ML20004D215
| ML20004D215 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 11/06/1980 |
| From: | Finkel A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20004D213 | List: |
| References | |
| IEB-79-01B, IEB-79-1B, NUDOCS 8106080614 | |
| Download: ML20004D215 (32) | |
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ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED 4
ELECTRICAL EQUIPMENT IEB 79-01B TECHNICAL EVALUATION REPORT PILGRIM 1 DOCKET NO. 50-293 DATED:
NOVEMBER 6, 1980 Licensee:
Boston Edison Company Type Reactor:
BWR, General Electric Company Size:
655 MWe 1
1 Prepared by Alan E. Finkel Engineering Support Section Reactor Construction and Engineering Support Branch, RI 18106080 0/V
s Contents Page 1.
Introduction...............................................
1 2.
Background and Discussion.............................,.....
1 2.1 General................................................
1 2.2 On-Site Verfication Inspections........................
1 l
2.3 Evaluation of Licensee's Report........................
1 3.
Ge ne ra l I nfo rma ti on.........................................
2 3.1 Identification of Class IE Electrical Equipment........
2 3.2 Service Conditions.....................................
2 3.3 Qualification Documentation............................
2 4.
Technical Evaluation.......................................
2 4.1 Identification of Safety Related Equipment............
3 4.2 Master List...........................................
4 Ser' ice Conditions...................................
4 4.3 v
4.3.1 Inside Containment LOCA.....................
4 4.3.1.1 Radiation..............................
5 4.3.1.2 Submergence............................
5 4.3.1.3 Chemical Spray.........................
5 4.4 High Energy Line Breaks (HELB)........................
5 i
4.4.1 HELB Inside Containment.....................
5 4.4.2 HELB Outside Containment....................
6 4.4.3 Recirculated Fluids.........................
6 4.5 Margins........./.....................................
6 4.6 Aging.................................................
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l 4.7 Documentation.........................................
8 4.8 Site Verification Inspection.........................
8 4.9 Equipment Data Review.................................
8 4
4.10 Conclusions...........................................
9 i
5.
Licensee Event Reports (LERs).........................
9 6.
References............................................
10 Appendix A, Test Reports and Analysis Lists...........
Appendix B, Equipment Status Lists....................
Appendix C, Licensee's Exceptions.....................
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INTRODUCTION 1.1 General The NRC Office of Inspection and Enforcement (I/E) issued Bulletin 79-01B, " Environmental Qualification of Class IE Equipment" in January 1980.
This bulletin required the licensee to perform a detailed evaluation of the environmental qualification of Class 1E electrical equipment required to function under postulated accident conditions and to submit a report on this action.
This document is a report on the evaluation of the licensee's response to this bulletin.
2.
BACKGROUND AND DISCUSSION 2.1 General The evaluation of the licensee's response was accomplished by performing an on-site inspection of selected class lE equipment and by examining the licensee's report for completeness and technical accuracy.
The licensee's report used in this evaluation is dated October 29, 1980, and therefore, does not include the response to the bulletin supplement which was issued on September 30, 1980 in the form of Generic Questions and Answers.
2.2 On-Site Verification Inspections The on-site inspection, made on selected lE equipment, verified proper installation of equipment, overall interface integrity, l
location with respect to flood level for equipment inside the containment, and manufacturers nameplate data.
The manufacturer and model number from the nameplate data was compared to information given in the Component Evaluation Work Sheets (CES) of the licensee's report.
If any discrepancies were noted between the installed equipment and the corresponding equipment addressed in the licensee's report, they are referenced in Section 4.8 of thir report.
The site inspection is l
documented by report number 50-293/80-24.
2.3 Evaluation of Licensee's Report Each component as addressed on the Comp <inent Evaluation Work Sheets l
(CES) of the licensee's report was examined for completeness and i
accuracy to the criteria given in the bulletin.
This examination assumed qualification documents (analysis, test reports, etc.)
referenced by the licensee in their submittal are acceptable.
The results of this examination are documented in Appedix B.
2 4
3.
General Information 3.1 Identification of Class IE Electrical Equipment The licensee's list'of systems was compared to the systems list issued by the Equipment Qualification Branch (EQB) and discussed in section 4.1 of this report.
It's recognized that there are differences in.nomenclatere of systems because of plant vintage and engineering design, therefore many of these systems may not exist or have different titles.
These differences will be addressed in the Safety Evaluation Report.(SER) that will be -
prepared ~for this site.
3.2 Service Conditions The service condition accident environmental, HELB/LOCA inside containment and HELB autside containment are indicated or discussed in the licensee's report and are based on the FSAR accident analysis and discussed on section 4.3 of-this report.
