4 The reauirerent for recorting in IES79-01B does not change the recortina reauirerents defined in the license conditions.
In general, cps should recort via 50.55e.
Operating plants should use the LER.
i When a determination has been made that reasonable assurance does not exist to ensure that the Class IE electrical ecuipnent component (s) can cerform their safety-relatec function, that is reportable.
Inadeauate or no data are factors in this determination.
The time and tecnnical judaements reauireo to make the determination should i
be based on tne significance of tnis specific eouipment, components, j
and the discrepancies.
Q.5 How coes the "Q" list review interface with the EQB effort? Can the NRC provide nore 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 selecteo.
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 equignent are those required to help bring the plant to cold shutdown and to nitigate the accident (LCCA, HELB inside or outside containment).
i "Miticate" includes safety-related functions such as containment isolation, and prevention of significant release of radioactive raterial.
In order to " pick out" the safety-related equipment, the licensee should cenerate a list of safety functions typically performed by plant safety systens.
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 precedures to perform that function during a LOCA or any llELB inside or outside containment.
If a plant specific safety function not listad in Table II is identified, that function and the corresponding systens or equipment to perforn the functicn should be added to the licensee's list.
The systens and equipment identified above should be included reoardless of the original classification when the plant received its operating license; i.e., scre control grade equipment will probably be named in emergency precedures.
Itowever, if plant emerqency procedures specify a preferred mode of accident mitigation involving equipment recognized by the licensee as unlikely to meet environrental 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
i
. i 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 monitor flow or temperature, etc.
Some of these functions are essential for correct ooerator action, to mitigate accidents, and prevent radioactive releases. When this is the case, the radiation 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 conconents 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, justificiation 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, etc.).
1 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 environ-ment.
Conclusions which show that equipment is unqualified should include a Sasis for continued plant operation.
TABLE II TYPICAL EOUIPMENT/ FUNCTIONS NEEDED FOR MITIGATICU OF A LOCA OR MSLB ACCIDENT Engineered Safeguards Actuation Reactor Protection Containment Isolation Steamline Isolation Main Feedwater Shutdown and Isolation Emergency Power 1
Emergency Core Cooling Containment Heat Removal Containment Fission Product Penoval Containment Combustible Gas Control Auxiliary Feedwater Containment Ventilation Containment Padiation Monitoring Control Room Habitability Systems (e.g., HVAC, Radiation Filters)
Ventilation for Areas Containing Safety Equipment Component Cooling Service Water Emergency Shutdown Post Accident Samolina and Moriitoring Radiation Ponitorina 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 wnicn 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 tystem 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 ccolant 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 systems.
(3) More specific identification of these types of equipment can be found in the plant emergency procadures.
Q.6 NUREG-05C3 was issued for comment.
'!ill 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 MUREG-0588.
No substan-tive changes are anticipated in NUREG-0583 or in referenced daughter standards.
A revision is anticipated, making corrections.
j Q.7 Can IEEE Std. 650(Standards for Oualification of Class IE static battery chargers and invertors for nuclear power generating stations) be used for qualifying the balarce 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 environments.
I l f l
Q.8 Provide the staff's definition of " central location" for qualifica-tion documentation. What documentation is expected to be maintained?
Will it be acceptable to maintain suarary test reports at the utility central file and provide a reference to the NSS3 Vendor's file for the actual test reports? Does NRC require test reports to be sub-4 mitted to support qualification?
4 A.8 The central location should be at tne utilities corporate head-quarters or plant site.
Both the D0R guidelines and NUREG-0588 1
specify that sufficient information must be available to verify that the safety-related electrical equipment has been qualified in S
accordance with the guidance and requirements.
Details for the information and documentation required for type tests, operating i
experience, analysis, and extrapolation of test data frcm operating j
experience are provided in Section 5 of NUREG-0588 and Section 8 of 3
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 sendor's facility.
1 The Licensee / Applicant must make the determination tnat necessary j
infornation and documentation, to support qualification of equipment, i
is in conformance with DOR guidelines and NUREG-0588.
This vendor information file must be maintained current, auditable and available 4
j 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.
l 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 i
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 establishnent and maintenar.ce of a centrally located
]
file which will contain the information necessary to verify the qualification adequacy of all safety-related electrical equipment.
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 facili-ties at different stages (0R, NTOL, CP)?
4 e
. A.11 The qualification information and data should be subnitted with the approoriate officer'; notarized sworn statements.
The format for the data should be in accordance with the format provided in I&E Bulletin 79-018 or the letters provided to the plants in the SEP progran.
Either format is acceotable.
Q.12 Is testing required of ecuipment 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 recuire that the nuclear instrumentation and its associated comoonents be enviror. mentally cualified for a LOCA or HELO.
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 peri" of at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in excess of the time assumed in the accident analysis. The staff has indicated that time is the most significant factor in terns of the margins required to provide an acceptable confidence level that a safety-related function will be conpleted.
Cur judgnent 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 equipment specifying confor-nance 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?
A.13 Sequential testing requirements are specified in NUREC-0588 and t!.
DOR guidelines.
