PLA-4274, Responds to NRC 950203 RAI Re Plant MSIV Leakage Control Sys
| ML18026A452 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 02/21/1995 |
| From: | Byram R PENNSYLVANIA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| PLA-4274, NUDOCS 9502270273 | |
| Download: ML18026A452 (62) | |
Text
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R.IWY'CCELERATEDRIDS PROCESSIi REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9502270273 DOC.DATE: 95/02/21 NOTARIZED: NO DOCKET N
. FACIL:50-387 Susquehanna Steam Electric Station, Unit 1, Pennsylva 05000387 50-388 Susquehanna Steam Electric Station, Unit 2, Pennsylva 05000388 AUTH.NAME AUTHOR AFFILIATXON BYRAM,R.G.
Pennsylvania Power 6 Light Co.
RECIP.NAME RECXPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Responds to NRC 950203 RAI re plant MSIV leakage control sys.
DISTRIBUTION CODE:
A017D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: OR Submittal:
Append J Containment Leak Rate Testing P
NOTES:
05000387 R
RECIPIENT'D CODE/NAME PD1-2 LA POSLUSNY,C XNTERNAL: ACRS NRR/DE/ECGB OGC/HDS2 RES/DSIR/SAXB EXTERNAL: NOAC NOTES:
COPIES LTTR ENCL 1
0 1
1 6
6 1
1 1
1 1
1 1
1 RECIPIENT ID CODE/NAME PD1-2 PD PI-LE CEggE 0
NRR/DSSA/SCS B RES/DE/SEB NRC PDR COPIES LTTR ENCL 1
1 1
1 1
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N iNOTE TO ALL"RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE lV'iSTE! COYTACT I HE DOCI:IIEYTCOYTROL DESK, ROOI~I Pl-37 IEXT. 504-20S3
) TO ELDIIX 'iTL YOI.'R XA!~IL'RO!~I DISTRIBUTIOY,LIS'I'S I:OR DOCI.!i!EX'I'S5'OL'ON "I'l;ED!
TOTAL NUMBER OF COPIES REQUIRED:
LTTR 18 ENCL 17
FEB 21 1995 Pennsylvania Power 8 Light Company Two North Ninth Street ~ Allentown, PA 18101-1179 ~ 610/774-5151 Robert G. Byram Senior Vice President Nuclear 610/774-7502 Fax: 610/774-5019 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Mail Station PI-137 Washington, D.C.
20555 SUSQUEHANNA STEAM ELECTRIC STATION MAINSTEAM ISOLATIONVALVE/LEAKAGE CONTROL SYSTEM: RESPONSE TO REQUEST FOR ADDITIONALINFORMATION Docket Nos. 50-387 and 50-388
Dear Sir:
The primary purpose ofthis letter is to provide the Pennsylvania Power &Light Company response to the NRC Request for Additional Information (RAI) of February 3, 1995.
This RAI resulted from the NRC Staff's evaluation of the radiological analysis portion of the Susquehanna Steam Electric Station (SSES) proposed amendment (PLA 4228, dated November 21, 1994) to increase Main Steam Isolation Valve (MSIV) leakage rate and to delete the MSIV Leakage Control System (LCS).
Enclosure 1 restates the Staff's five specific RAI questions followed by PP&L's response to each.
In addition, during the January 24, 1995 meeting regarding PP&L's proposed amendment, PP&L indicated it would be providing a supplemental submittal to address compliance to Appendix A of 10 CFR 100 and to request continued exemption from Appendix J to 10 CFR 50. Enclosure 2 highlights the MSIV Alternate Treatment System Seismic Margin Evaluation performed by PP&L.
This analysis supports the conclusion that this alternate treatment method will remain functional following a Safe Shutdown Earthquake; thus meeting the seismic criteria of 10 CFR 100, Appendix A.
provides the supporting information for the continued exemption to 10 CFR 50, Appendix J.
Questions regarding this supplemental information should be directed to Mr. J.
M. Kenny at (610) 774-7904.
Very truly yours, 1
R.
yr Enclosures r~pat+c r p 9502270273 95022i PDR ADOCK 05000387 P
PDR FILES A17-2/R41-2 PLA-4274 Document Control Desk CC:
NRC Region I Ms. M. Banerjee, NRC Sr. Resident Inspector - SSES Mr. C.
Poslusny, Jr., Sr. Project Manager - OWFN Mr. W.
P. Dornsife, PA DER
ATTACHMENTTO PLA-4274 Page I of32 ENCLOSURE1 RESPONSE TO REQUEST FOR ADDITIONALINFORMATION
...9502270273
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ATTACHMENTTO PLA-4274 Page2of32 (1)
Confirm the assumptions used to calculate LOCA doses in Table 15-5 of Susquehanna Safety Evaluation Report (NUREG-0776) issued in April 1981 and identify the basis for any changes.