3.3 Qualification Documentation Appendix A is a list of documents (test reports, analysis-letters, etc.) used by the licensee in determining the environmental qualification of plant equipment for Pilgrim Unit #1.
These references have been tab,ulated by the licensee and are indicated on the applicable CES of their report.
4.
Technical Evaluation The basis for the technical evaluation is the infoymation provided by the licensee, Boston Edison Company, for:the Pilgrim l' and the inspection of the as-installed equipment of the Main Steam ADS system which is located l
l in the containment (IE Inspection Report 50-293/80-24).
l l
Utilizingtheinformationidentifjedabov9,0588,andthesupplementsthestaffassessed i
in relation to the D0R guidelines, NUREG to' IEB 79-01B which provides the Commission's requirements and staff positions.
The quality control measures. utilized by the licensee included using l
experienced consultants to perform the tasks required by IEB 79-01B.
Independent technical overview of each part of the effort was performed by the licensee's engineering staff.
In addition, an extensive review of l
the final response and sign-off approvals by various levels of the licensee's engineering management was required.
l l
i 3
4.1 Identification of Safety-Related Equipment The licensee reviewed his documentation to establish the systems required to achieve a safe shutdown or provide isolation for the events identified in IEB 79-018.
These systems were then evaluated against the D0R guidelines.
The systems identified and included in this evaluation are:
1.
Main Steam Line Isolation Valves 2.
Control Rod Drive System 3.
RHRS LPCI Mode Torus Cooling Mode 4.
HPCIS 5.
Automatic Depressurization System 6.
Core Spray System 7.
Primary Containn.ent and Reactor Vessel Isolation 9.
Standby AC Power System 10.
DC Power System 11.
Standby Gas Treatment System 12.
Incident Detection Circuitry 13.
Reactor Building Closed Cooling Water 14.
Salt Service Water 15.
Main Control Room Environmental Control 16.
Reactor Building Isolation Control 17.,
Torus Water Temperature and Level Indication 18.
Equipment Area Cooling System The list of systems including those that were excluded was provided to the Environmental Qualification Branch (EQB).
The EQB compared the list to a "Q" list developed by the staff and to the lists provided by similar facilities to determine the completeness of the licensees response.
Based on the information provided by the licensee and the comparison,
it has been determined that the systems identified are within the guidance provided in Section 3.0 and Appendix A of the D0R Guidelines and are acceptable with these exceptions:
1.
Post Accident Sampling and Monitoring System This system is being reviewed to comply with NUREG-0578 and was excluded per the staff's direction in the first supplement to IEB 79-018.
However, in the second supplement to IEB 79-018 the scope was expanded to include the environmental qualification of the electrical equipment being evaluated in accordance to NUREG-0578.
We, therefore, require that the electrical equipment be identified and the information required by IEB 79-018 be provided.
Reference Appendix C.
l
4 2.
Radiation Monitoring System The basis for excluding this system is the same as item 1, therefore, we require the same.
3.
"Q" List The acceptability of the licensee's list in paragraph 4.1 will l
be evaluated by the Envircnmental Qualification Branch (EQB) and addressed in the Safety Evaluation Report (SER) to be issued by February 1981.
4.2 Master List The licensee developed a master list based on his system evaluation
(
as required by IEB 79-01B. of the licensees 90 day response includes a list of references which provided the basis for l
including or excluding specific components / equipment from having a
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detailed data work sheet as required by IEB 79-018.
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We have reviewed the supporting basis for the inclusion or exclusion of equipment provided in the references and have concluded that Reference event 39, event 41 item 7, item 7.1.a, and 7.1.b are l
acceptable.
4.3 Service Condition 4.3.1 Inside Containment LOCA The licensee provided temperature and pressure profiles for the Pilgrim 1 containment resulting from a LOCA.
These curves, FSAR Figures 14.0, 31 and 32 are included in the licensees 90 Day Response Report.
The maximum environments identified are:
Temperature:
290 F Pressure:
44 PSIG Humidity:
100% R.H.
Chemical Spray:
NA Radiation:
4 x 10 Rads The delay time Vr65 the~ event to the initiation of safety injection for the spectrum of breaks is indicated in FSAR ammendment 34 and profile la and Ib of IEB 79-018 90 day submittal.
The licensee stated in ammendment 34 of the FSAR that the service conditions in the containment will return to the levels that existed prior to the event in less than 30 minutes.