Licensees must follow the test requirements of '~a 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 nrograms in progress should be evaluated regarding the ability to ccmply 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 4
3-l failure nechrism, when the failure occurred, etc.), may be very diffi: ult to achieve.
If such a failure occurs it may be more prudeat 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 Eulletin 79-010, 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 spectrun of breaks including main steam and feedwater line breaks.
The parameters to be considered are:
temperature, pressure, humidity, caustic spray, l
radiation, duration of exposure, aging and submergence.
Nochanical and flow-induced vibrations and seismic effects will be < onsidered l
sepa ra 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 nay be available from earlier HEL3 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 COR Guidelines and MUREG-0583 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 nininum high temperature conditions in pressure-suppression-type l
containnents, 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 i
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 i
function of time should be based on analyses in the FSAR.
- However, l
these conditions should bound those expected for coolant and steam line breaks inside the containment with due consideration of l
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 inpact of the spray on required equipment should be accounted for.
. The adecuacy of a clant-specific profile is dependent on the assump-tion and design considerations at the time the profiles were developed. The 00R guidelines and NURES-0588 provide guidance and considerations recuired to deternine 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 cuality assurance program, can it be qualified to meet IEEE Std. 323-74 requirements?
A.16 The OA programs should accommodate ar/ ncreased scope due to the new environmental qualification docu: er:ation requirements.
Proce-dures incorporated by the licensee f ir data acquisition should be documented and available for staff rev.ew upon request.
Requirements for QA programs are provided in Par + f 0, Appendix B, of the Code of Federal Regulations.
Part E1, Appendix B of the Code of Jederal Regulations states that the aoplicant/ licensee shall be rest)nsible for the establishment and execution of quality assurance rograms.
Specifically in r
purchasing parts or ccmponents, c is the responsibility of the licensee / applicant to ensure raat the applicable quality assurance procedures for their plant r.e met.
In determining the quali.1 cation status of existing equipment purcnased frcn a vendr., where a QA program did not exist, the utility should consiaer the following:
1.
The comolexity of design, complexity of manufacturing process, and end use.
2.
Past perfornance of vendor.
3.
Past operating history of products, especially similar products, made by vendor.
4 Procedures, equipment, and results of environmental qualifica-tion testing relative to those for other equipment for which a OA 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 I
. memorandum.
Af ter 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." Monavailabili 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 weichinn 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 NUREG-0588 shall be documented by the licensee and records shall be available for audit, upon request by the NRC.
Q.18 DDR 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 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 /505 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 containnent.
NUREG-0578 provides the radiation source term to be used for deter-mining the qualificaticn 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.
NUPEG-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 terms for equipment qualification, the licensee must ensure consid: ation is given to those events which provide the most bounding conditions. The following table summarizes these considerations:
'Outside Containment NUREG-0578 NUREG-0588 (100/50/1 (10/10/0 in RCS) in RCS)
I i
i
. Inside Containrent Larcer of MUREG-0588 NUREG-0588 (100/50/1 (10/10/0 in containment) in RCS) 1 or NUREG-0578 (100/50/1 in RCS) 0.19 Can oarma eouivalents be used rather than beta exposure for radiation qualification?
A.19 Yes.
Gamma eauivalents may be used when consideration of the contri-butions of beta exposure nave been included in accordance with the guidance given in the 00R guidelines and HUREG-0588.
Cobalt 60 is one acceotable gamma radiation source for environmental qualification j
of safety-related equipment.
Cesium 137 may also be used.
Q.20 If a piece of equipment will become submerged after completing its I
required action, nust it be qualified for submergence?
A.20 If the equipment (1) ceets the guiadance and requiremants of the 00R guidelines or i.UREG-05'8 for the LOCA and HELB (small and large i
breaks) accidents and (2) licensees demonstrate that its failure will not adversely affect any safety-related function or mislead the operatcr after submergence, the equipment could be considered exempt from that portion (submergence) of qualification.
Q.21 What qualification is required of Reactor Pressure Vessel internal instrumentation (e.g., thermocouples) and new instruments required as the result of T"I 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 MUREG-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.
0.22 Is qualification "by use" an acceptable method (e.g., CRDM's in BURS)?
A.22 Qua'ification by use has limited application.
Often the equipment has never seen the harsh environment and no conclusions can be drawn as ;o its operability in a harsh environment.
Some qualification
. 4 based on operating experience is a recognized method subject to the requirements of flVREG-0588 and the Guidelines.
Credit can be taken for the retural aging of the equipment and for the location of the equipnent m* ntbor portions of the overall qualification information.
Q.23 How long shot a "long term" equipment be qualified for environmental l
qualification?
A.23 "Lono tern" for the purpose of qualifying equioment for a harsh environnent 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 be provided with the equipment qualification documentation.
1.
The time period over which the equipment is reouired to bring the plant to cold shutdown and to mitigate the consequences of the accident.
l 2.
The ability to change, modify or add equipment during the i
course of the accident or in nitigating its effects which will provide the same safety-related function.
Q.24 Why do we want compcnent surface temperature rather than the bulk environment temperature?
A.24 Temocrature 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..ch 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.
4 n.
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