Differences in assumptions shown in Table 15-5 of the SSES SER as supplied in NRC RAI of February 3, 1995 and those used in the power uprate update of the SSES FSAR Section 15.6.5 DBA-LOCA radiological analysis are shown below. Any differences in parameters are described and the SSES FSAR value is presented.
If a parameter shown in Table 15-5 is essentially the same as used in the PPAL DBA-LOCAanalysis, the value is given and the indication (Equivalent) appended.
Power Level:
The current power level used in the SSES DBA-LOCA analysis is 3616 MW(t) which represents 105% of the recently uprated SSES core thermal power. (PPhL Calculation EC-RADN-1009).
Operating Time:
3 years (Equivalent)
Core Fraction Released to Drywell:
100% Noble Gases 25% lodines (Equivalent)
(Equivalent)
Containment Free Volume:
Reactor Building Free Volume:
3.89x10 ft Primary Containment Leak Rate:
1.0%/day
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Reactor Building Leak Rate:
100%/day (Equivalent)
(Equivalent)
(Equivalent)
The Reactor Building Free Volume is obtained from SSES Architect-Engineer (Bechtel) calculations of free volumes of ventilation zones I, II, and III, and has a value of 5,755,200 ft
. Individual ventilation zone free volumes are shown in FSAR Table 15.6-22 as:
ZoneI:
1,488,600 ft Zone II:
1,598,600 ft Zone III: 2,668,000 ft Total:
5,755,200 ft3 Bypass Leakage:
5 SCFH Reactor Building Mixing Fraction:
50%
(Equivalent)
(Equivalent)
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Reactor Building Exhaust Maximum Flow Rate:
ATTACHMENTTO PLA-4274 Page 3 of32 The initial flow rate of the SGTS.system is 10,500 cfm which-occurs'during drawdown."As -1/4".-.
. water gauge pressure is achieved in secondary containment. (Reactor Building) at approximately
" 180 seconds, the Reactor Building Exhaust Flow rate flow-drops to 3997 cfm (Zone I+II+III mixing) for the remainder of the accident course.(FSAR Table 15.6-22).
SGTS Filter Efficiencies:
Elemental Organic Particulate 99%
99%
gg (Equivalent)
(Equivalent)
(Equivalent)
Exclusion Area Boundary:
Low Population Zone Distance:
Atmospheric Diffusion (X/Q) values:
3 miles (Equivalent) 1800 feet (SSES FSAR Section 2.1.1.2)
The Five-percentile offsite atmospheric diffusion factors (X/Q) were determined by the SSES Architect-Engineer, and are shown below as obtained from SSES FSAR Table 2.3-119. A full discussion of the determination of these values can be found in SSES FSAR Section 2.3.4.
Time Period 0-2 hours 0-8 hours 8-24 hours 1 -4 days 4-30 days Location Exclusion Area Boundary Low Population Zone Low Population Zone Low Population Zone Low Population Zone X/Q (sec/m3) 9.60x10 2.18x10 2.83x10 1.43x10 1.08x10
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ATTACHMENTTO PLA-4274 Page 4 of32 (2)
Provide the following additional information and FSAR or other references where the information can be found as appropriate:
- Main Steam Line (MSIVto Main Steam Drain Line):
Inside Diameter Length'nsulation Thickness'ipe Thickness'ooling Rate(or provide density, heat capacity, and thermal conductivity of steel pipe)
- Drain Line (Drain line to Main Condenser):
Inside Diameter'ength'nsulation Thickness Pipe Thickness'ooling Rate(or provide density, heat capacity, and thermal conductivity of steel pipe)
See attached sketches Outboard MSIVto Condenser (Drain Line Pathway)'nit 1
& 2 (from PP&L calculation EC-083-0512, MSIVILCS Radiological Dose Calculation Data Sheet-Data Verification )
See Main Steam Line and Drain Temperature Profile charts (charts provided by CEIBWROC)
- Effective Condenser Volume (above drain line):
Effective condenser volume (above drain line) value used in the radiological calculation 98601 ft
'his includes the free air space volume of the LP Turbine..
The actual condenser free volume (above drain line) including the free air space volume of the LP Turbine - 117371 ft
'hus, the value used in the radiological calculation is conservative.
pdEg +etrrta 9999 vg +)
Cat 9915iOI Dept.
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- Control Room Parameters:
The following data are drawn from the SSES FSAR and PP8 L Calculation EC-RADN-1009 which updates the Design Basis LOCA Analysis for the Power Uprate Project. Calculation EC-RADN-1009 updates DBA-LOCA doses for SSES FSAR Section 15.6.5 and supersedes the DBA-LOCA analysis performed by GE.for power uprated conditions. This calculation uses ICRP 30 dose conversion factors and the TACT5 computer program which has been validated and verified for design use at SSES.