5 4.3.1.1 Radiation 6
The 4 x 10 Radyidentifiedbythelicenseeisless than the 2 x 10 Rads identified, as acceptsble in the DOR guidelines, Section 4.1.2.
The licensee performed calculations using the methods described in TID-14844 and presented in FSAR Table 14.0.19.
The radiation results that were calculated by the licensee areprovidedinservicgprofiles1c,2a,and2bof the licensee submittal.
The licensee is re-evaluating LOCA Radiation Levels outside containment using NUREG-0578.
We have concluded that the above information is unacceptable.
The staff's position in relation to radiation analysis is provided in the second supplament4 to IEB 79-018.
In addition, the supplement expanded the scape to include the environmental effects on electrical equipment being evaluated in accordance to NUREG-0578.
Reference Appendix C, 4.3.1.2 Submergence The licensee 'centified the only area with the potential for submerged equipment post - LOCA is the interior of the torus.
No safety related electrical equipment exists i,n the ares.
4.3.1.3
^hemical Spray The licensee stated that no chemical solutions are used in systems required for the accidents presently under consideration.
4.3.1.4 Accuracy The licensee's answer is unacceptable.
Reference Appendix C for their statement on accuracy.
4.4 High Energy Line Breaks (HELB) 4.4.1 HELB Inside Containment The Pilgrim 1 facilty in their FSAR Appendix G for Event 39 - Pipe Break Inside Primary Containment up to and including a DBE-LOCA listed the systems that are required to function during this event.
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4.4.2 HELB Outside Containment The licensee indicated in their iSAR Appendix G event 41, Pipe Break Outside Primary Containment the systems that are required to function during this event.
The licensee analysis, event 41, included the following co.7siderations and actions:
1.
Protective enclosures 2.
Electrical equipment arrangement 3.
Qualified equipment in the hostile environment 4.
Consiceration of loss of air conditioning and loss of offsite power.
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We have concluded, based on the considerations, information and actions taken, that the licensee meets the requirements of the DOR guidelines, Section 4.3.1.
The acceptability of the licensees basis for specific equipment ~ s0bjected to HELBs outside of containment is included in Section 4.9 of this report.
4.4.3 RecirculatedFilids g.
The licensee indicated that the hostile environments, event 39 evaluation, in the Reactor Building' areas containing post LOCA recirculation flow have been reviewed i
using the integrated dose for the surface of a 24 inch Schedule 80 pipe.(service pgofile 2a) and that the radiation is no greater than 7.1 x 10 Rads.
l The acceptability of the parameters identified-and the basis for specific equipment qualification are included in Section 4.9 of this report.
In addition, the value for the radiation environment may be modified by the licensee i
l in re evaluating LOCA Rad levels in response to NUREG-0578.
Appendix C states the licensee position on the study they have completed to date.
4.5 Margins The DOR Guidelines indicate that special consideration was given to the time required to remain functional when establishing the criteria in Section 5.2 of the guidelines.
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7 120,F and the profiles indicate that the temperature returns toT 120 F within 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> of the event.
NUREG-0588, Section 3(4),
requires that a type test be for a minimum of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in duration when the functional requirement is within the. first seconds or minutes of an event and the DOR guidelines, Section 5.2, requires 3
that the test duration be at least as long as the period from initiation j
until the service conditions return to the level that existed prior to the event.
Therefore, any type test that exceeds the functional operebility time by 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or longer meets the requirements defined in NUREG-0588 and the DOR guidelines for margin in relation to test duration Tor j
this facility.
The other consideration identified in the DOR guidelines in relation to the methods of qualification, other than identified specifically l
in this report will ba addressed in the Safety Evaluation Report (SER) which will incorporate an audit of selected analysis and test reports identifieo in Appendix A.
4.6 Aging i
The licensee indicated that an ag 4g study of the components subjected to harsh environments is still ar tstandingitgm.
Details of the licensee's effort is included in sir submittal, Print Out 2II.
The' licensee has identified the components which are still listed as i
l requiring data.
The DOR guidelines, Section 7, does not require a qualified life be established for all safety related electrical equipment, however, the following actions are required:
I 1.
Detailed comparison of existing equipment to the materials identified in Appendix C of the D0R guidelines.
The first supplement 4 to IEB 79-01B requires the licensees to utilize the 1
table and identify any aMitional materials as the result of their effort.
2.
Establish an ongoing program to review surveillance and maintenance records to identify potential age related degradations.
3.
Establish component maintenance and replacement schedules which include considerations of aging characteristics of the installed components.
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8 Ve, therefore, require that the licensee provide the details of a program which will include a continuing effort to obtAin data on existing materials and address the actions identified above.
