ML20135E684

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Proposed Tech Specs,Requesting Amend to Section 3.4.2 to Revise SRV Configuration to Include Only 13 of Current 18 SRVs
ML20135E684
Person / Time
Site: LaSalle  Constellation icon.png
Issue date: 12/02/1996
From:
COMMONWEALTH EDISON CO.
To:
Shared Package
ML19310D739 List:
References
NUDOCS 9612110423
Download: ML20135E684 (27)
v • d • e


Text

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I ATTACHMENT B j i

TECHNICAL SPECIFICATION CHANGES FOR LASALLE UNIT 1 i I

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SUMMARY

OF PROPOSED CHANGES FOR LASALLE UNIT 1 (NPF-11) l Section 3.4.2 SRV safety valve function lift setting listing revised to Page 3/4 4-5 include only those SRVs not removed by the proposed amendment; Number of required SRVs changed Bases Section Number of SRVs changed to reflect new configuration B 3/4.4.2 Page B 3/4 4-2 )

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9612110423 961202 PDR ADOCK 05000373 P PDR

l REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 12 3.4.2 The safety valve function of JT of the below listed reactor coolant  :

system safety / relief valves shall be OPERABLE with the specified code safety i valve function lift setting *f; all installed valves shall be closed with OPERABLE position indication. ,

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l a. 2 A' safety / relief valves 91205 psig 13% >

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b. 3 # safety / relief valves 91195 psig 13% ',
c. 2t safety /re'ief valves 91185 psig 13%

, d. 4 safety / relief valves 91175 psig 13% ,

l e. 2 safety / relief vahes 91150 psig 13% .

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. I!

ACTION: l l a. With the safety valve function of one or more of the above. required  !

safety / relief valves inoperable, be in at least HOT SHUTDOWN within i

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.  !

b. With one or more of the above requi' red' safety / relief valve stem e position indicators inoperable, restore the inoperable stem position  !

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indicators to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. '

SURVEILLANCE RE0VIREMENTS l

4.4.2.1 The safety / relief valve stem position indicators of each safety / relief valve shall be demonstrated OPERABLE performance of a:

a. CHANNEL CHECK at least once per 31 days, ,a
b. CHANNEL CALIBRATION at least once per 18 s nths.** ,1 l 4.4.2.2 The low-low set function shall be demonstrated not to interfere with

! the OPERABILITY of the safety / relief valves or the ADS by performance of a l CHANNEL CALIBRATION at least once per 18 months. ,i '

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  • The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. Following testing, lift settings shall be within 11%.
  1. Up to two inoperable valves may be replaced with spare OPERABLE valves with r lower setpoints until the next refueling outage.
**The provisions of Specification 4.0.4 are not applicable provided the j surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

LA SALLE - UNIT 1 3/4 4-5 Amendment No.113 l

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.d g gT JORLANT SYSTEM '

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3/4.4.2 SAFETY / RELIEF VALVES j

The safety valve function of the safety / relief valves operste ta prevent the reactor coolant system free being presser 12ad above the Safety Limit of

, _t 1325 o,Lig_in accord.nce with __ASME Cgse. Analysis has shown that witt, the j v7tto Tsafety function or one- oT ichteep safety / relief valves inoperable the j

reactor pressure is limited to withTn ASME III allowab_13 values for the worst case upset transient. Therefore, operation with any K .SRV'c casable of opening is allowable, although all installed SRV's mustD c' osed and have position indication to ensure that integrity of the primary coolant boundary is known to exist at all times. ~

Demonstration of the safety / relief valve lift settings will occur only i

during shutdown and will be performed in accordance with the provisions of j 5pecification 4.0.5. -

I 3/4.4.3 REACTOR CDOLANT SYSTEM LEAKAGE l 3/4.4.3.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are pro-1 vided to monitor and detect leakage from the reactor coolant pressure boundary.