Use of ICRP 30 dose conversion factors is consistent with the BWROC methodology used to calculate doses for SSES due to the Increased MSIV Leakage/LCS Elimination analysis. EC-RADN-1009 is now the calculation of record for the SSES DBA-LOCA.
Where original FSAR parameters have been carried through into the EC-RADN-1009 analysis unchanged, the original FSAR reference is given.
Parameter Value Reference Remarks Control Structure Free Volume 510,000 ft3 Control Room Free Volume 11p ppp ft3 FSAR 6.4.2.3 FSAR 6.4.2.3 Control Structure Volume Murphy-Cam pe Gamma Volume 1
Emergency Makeup Air Intake Unfiltered Inleakage Into Control Structure Control Room Recirculation Flow Rate Intake Charcoal Absorber Thickness Intake Filter Charcoal Absorber Iodine Removal Design Basis Intake Filter Charcoal Absorber Removal Efficiency as specified in Tech. Spec.
Filter Testing 5810 CFM 10 CFM 0 CFM*
4 II 99'yo Methyl Iodide Penetration of less than.75'/o after 720 hrs operation FSAR 6.4.2.3 FSAR 6.4.2.1 FSAR 6.5.1.2.2 FSAR 6.5.1.2.1 SSES Technical Specifications 3/4.7.2 Filtered Outside AirSupplied to Control Structure Assumed
- SSES does not employ filtered recirculation 1)
K.G. Murphy and K.M. Campe, "Nuclear Power Plant Control Room Ventilation System Design For Meeting General Design Criterion 19", 13th AirCleaning Conference, August 1974.
- Control Room X/Q and Occupancy Factors Used:
Time Period 0-8 hrs 8-24 hrs 1-4 days 4-30 days Wind Speed Adjustment Factor 0.67 0.50 0.33 Wind Direction Adjustment Factor 0.88 0.75 0.50 Wind Adjusted Control Room X/Q (sec/m
)
3.32x10+
1.96xt 0+
1.25x10+
5.48x10 5 Occupancy Adjustment Factor 0.60 0.40 Overall Control Room X/Q (sec/m3) 3.32xt 0+
1.96xt 0+
7.64x10 5 2.19x10 5 Note:
The wind and occupancy adjustment factors are from Murphy-Campe.
The derivation of the unadjusted Control Room X'Q (3.32x10+ ser Jm3) is given in SSES FSAR Section 15B.2.
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- SCTS Parameters Flow Rate:
ATTACHMENTTO PLA-4274 Page 10 of32 10,500 CFM until -1/4" Water Gauge; 3997 CFM thereafter for 3 Zone Mixing (FSAR Table 15.6-22)
Charcoal absorber Iodine Removal Efficiency:
99% for all Iodine species - (FSAR Table 15.6-22)
Reactor Building Pressure Drawdown Time:
3 minutes - (FSAR Table 15.6-22)
- ECCS Leakage Outside Containment Suppression Pool Water Volume:
Minimum Water Volume - 122,410 ft Maximum Water Volume - 131,550 ft (FSAR Table 6.2-1)
- Maximum Allowable ECCS Leak Rate Outside Containment:
SSES does not have a Technical Specification which directly addresses this leakage pathway.
However, Technical Specification 3.6.1.2(e) limits leakage for all containment isolation valves in hydrostatically tested lines which penetrate the primary containment to 3.3 gpm. In the SSES DBA-LOCA analysis, 5 gpm is assumed for circulating water Engineered Safety Feature (ESF) leakage into secondary containment (Reactor Building which is treated by the Standby Gas Treatment System).
NUREG-0776 assumed 1 gpm for this leakage.
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ATTACHMENTTO PLA-4274 Page 11 of32 (3)
Provide Integrated Iodine Releases for SSES The following figures have been provided by BWROG and represent iodine release data based upon the dose analysis performed by BWROG in support of the SSES proposed Technical Specification Amendment Nos. 178 and 132 to License Nos. NPF-14 and NPF-22, respectively for the elimination of the MSIV-LCS with increase in MSIV leakage.