In addition, we require the licensee provide a schedule for implementation of the program that identifies problem components.
4.7 Documentation 4
The second supplement to IEB 79-01B and the order,5 No. CLI-80-21, requires the licensee have the 60cumentation and data identified in the detailed worksheets which supports the qualification of the safety related elgettical equ % ment available for:NRC: audit.
The second supplement identifies the type of information required and -
the locations where the records are to be maintained.
The staff requests the licensee provide a response to the order and-supplement which discusses their compliance and identifies.any deviations.
- 4. 8 Site Verification Inspection An inspection of the installed compcnents associated with the Main Steam ADS system was conducted on April 9-10, 1980 at the Pilgrim I facility.
The details of this irapection are included.in IE inspection report 50-293/80-24.
The Jetailed identification of the components and the observations i
rec'orded will be addressed in the SER which will incorporate an I
audit of selected analysis and test reports identified in Appendix A.
4.9 Equipment Data Review The equipment listed in Appendix B is the status of the latest data submittedbythelicenseeintgeirresponsetoIEB79-018.
Appendix B identifies the licensee data in a format that allows-the reviewer a quick look status of each listed component.
The first four columns are self explanatory while the next three columns are defined as follows:
Enviror, ment l
l The listing in this column identifies the environment that appears to have some question as.to whether or not it's in compliance with the requirements of the licensee.
Category As listed below a category I through V has been assigned to the environment for a specific component or group of components as l
listed.
l
9 l
Remarks l
The remarks column was used to identify the environmental condition I
associated with the category number, or identify the system location when the licensee indicated that data was being looked for or an analysis was in progress.
An example of this lack of data environment information in the licensee submittal is the requirement for aging.
The equipment has been listed and identified in one of the following categories:
I Qualified for Plant Life II Qualified With Restrictions III Exempted From Qualification IV Qualification of Equipment Unresolved, and L
V Equipment Not Qualified The number in the ( ) in the component block on the table indicates the number of identical components listed, but may have a different title within the report.
4.10 Conclusion This evaluation is based on the on-site insgection, the information supplied by the licensee in their submittal, their FSAR, and the l
assumption that the Qualification Documentation (Test Reports, Ana,1ysis, Letters, etc.) are acceptable.
9 8
The Region I reviewer using the guidance and inst.ruction for the evaluation of licensees data submittals and the site verification inspection that was performed to verify the IE Bulletin 79-01B, l
January 1980datasubmittalinforgion,findsthelicenseetobein l
accordance with the NRC direction except as listed in Appendix B l
and C of this report.
The results of this evaluation does not necessary imply that the equipment is unreliable, unsafe or represents a significant safety l
issue; it does imply that additional information is required and that the items in Appendix B and C will be evaluated by the Equipment Qualification Branch (EQB) and addressed in the Safety Evaluation Report (SER) to be written for this licensee by February, 1981.
5.
Licensee Event Reports (LERs)
No licensee event reports were submitted by the licensee, associated with their evaluation of IEB 79-01B, as of October 29, 1980.
1D 6.
References 1.
IEB 79-018, Memo to V. Thomas (NRC) from A. Finkel (NRC) dated August 18, 1980.
2.
EQ Branch Comparison of systems and parameters.
Boston Edison Coapany, Revised and Updated Response to IEB 79-01B, dated October 29, 1980.
4.
Supplement Information to IEB 79-01B, dated February 29, 1980, and September 30, 1980, and October 24, 1980.
5.
Order requiring licensees implement requirements of Commission Menorandum and Order of May 23, 1980 (CLI-80-21).
6.
Division of Operating Reactors (DOR), " Guidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors", Enclosure 4 to IEB 79-018.
7.
NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety Related Electrical Equipment", dated December 1979.
8.
Inspection Requirements for Verifying Reactor Licensee Responses to IE Bulletin No. 79-018, dated Spril 25, 1980.
9.
IE ' Support and Review of Environmental Qualification of Electrical Equipment at Operating Reactors, dated October 10, 1980.
10.
Boston Edison Company, Responses to IEB 79-01B, dated March 4, 1980, March 12, 1980, April 18, 1980, April 22, 1980 July 22, 1980, and October 29, 1980.
~5E
App =ndix A Test Reports and Analysis References l
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1.
Kerite Letter to Licensee 7-21-80 and certified test report summaries 2.
OKonite engineering report 127, 10-24-71 and letter 4-9-80 3.
GE letter G-HK-9-44, and G-HK-9-72 6-5-79 4.
Rockwell report 2792-03-02 5.
Namco test report EA'/40 2-20-78 6.
Atwood and Morrill test report STR-060578-1 7.
Limitorque report 600376A and letter 2-17-79 (Memo 80-184 aging) 8.
Limitorque letter to licensee 4-27-79, TWX 5-17-79 and 5-29-79 Test report 600198 1-2-69 9.
EPAG-060, 009, 010, 046 10.
SPSD-QC-206 11.
EPAG-055, 007, and EPAG-047/AEPAG-3 12.
Conax test report IPS-42 Rev. A 13.
ASCO test report AGS21678/TR 14.
GE letter G-HK-9-123 9-23-79 15.
OKonite letter to licensee 6-4-79 and 7-9-79 16.
FIRL F-C3781-1 and FIRL F-3781-2 17.
P&CS Memo to File #80-12 18.
P&CS Memo to File #80-125 re' 19.
BPCo SCEG 20.
GE test report 145C3008, P&CS Memo 80-202 (Aging),'and Memo 80-238 (RAD) 21.
GE test report 145C3009 22.
GE test report 145C3007 23.
Terry test report E/L20397, Rev. 4, and Terry letter to licensee 11-29-79 24.
GE Test report 145C3031 25.
Memo P&CS80-109 26.
Limitorq'ue test report B0003, 80027 and letters 4-27-79, 4-30-79, and 6-11-79.
27.
FIRL F-C3271 (Break Only) 28.
Limitorque test report B0009 29.
GE test report 145C3012, 145C3011, and 145C3010 30.
Wyle test report 43854-1 31.
P&CS Memo 80-238 (RAD) 32.
P&Cs Memo 80-186, August 18, 1980 entitled " Efforts of PBOC Short-term Elevative Ambient Temperatures on the Operating Termperatures of Various Electrical Equipment"
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Appendix B Equipment Status List o
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The following coding is used by the licensee to identify the status of the components listed in printout section 2II of the October 29, 1980 IE Bulletin No.79-01B Final Response.
Components which have been extrated from the above report and listed in this Appendix are coded in accordance with the requirements of section 4.9 of this report.
Licensee Codes:
l QUAL STATUS - This field lists the results of BEco.'s qualification evaluation.
l Acronymic designations were used to facilitate future work.
Possible designations and their respective meanings are as follows:
NYI - Not Yet Installed JC0 - Justified for Continued Operation QPS - Qualified to Plant Speci:ic Requirements DOR-A - Qualified to 00R Guidelines except for aging DDR - Qualified to 00R Guidelines l
UQ - Unqualified (Reference LER)
NYQ - Not Yet Qualified OUAL PLAN - T,his field lists future qualification efforts and forecast completion dates.
Again, acronymic designations were used as follows:
FT - Future Qualification Test RA - Radiation Analysis AA - Aging Analysis HT - Heat Transfer Analysis l
ER - Equipnent Replacement II - Installation Inspection FE - Final Evaluation l
I Cumponent Kiraa f.
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,,3g la teriory Remarks E our RIR (fnit Cooler (4)
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QUAL PLAff - AA-9/81. RA-9/81. & FC-9/81. QJAL STAIUS - JCO.
HotorOperator(1)
L imitorque
$ 10 2 m
Radiation IV Peericss Itw21226 Peericas flotor - QUAL PLAN - RA-l/al, and IT-1/al.
Hotor Operator (4)
Limiturque SHU 0 m
Radiation IV Reliance Motor Rellarge 441101 441101-CY - INAl PIMI - RA-9/SI, and TC-9/01.
DutletDamper(2)
Acilonator it94mlo6 m
Radiation IV Standby r.as Ire.itment - 1)tlAL 11 AN - FR-II/01.
A ing IV 9
Pressure Switch (13)
Static-o-LN AAl a
Heat Irans IV Ring QUAL PLAN - Ili-4/01, and FE-4/81 Pecssure Switch (2) 3arts Dale D2T-Al25!
iteat Irans IV a
f)tlAl PIAN - lli-7/81 dpd FE-l/81 Pressurcswitch(4) ilarks Dale B2T-Al25!-
a lleat Trans IV 94L PLAN - 111 4/81, and FE-4/81.
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Invirors-el$rfal e2 t t.en t IEfIgor 8C28t it y
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l3 4AL-x Agin9 IV llAVr 'lCCS Un'il 'Coolcr's' - QUAL PI'AN. AA-7/8ll and STAlus 00R-A.
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,I Penetration Conan Y
The Test curve #5 did.v t s:eet the 00R Guidelines for temperature umf time.