(See six following figures)
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2.0E+5 I-131 Integrated Release to MSIVs from Containment First 24 Hours 1.8E+5 1.6E+5 1.4E+5 Eiemeental+ Particulate+
Organic'.2E+5 1.0E+5 0
8.0E+4 Elemental+ Particulate 6.0E+4 4.0E+4 2.0E+4 O.OE+0 0
10 Time in Hours 15 20 25
1.6E+6 l-131 Integrated Release to MSIVs from Containment 1 Day to 30 Days 1AE+6 Elemental
+ Particulate+ Organic 1.2E+6 Ele entai+ Particulate 1.0E+6 8.0E+5 0
6.0E+5 4.0E+5 2.0E+5 O.OE+0 0
400 Time in Hours
1-131 Integrated Released from MSIVPiping to Condenser First 24 Hours 1AE+5 Resuspended Organic component negligible 1.2E+5 y
Elemental + Particulate+ Organic 1.0E+5 8.0E+4 0
6.0E+4 4.0E+4 Elemental+ Particulate 2.0E+4 O.OE+0 0
10 Time in Hours 15 25
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4.0E+5 1-131 Integrated Released from MSIV Piping to Condenser 1 Day to 30 Days Elemental+ Particulate+ Organic+ Resuspendecf Organic 3.5E+5 3.0E+5 Elemen al.+ Particulate+ Organic 2.5E+5 Ele ental+ Particulate 2.0E+5 O
1.5E+5 1.0E+5 5.0E+4 O.OE+0 0
Time In Hours
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2.0E+2 I-131 Integrated Release to Environment from Condenser First 24 Hours 1.SE+2 Elemental+ Particulate+ Organic+ Resuspended Organic 1.6E+2 1AE+2 1.2E+2 1.0E+2 O
8.0E+1 6.0E+1 Elemental+ Particulate+ Org nlc 4.0E+1 Elemental+ Particulate 2.0E+1 O.OE+0 0
io Time in Hours 15 25
4.5E+4 integrated I-131 Release to Environment from Condenser 1 Day to30Days 4.0E+4 Elemental+ Particulate.+ Organic+ Resuspended Organic 3.5f+4 3.0E+4 2.5E+4 2.0E+4 1.5E+4 Elemental+ Partfcui e+ Organic 1.0E+4 5.0E+3
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Elemental+ Particulate, 0.0E+0 0
Time in Hours
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ATTACHMENTTO PLA-4274 Page 18 of32 (4)
Provide a discussion of the methodology used to determine thyroid dose estimates reflected in the FSAR.
The SSES FSAR is being updated to reflect power uprated conditions. The following discussion refers to the methodology employed to assess radiological consequences associated with a DBA-LOCA for SSES power uprate conditions.
PPRL Calculation EC-RADN-1009 was performed to assess offsite and control room doses resulting from a design basis accident-loss of coolant accident (DBA-LOCA) for the Susquehanna Steam Electric Station (SSES) at power uprated conditions. These doses are reported in FSAR Section 15.6.5. The calculation employs the TACT5'omputer program which has been validated and verified for design basis use at PPRL. Dose conversion factors are based on ICRP 30 guidance.
For the dose assessment of MSIV/LCS deletion with increased MSIV leakage, a specific analysis in Calculation EC-RADN-1009 is performed to supply "base" doses with MSIV leakage set to zero. (All other parameters are based on DBA-LOCA input).
BWROG methodology as reviewed by USNRC is used to assess further dose contributions due to the deletion of the MSIV/LCS with increased MSIV leakage. These contributions are added to the base doses to provide the total doses associated with operation of SSES with deletion of MSIV/LCS along with increased MSIV leakage.
The TACT5 program is used to provide 2-hour Site Boundary (Exclusion Zone), 30-day Low Population Zone and 30-day Unprotected Control Room thyroid doses (excluding MSIV/LCS contribution as noted above)
~ TACT5 computes the transport of activity from drywell through reactor building to environment using a Regulatory Guide 1.3 source term released into the drywell at accident time T 0.
Design Basis input parameters to the TACT5 analysis are taken from sections 15.0, 15.6.5, and Appendix 15B of the SSES FSAR.
Some of these parameters are specifically described in responses (1) and (2) to this RAI. TACT5 applies atmospheric diffusion factors to calculate radionuclide concentration at the dose receptor point and dose conversion factors calculate off-site thyroid doses at the above stated off-site receptor points. Protected control room thyroid doses are calculated from the TACT5 supplied unprotected control room thyroid doses using Murphy-Campe~ methodology to determine an Iodine Protection Factor (IPF) which is used as a divisor to the unprotected control room dose. This procedure provides the protected SSES control room 30-day thyroid dose.
BWROG calculated additional dose contributions are then added to the respective TACT5 determined receptor point dose to provide total doses associated with MSIV/LCS elimination with increased MSIV leakage.
NUREG-5106, "User's Guide for the TACT5 Computer Code", prepared for USNRC, june 1988.
K.G. Murphy and K.M. Campe, "Nuclear Power Plant Control Room Ventilation System Design For Meeting General Design Criterion 19", 13th AirCleaning Conference, August 1974.