Penetration Physical l
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ilie Irst tuive 86 did viot meet the IX1R Guidelines for temperature aim) time.
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grial men t ment ategory Remasks lia ouT Piessure XMIR (4)
Roseemnt 115?
m IV PHIT TEI Ild517.ttID - Requf re 5thedule Fran Licensee.
SolenoidValve(2)
ASCO flVA 90-405 x
IV Ptt Program /CE SIL128 Replaces All age Sensitive Casponents.
IJtlAt PLAN - t[-6/01.
SolenoldValve(14)
ASCO Hill 320A x
Radiation IV 22 Io le e cplat a.1 prior to 6-111 QtlAL PL Aff - IR-il/81.
Solenoid Valve (2)
ASCO WP-18-83tfi a
Radiatinn IV (i.'AL PLAN - AA-4/81, RA-4/81.111-9/191 anf II-9/al.
J6 Aging IV lleat Trans IV TemperatureElement(2)
Electric 35-4 22-fi-Aging ly Existing Ril?s in be replaced - QlfAL Pl.AN [R-12/00 - QUAL 51Arus.ro.
Themo-139 meter themnstat (7)
Johnsore x
Aging IV Controls in he replaced l>y 11/30 HVAC TLCS Unit Coolers QUAL PLAN -
(R-2/81; QUAL SIAIUS - N7Q.
Solennld Yalve (4)
AVC0 C5159 Aging IV QUAL fl Aft Steane Tunnel tocation AA - Test scheduled 2/81 and TE-procedure 4/01 for (1im Sw) and (SV).
Limit Switch (4) flAllC0 IA140 Agi g IV Switches (6)
Elec t:c-24-40 W-x Radiation IV QUAL PL AN - RA-2/81, llT-2/01, AA-2/81 and FE-J/al.
Switch Lights (6)
GE ET-16 Heat Trans IV s
T1rminalBlocks(6)
GE EB-25 Aging IV m
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Appendix C Licensee's Exceptions i
e l
l i
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The licensee has identified the following as "Outstancing Items", and " Generic Evaluations".
Included in this appendix is Hostile Environments, Events 39 and 41.
These are direct quotes from the licensee's letter that was part of the IEB 79-018 final report.
The reference in this quote to reports and analysis are not included in this TER.
Outstanding Items Radiation The NRC, in response to Question 9 of Supplement 1 to 79-018, advised that it considered NUREG-0578 and NUREG-0588 to be consistent in their requirements for radiation evaluations.
Based on this guidance, BEco used the TID-14844 source terms and distribution methodology presented in Section 14 of Pilgrim's FSAR.
This approach is considered to be consistent with the requirements of NUREG-0588, Section 1.4 (1) and Appendix D for DBA-LOCA.
Supplement 2, Question 18 requested new evaluations assuming all source term inventories remain in the coolant.
Boston Edison has initiated radiation analyses but has not yet incorported this requirement into its equipment qualification evaluations.
If revised radiation dose values result from the analyses, equipment qualification evaluations will be conducted to the revised values.
Our conclusions are not expected to change as a result of this work because of the conservatism already included in our evaluations.
For example, we have used the contact dose for a 24 inch pipe containing post-loca fluid for radiation evaluations outside containment.
Generic Evaluations Boston Edison has developed, for equipment located outside containment, technical papers addressing the ability of electrical equipment to function in spite of aging, radiation and short-term thermal transients.
These papers, discussed in detail in the following section, have been used in the performance of generic qualification evaluations.
In some cases, Boston Edison determined that before a final equipment qualification evaluation could be made, verification of the applicability of these generic evaluations was required.
For each l
item, this determination was reflected in the computer print-outs for evaluations.
The component was classified as being justified for continued operation (JCO) or as meeting D0R guidelines except for aging (DOR-A).
The outstanding verification effort (eg-Radiation Analysis (RA) and/or Aging Analysis (AA)) was listed under the Qualification Plan section along with its forecast completion date.
Justification for Continued Operation As the attached computerized summary reports reflects, the qualification of many items has been found to meet the D0R guidelines.
For those items requiring additional effort, discrete tasks and target completion dates have been established and reflected in the qualification plan for each item.
In the interim, generic qualification evaluations support continued operation for many items.
Boston Edison has prepared technical papers a,ddressing the ability of electrical
2 equipment to function in spite of aging, radiation, and short-term thermal transients.
P&CS memo 80-257 entitled " Simplified Methodology for Determining Significance of In-Service Aging and its Effects on Equipment Performance,"
presents a method to determine if thermal aging is a significant failure mechanism for safety related electrical equipment.