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ATTACHMENTTO PLA-4274 Page 19 of32 (5)
Was there any test conducted for unfiltered in-leakage to control room ever after the plant was constructed and operated?
'he'control structure is maintained at a positive pressure (+1/8'-'.g:) relative to the atmosphere.
The control structure positive pressure ensures that no inleakage'rom the turbine or. reactor buildings will occur. A surveillance,-SE-030-002, is performed every 18 months to assure that. the control structure is maintained at a positive pressure.
During an accident a small percentage of air is exhausted due to building leakage and fresh air from outside is the makeup source for this air. If the outside air is contaminated the CREOASS system will remove any radioactive iodines.
In Section 6.4 of the Standard Review Plan (SRP), control rooms are categorized by design features, Susquehanna's control room meets the design features of a type 3 control room.
The SRP recommends a conservative value of 10 CFM for inleakage into Type 3 control rooms.
The actual inleakage into the control room has never been measured although the design features suggest that no inleakage would occur.
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ATTACHMENTTO PLA-4274 Page 20of32 ENCLOSURE 2 MSIVALTERNATETREATMENTSYSTEM MARGINEVALUATION
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(I ATTACHMENTTO PLA-4274 Page 21 of32 MSIV LEAKACEALTERNATETREATMENTSYSTEM MARCIN EVALUATION Seismically analyzed piping within the MSIV Leakage Alternate Treatment Method includes the main steam line from containment isolation valves to the turbine stop valves, the bypass piping from the main steam line to the main condensers, the main steam drain line header from containment isolation valves to in-line pipe anchors, and portions of main steam branch connection lines to in-line pipe anchors.
Design methods for these analyzed lines are consistent with Seismic Category I qualification methods for the SSES and design margins are accordingly adequate to assure acceptable seismic performance.
Certain main steam piping and components that are part of the alternate MSIV leakage treatment system are not currently classified as Seismic Category I and, therefore, seismic piping analyses do not exist. The BWR Owners'roup has determined that the main steam piping is extremely rugged and B31.1 design requirements lead to a considerable reserve margin.
The same conclusion was reached after the BWR Owners'roup performed post-earthquake reconnaissance to investigate past performance of non-seismically designed piping in non-nuclear applications which experienced high magnitude earthquakes far in excess of the Design Basis Earthquake (DBE) at SSES.
Seismic walkdowns were performed to verify that SSES piping and instrumentation within the alternate MSIV leakage treatment system are free of impact interactions from falling and the proximity or differential motion hazards.
Conditions outside the experienced database boundary (outliers) have been reviewed to demonstrate reasonable assurance-of the integrity of the associated piping systems and components under normal and earthquake loading.
In addition, a
representative bounding pipe support sample on the 4" main drain line was reviewed to demonstrate pipe support margins.
These reviews demonstrated that the non-seismic analyzed piping systems consisting of welded steel pipe and standard support components, are consistent with the construction standards associated with the seismic experience database piping systems.
Reviews also demonstrated that adequate design margins exist.
Main Steam Drains to Condenser i(
The main steam drain line to the condenser consists of safety (Class
- 2) and non-safety related pipirIg.
The safety related pipe and portions of the non-safety piping up to in-line pipe anchors downstream of isolation valves HV-1/241F019 and F020 were seismically analyzed.
These piping systems were designed in accordance with the ASME Code,Section III, Class 2 and ANSI B31.1
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(i ATTACHMENTTO PLA-4274 Page 22 of32 requirements, using response spectra analysis techniques.
The piping design code used was the ASME III, Class 2, 1971 edition including Winter 1972 Addenda and B31.1, 1973 Edition. Support types include springs, struts, and snubbers and their design was governed by AISC, ANSI, B31.1, and MSS SP58.
Design methods for the seismically analyzed drain piping are consistent with Seismic Category I
qualification methods for SSES.
Therefore, the design margins associated with these systems and their supporting structures will be adequate to insure piping system integrity under projected seismic performance.
The remaining main steam drain and associated piping were analyzed for dead weight and thermal loads using computer analysis and spacing criteria.
This piping is similar to piping found in the seismic experience database.
The seismic verification walkdowns identified minor interaction issues that could be potential sources of damage.
Actions have been initiated to resolve these issues.
The supplemental field verification determined that the support types used are considered to have good seismic performance.
The system is predominantly supported for dead weight utilizing rod hangers.
Component designs are constructed from standard support catalog parts typically consisting of clamps, threaded rods, weldless eye nuts, turnbuckles, welding lugs and are attached to either concrete or structural steel.
These support types are designed to resist vertical loads in tension. Design capacities are provided by manufacturers'oad rating data sheets.
Load capacity ratings for component standard supports are typically based on testing and utilize a factor of safety of five in accordance with MSS SP-58.
The load on which the load capacity data (LCD) is based is therefore a factor of five higher than the catalog load rating. The margin capacities for each support component are taken as the LCD x 5 x 0.7 (EPRI NP-6041).
Including thermal effects on allowable loads, component standard supports designed by load rating is calculated as follows:
TL x 0.7Su/Su*
where:
TL:
Su:
Su*:
Support test load is less than or equal to load under which support fails to perform its intended function; TL - LCD x 5 Material ultimate strength at temperature Material ultimate strength at test temperature Structural steel support members are evaluated using section strength based on the plastic design methods in Part 2 of AISC or 1.7 times the AISC working stress allowables.
Concrete anchor bolts are evaluated using data from the A46/SQUG criteria, Appendix C.
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II ATTACHMENTTO PLA-4274 Page 23 of32 In addition to the resolution of the walkdown outliers, seismic'margin.evaluation of the anchor bolts and structural steel members of a random bounding sample of pipe supports on the critical 4-inch main drain line was conducted.
This assessment is more conservative and than, the evaluation referenced in NEDC-31858P.
The objective of the evaluation of the non-seismic main steam drain piping is to demonstrate that piping position retention will be maintained during a seismic event to provide assurance that the pipe supports will behave in a ductile manner and that all lines are free of known seismic hazards.
In addition, it will establish that these SSES piping systems will perform in a manner similar to piping and supports that have been observed to demonstrate good seismic performance.
The evaluations'oal is to produce High-Confidence-Low-Probability of Failure for the selected pipe support sample.
This should provide the desired reasonable assurance of good seismic performance.
In addition to the seismic DBE loads, dead-weightand operating mechanical loads are accounted for.
Operating mechanical loads for this system are thermal expansion loads and design dead weight support loads are consistent with tributary area weight procedures.
The actual piping stresses and pipe support loads due to these design loads are obtained by performing ME101 computer analyses.
The anchor bolts are evaluated in accordance with SQUC Appendix C anchorage data.
The anchor bolts are HILTI KWIKor equivalent as identified on the pipe support drawings and concrete strength is 4000 psi or greater.
Two sizes of anchor bolts have been used for the selected pipe supports.
The allowable loads per Table C.2-1 of SQUC Appendix C are:
Pullout Capacity Shear Capacity Min Spacing Min Edge Dist.
3170 Ibs 3790 Ibs 6.25" 6.25" 4690 Ibs 5480 Ibs 7 5 II 7 SII The piping consists of the main steam, RCIC, and HPCI turbine steam line drains from valves F019, F026, and F029 to the HP condensers.
The SSES Unit 1 model envelopes approximately 283 feet of 4 inch schedule 120 piping which is supported by 23 hangers.
Figure 1
presents piping configuration. The types of hangers are as follows:
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ATTACHMENTTO PLA-4274 Page 24 of32 Anchor Spring Structural Cantilever (vertical restraint) - 07 Single Member (lateral restraint) - 03 Hanger Rod Strut 01 10 10 01 01 Total 23 Figure 2 shows the configurations of the selected pipe supports.
The following table shows the mark number, type, and restraint directions of the selected pipe supports.
EBD-114-H13 EB D-1 14-H23 EBD-114-H25 EBD-114-H30 EBD-214-H1 6 EBD-214-H20 EBD-214-H22 EBD-214-H27 Structural Structural Structural Anchor
. Structural Structural Structural Spring and Strut Vertical Lateral Lateral 6-way Vertical Vertical Vertical Vertical and lateral Four SSES Unit 1 and four SSES Unit 2 pipe supports are selected for the seismic margin evaluation.
The selected population is obtained based on an overview of the support configurations and the original design qualification.
This population includes both vertical and lateral restraint type supports.
Since non-safety design practices normally do not include lateral restraints, a seismic event will result in these types of supports being subjected to increased loadings.
Therefore, pipe supports with lateral resisting capacity are natural candidates for inclusion in the selected population.
In
- addition, some supports are also selected based on the indication that minimum remaining component stress or load margins existed between documented and allowable design values.
In this respect, the selection process is a conservatively biased one.
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ATTACHMENTTO PLA-4274 Page 25 of32 The original seismic design and seismic floor curves generated are documented in Calculation 0 37 of Volume 19 ETO 12B titled "Turbine Building and Auxiliary -Bay-Seismic" for-the Turbine Building. The seismic design included the development of three lumped mass models for the east-west, north-south, and vertical directions.
The seismic floor curves were generated to determine seismic anchor forces and displacements for the piping systems that are attached to the Turbine Building.