Primarily, it uses Arrhenius methodology to infer increased life factors from manufacturer's information on expected life at some temperature, when, the maximum continuous operating temperature is lower.
P&CS Memo 80-186 dated August 18, 1980 entitled " Effects of PB0C Short-term Elevation Ambient Temperatures on the Operating Temperatures of Various Electrical Equipment," justified the use of device specific heat transfer analyses to demonstrate the ability of electrical components to function under the subject thermal effects.
Analytical and empirical data were used to demonstrate that a component's thermal capacitance precluded it from experiencing the ecmpartment maximum ambient temperatures.
Based on this, specific heat transfer analyses will be performed on those components whose test environment temperature profile does not envelope the compartment ambient temperature profile.
P&CS Memo 80-238 entitled " Radiation Effects on Organic compounds used in Safety Related Electrical Equipment Located Ouside a BWR Primary Containment,"
establishes the conclusion that for outside containment equipment most of the organic materials, with a few exceptions will perform adequately at integrated doses of 1.0 Megarads gamma and lower.
Radiation effects on materials properties are identified and discussed.
Empirical results from a variety of tests and studies supporting the conclusion are provided as attachments.
From the abov'e discussion it is apparent that although verification of the applicability of the generic evaluations continues to be an outstanding item for some components, each component has been carefully reviewed and it is our judgement that the components will ultimately be found to be qualified.
With respect to the outstanding radiation analyses dictated by Supplement 2, it has l
been demonstrated both in the preceding section and in the Hostile Environment section, that the evaluation conclusions should not be affected by this effort.
Accuracy Accuracy information is applicable to instruments only.
No instruments in containment are under the scope of 79-01B (none are required for LOCA).
Instruments outside containment are not exposed to the LOCA temperature, pressure effects.
Some instruments outside containment do experience the effects of PBOC, principally directly related to instrument internal temperatures not ambient temperatures.
For PB0C such shifts are short-term in nature and are significantly less than the shifts due to steady-state elevated temperatures.
Test accuracies, if available, are presented on the appropriate Test Curve sheet.
l l
l
~,.
3 Hostile Environments - EVENT 39 "For Event 39 the inside containment environmental profiles developed in the FSAR considered 2 limiting cases.
- 1) Environment ~due to DBE-LOCA and 2)
Environments due to the envelope of smaller size breaks.
The Drywe: 1 Pressure, Temperature, time curves for both cases are shown as Service Profiles (enclosed) la) (Reference FSAR Figures 14.0-31 & 32) and ib) (Reference FSAR Amendment 20 Response to comment 5.2.1 Figures 5.2.1.2 thru 5.2.1.4).
Gamma radiation exposure due to the DBE-LOCA using the assumptions given in TID-14844 were developed and are presented as FSAR Table 14.0.19 - enclosed.
The fission product source terms postulated in TID-14844 are conservative considering that the core standby cooling systems are designed to protect against such gross fission product releases.
Equipment capabilities were reviewed using these TID-14844 gamma dose levels.
For cable, wire, and splices not contained within protective enclosures evaluations were conducted assuming a total 8
integrated dose (beta and gamma) of 2x10 Rads.
No hostile environments (temperature, pressure, himidity) due to DBE-LOCA inside containment are experienced by safety related electrical equipment outside primary containment.
Radiation levels due to LOCA at electrical equipment locations in the Reactor Building are significantly below the threshold i
l values for most plastics, elastomers and insulating compounds.
The integrated TID-14844dosesforacompartmentcontaininggcorespraypumpandassociated piping are given'in Table 14.0.19 as 7.1 x 10 Rads.
Based on these considerations, l
Reactor Building areas would not experience hostile environment due to Event j
39.
For the purposes of this evaluation equipment located in Reactor Building areas contain,ing post-LOCA recirculation flow have been reviewed using the integrated contact dose for the surface cf a 24 inch Schedule 80 pipe (Service Profile 2a).
Unless otherwise indicated, the considered " Required" dose included radiation received during normal operation.
In response to NUREG-0578 Boston Edison is re-evaluating LOCA Radiation levels outside containment.
If the results of this work indicate that the Table 14.0.19 levels are inappropriate, evaluations under 79-01B will be conducted to the revised levels."
l Hostile Environments - EVENT 41 s
"The environmental effects of PBOC are considerably less severe than those generated by DBE-LOCA.
The principal effects involve high humidity and short-term elevated temperatures.
Radiation effects to electrical equipment are insignificant as are the effects of the short-term pressurizations.