Therefore, seismic floor curves were only generated for 1/2% (OBE) and 1.0% (DBE) equipment damping values.
The IPEEE generic methodology is recommended by the BWR Owners'roup to demonstrate seismic adequacy of the non-seismically designed main steam drain lines.
This generic methodology is described in EPRI NP-6041.
Currently and during the original seismic qualification of piping systems at SSES, an equipment damping value of 1.0% is utilized in the seismic piping analyses for DBE loading. However, a realistic equipment damping value of 5.0% is recommended as part of the 'generic methodology.
This will result in a lower seismic response which is appreciated in the seismic evaluation of the identified outliers and margin evaluations.
The existing 1/2% and 1% floor response spectrum curves are extrapolated to generate the 5% DBE floor response spectrum curves for the Turbine Building.
The ratio of the allowable stress to actual calculated stress for the 4-inch main drain line is as follows:
Design Condition Sustained Loads Occasional Loads Thermal Expansion Maximum Computed Stress (psi) 4494 14593 5691 Allowable Stress (psi) 15000 (SH) 36000 (2.4 SH) 22500 (SA)
Ratio 3.34 2.47 3.95
. Pipe Support Margins:
The ratio of the allowable load to actual calculated load for the selected pipe supports is as follows:
Retie. (See Note 1)
EB D-114-H13 EB D-114-H23 EBD-114-H25 EBD-114-H30 1.40 1.91 1.91 See Note 2
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jl ATTACHMENTTO PLA-4274 Page 26 of32 EBD-214-H16 EB D-214-H20 EBD-214-H22 EBD-214-H27 1.08 1.93 1.05 1.27 Note 1:
Ratio represents the lowest safety factor for the most stressed elements (anchor bolts, base plate, or structural steel member) of the pipe support being analyzed.
Note 2:
This is an anchor which is welded to an embedded plate and it has 3/4" HILTI KWIK bolts.
Since piping on one side of the anchor was not analyzed, plastic shear moment capacity of the pipe were utilized for anchor design.
Therefore, the anchor bolts have been designed for the worst possible loading and, consequently, they are judged to be adequate for DBE loading.
In conclusion, it has been demonstrated that the proposed MSIV Alternate Treatment System complies with Appendix A of 10CFR100 by using a combination of the NEDC 31858P experience-related seismic methodology comparisons and confirmatory seismic analyses of the main 4" drain line and a sampling of critical drain line pipe supports.
The results above show that design margin still exists between actual and allowable loads for piping and pipe supports when seismic load conditions are applied.
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ATTACHMENTTO PLA-4274 Page 29 of32 ENCLOSVRE3 EXEMPTIONTO 10 CFR 50 APPENDIXJ
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ATTACHMENTTO PLA-4274 Page 30 of32 The proposed action is to maintain the current exemption from 10 CFR 50, Appendix j, Sections II.H.4 and III.C.2, which exempts Susquehanna Steam Electric Station (SSES) from the requirement to perform Main Steam Isolation Valve (MSIV) leak rate testing at the calculated peak containment pressure related to the design basis accident, and Section III.C.3 requiring that the MSIV measured leak rates be included in the combined local leak rate test results.
Pennsylvania Power and Light (PPAL) is proposing to re-apply for the exemption based on a proposed change to the current exemption description; specifically, the elimination of the MSIV Leakage Control System (LCS), use of the alternate leakage treatment method, and increasing the assumed MSIV leakage rate from 11.5 standard cubic feet per hour (scfh) to 100 scfh per steam line, not to exceed a total of 300 scfh for all four main steam lines. The proposed exemption will maintain the leak rate testing of the MSIVs at a reduced pressure (i.e., 22.5 psig when applied between valves) and will continue to exclude the measured leakage from the combined local leak rate test results.
SSES Unit 1 and 2 received an exemption from 10 CFR 50, Appendix j, Sections II.H.4, III.C.2, and III.C.3 based on the NRC conclusion that the SSES MSIV leak testing methods were acceptable alternatives to the requirements.
This conclusion was included in the SSES Safety Evaluation Report (SER) (i.e., NUREG-0776).
In addition to the SSES leak testing program the SER described that in the event of a Loss-of-Coolant Accident (LOCA), the MSIV Leakage Control System will maintain a negative pressure between the MSIVs, and the effluent will be discharged into a volume where it will be processed by the standby gas treatment system before being released to the environs.
Furthermore, the SER described that a radiological analysis including this potential source of containment atmosphere leakage was performed.
The analysis was based on an assumed leak rate limit for a main steam line through the MSIVs of 11.5 scfh, and the MSIVs were planned to be periodically tested to ensure the validity of the radiological analysis.
The NRC concluded that the current SSES testing procedure where two valves on one steam line are tested simultaneously, between the valves, utilizing a reduced test pressure (i.e., half the peak containment pressure or 22.5 psig applied between the MSIVs) was acceptable, and excluding the MSIV test leakage rate from the combined local leak rate was acceptable because the MSIV leakage had been accounted for separately in the radiological analysis of the site.
PLA-4228, dated November 21, 1994, entitled, "Proposed Amendment No. 178 to License No.
NPF-14 and No. 132 to License No. NPF-22: Increase of MSIV Leakage Rate and Deletion of Leakage Control System,"-supports*the planned modification to eliminate the QSIV Leakage Control System and utilize an alternate leakage pathway.
This proposal is based on the Boiling Water Reactor Owner's Group (BWROG) method summarized in General Electric Report NEDC-31858P, Revision 2, "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control System."
Therefore, the description of the MSIV Leakage Control System and the assumed
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ATTACHMENTTO PLA-4274 Page 31of32 MSIV leak rate will no longer be accurate once the proposed modification is performed and the implementing Technical Specification (TS) change is approved.
'As stated in the November 21, 1994, proposed amendment, a plant specific radiological analysis has been performed in accordance with NEDC-31858P, Revision 2, to assess the effects of the proposed increase to the allowable MSIV leakage rate in terms of Control Room and off-site doses following a postulated design basis LOCA. This analysis utilizes the hold-up volumes of the main steam piping and condenser as an alternate method for treating MSIV leakage.
The radiological analysis uses standard conservative assumptions for the radiological source term consistent with Regulatory Guide (RG) 1.3,
'Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-Of-Cooling Accident for Boiling Water Reactors,'evision 2, dated June 1974.
The analysis results demonstrate that dose contributions from the proposed MSIV leakage rate limitof 100 scfh per steam line, not to exceed 300 scfh for all four main steam lines, along with the proposed deletion of the MSIV Leakage Control System, result in an acceptable increase to the LOCA doses previously evaluated against the regulatory limits for the off-site doses and Control Room doses contained in 10 CFR 100, and 10 CFR 50, Appendix A, General Design Criteria (GDC) 19, respectively. of the November 21, 1994 submittal shows the new calculated doses resulting from the proposed changes.
Based upon the description contained in the SER, the NRC concluded the current exemption was acceptable based on: the method of MSIV testing (i.e., 22.5 psig test pressure when applied between MSIVs on a single steam line); that a radiological analysis was performed, assuming a 11.5 scfh MSIV leak rate, and the MSIVs.-would be periodically tested to ensure validity of the radiological analysis (i.e., verify that the MSIV leakage rate assumed in the radiological analysis is not exceeded per TS 3.6.1.2.c.); and a MSIV leakage rate measured during testing is accounted for separately in the radiological analysis of the site. The proposed changes do not adversely affect the current exemption because the modification and implementing TS change will not alter the procedural method of MSIV testing (i.e., test pressure will remain at 22.5 psig when applied between MSIVs); is based on the results of a radiological analysis, where the proposed leakage rate is assumed, and the resulting doses are still within regulatory limits, and the MSIVs will be periodically tested to assure the validity of the analysis (i.e., verify that the proposed MSIV leakage rate assumed in the radiological analysis is not exceeded per proposed TS 3.6.1.2.c.); and the MSIV leakage willstill be accounted for separately in the radiological analysis of the site.
The NRC may, upon application, grant exemptions from the requirements of 10 CFR 50, where special circumstances are present.
10 CFR 50.12(a)(2)(ii) states the NRC may grant exemptions from the requirements of 10 CFR 50 whenever application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.
PPLL has concluded that: the current MSIV leak rate testing method (i.e., test pressure of 22.5 psig when applied between MSIVs) is an acceptable method; the proposed alternate MSIV leakage pathway, and the calculated doses obtained by performing a radiological analysis, which assumed MSIV leakage rate limit of 100 scfh per steam line, not to exceed a total of 300 scfh for all four main steam lines, are within the limits of 10 CFR 100 and GDC 19; it is acceptable to continue to exclude the measured MSIV leakage rate from the combined local leak rate, since the leakage is accounted for separately and continues to meet the underlying purpose of the rule.
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ATTACHMENTTO PLA-4274 Page 32 of32 The existing exemption is proposed to be retained, and is justified, since the proposed changes would not result in a significant increase to the'LOCA doses previously evaluated. against the off-site and Main Control Room dose limits contained in 10 CFR 100 and 10 CFR SO, Appendix A, General Design Criteria 19, respectively.
As described in the preceding "Justification For Exemption," the method of calculating the revised doses is highly conservative, and the doses resulting from a postulated design basis LOCA are below regulatory limits.
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