For Event 41, no abnormal environments are experienced inside containment or in the Control Room areas.
Effects on other plant areas were formally investigated and presented as Amendment 34 to Pilgrim #1 FSAR.
The analysis as developed in that section was primarily concerned with structural and piping system capabilities and the effects of compartment pressurization, jet impingement and pipe whip.
The temperature profiles developed by the analysis were extremely conservative particularly for areas removed from the actual breaks.
4 Those pipe breaks outside containment which create hostile environments are breaks in the Main Steam lines, HPCI Turbins Steam Lines, RCIC Turbine Steam Lines and RWCU System Piping.
Closure of the MSIV's due to a steam line break will generate a reactor trip.
For breaks in the main steam system termination of blowdown will occur within 5.5 seconds.
The maximum duration of blowdown for all other PB0C cases considered is less than 26 seconds.
The arrangement of electrical equipment is such that, in gereral, only electrical equipment associated with the system within which the break occurs is located in break compartments.
Because the affected system is disabled due to the hypothesized pipe break this equipment need not function.
For the other plant areas affected by these breaks the abnormal environments are of shorter duration and substantially lower termperatures.
The short term abnormal environments experienced in these less affected areas are considered no more severe than those found in many areas of conventional power plants and industrial plants where similar devices have satisfactorily performed.
The inherent ability of these devices to tolerate short term transients can be established by recognizing that their active intervals are insulated from such short term effects by their enclosures and the conduit systems which interface with them.
The capability of enclosures to provide such e level of protection has been well known.
The NRC in NUREG-0458, "Short Term Assessment of the Environmental Qualification of Safety Related Electrical Equipment of SEP Operating Reactors", and in various licensing submittals has recognized this capability and acceptably reviewed equipment qualifications based on models which predicted the devices response to such short time events.
As the nuclear industry's experience in equipment qualifications has progressed, the emphasis on the demonstration of qualification has further shifted to type l
testing.
Since the time, a significant amount of testing has been conducted on devices generically similar to those used at Pilgrim #1.
The successful completion of these tests decanstrated the inherent ability of such devices to function during short term elevated temperaturs and substantiates the engineering analysis previously conducted which predicated their acceptability.
The use of equipment at Pilgrim #1 identical to the type supplied to other BWR's by GE and the specification of generically similar equipment to that used throughout the industry for Balance of Plant equipment insures a capability consistent with other plants.
1 e
- ~, -, - -
,m
c 5
Boston Edison is proceeding with searches and evaluations of recent qualification testing applicable to Pilgrim #1 equipment exposed ot PBOC environments.
When documentation cannot be found which is applicable to the equipment at Pilgrim, scheduled replacement with type tested units will occur.
In addition we have
~
completed developing realistic environmental profiles for all plant areas affected by PBOC.
The profiles are attached.
These PB0C environmental profiles were developed using Bechtel Corporation's latest version of "FLUD".
FLUD is one-dimensional computer code designed to handle problems dealing with gas flows between interconnected compartments.
Basic equation of mass and energy conservation are used along with quasi-steady state flow equations to calculate the transfer of mass and energy among the various compartments that comprise the system.
Long term compartment cooling is achieved by considering heat-transfer into compartment walls using appropriate heat-transfer mechanisms."
a
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s e.
6 Table 14.0.19 DOSE RATES FOR VARIOUS EQUIPMENT OR LOCATIONS BASED ON i
TID-14844 FISSION PRODUCT RELEASE ASSUMPTIONS Location or Max Dose Integrated Dose (Rad) For Equipment Rate (R/Hr) 12 Hours 2 Days 80 Days 180 Days 4
4 5
5 5
Surface 24 inch 1.1 X 10 5.9 X 10 2.0 X 10 4.4 X 10 6.2 X 10 80 pipe 5
4 6
7 7
l Interim Surface 7.8 X 10 3.8 X 10 9.4 X 10 1.8 X 10 2.6 X 10 Drywall I
2 3
3 3
Floor of Corner 2.6 X 10 1.0 X 10 1.0 X 10 8.0 X 10 7.1 X 10 Compartment Containing Core Spray Pump Seals 4
4 5
5 5
Pump Seals 1.1 X 10 5.9 X 10 2.0 X 10 4.4 X 10 6.2 X 10 2
2 3
4 4
Sscondary 1.0 X 10 4.2 X 10 3.8 X 10 1.1 X 10 2.6 X 10 Containment Grd.
2 3
4 4
5 Refueling Floor 4.2 X 10 1.7 X 10 1.6 X 10 4.5 X 10 1.1 X 10 I
I.
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