These detection systems are consistant with the recommendations of Regulatory j

Guide May 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems,"

1973.

j 3/4.4.3.2 OPERATIONAL LEAKAGE 4

The allowable leakage rates free the reactor coolant systes have been based on the predicted and experimentally observed behavior of cracks in pipes. The norsally expected background leakage due to equipment design and the detection f i

capability of the instrnamentation for detersining system leakage was 'also considered.

! The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for unidentified leakage the probability is small that tM imperfection or crack associated with such leakage would grow rapidly.

! Hows.wr, in all cases, if the leakage rates exceed the values specified or the leakage h located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to allow further investigation and corrective action.

The Surveillance Requirements for RCS pressure isolation valves provide i added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LDCA. Leakage from the RCS pressure i

isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

3/4. 4. 4 CHEMISTRY The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant. Chloride limits are specified to prevent stress corrosion cracking of the stainless steel.

' The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the higher limit on chlorides is permitted during POWER OPERATION.

During shutdown and refueling operations, the tamperature necessary for stress corrosion to occur is not present so high concentrations of chlorides

are not considered harmful during these periods.

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LA SALLE-UNIT 1 B 3/4 4-2 i Amendment No. 60

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TECHNICAL SPECIFICATION CHANGES FOR LASALLE UNIT 2 I

SUMMARY

OF PROPOSED CHANGES FOR LASALLE UNIT 2 (NPF-18)

I Section 3.4.2 SRV safety valve function lift setting listing revised to Page 3/4 4-6 include only those SRVs not removed by the proposed arnendment; Number of required SRVs changed Bases Section Number of SRVs changed to reflect new configuration B 3/4.4.2 Page B 3/4 4-1a i

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l REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of of the below listed reactor coolant system safety / relief valves shall be OPERABLE with the specified code safety valve function lift setting *#; all installed valves shall be closed with OPERABLE position indication.

a. 2/ safety / relief valves 91205 psig 13% *
b. St safety / relief valves 91195 psig 13% .
c. 2 / safety / relief valves 91185 psig 13%
d. 4 safety / relief valves 91175 psig 13%
e. 2 safety / relief valves 91150 psig 13%

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. '

i ACTION:

a. With the safety valve function of one or more of the above required safety / relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b. With one or more of the above required safety / relief valve stem position indicators inoperable, restore the inoperable stem position indicators to OPERABLE status w'ithin 7 days or be in at least HOT ,i SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the i following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. I SURVEILLANCE RE0VIREMENTS l

4.4.2.1 The safety / relief valve stem position indicators of each safety / relief valve shall be demonstrated OPERABLE by performance of a:

a. CHANNEL CHECK at least once.per 31 days, and a *
b. CHANNEL CALIBRATION at least once per 18 months.**

4.4.2.2 The low low set function shall be demonstrated not to interfere with the OPERABILITY of the safety / relief valves or the ADS by performance of a CHANNEL CALIBRATION at least once per 18 months.

  • The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. Following testing, lift settings shall be within il%.
  1. Up to two inoperable valves may be replaced with spare OPERABLE valves with lower setpoints until the next refueling outage.
    • The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

LA SALLE - UNIT 2 3/4 4-6 Amendment No. 98

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_R[ ACTOR COOLANT SYSTDI i

l 3/4. 4. 2 $AFETY/ItfLIEF VALVES  !

The safety valve function of tha safety / relief vale.,s orgrats ta :rev6nt the recetar ceolant system free being Trasurized above the Safety Limit of 1325 psis in accordance withmthe M qads. Analysis has shown that with the i free sarety function of one of tAttaishteerDasfety/ relief valves inoperable the reactor pressure is limited to witA1n M III allowable values for the worst I i

case upset transient. Therefore, operation with ary 33C3RV's capable of opening is allowable, although all installed SRV's must te closed and have position indication to ensure that integrity of the primary coolant boundary is known to exist at all times. '

Demonstration of the safety / relief valve lift settings will occw only during shutdown and will be performa$ in accordance with the provisions of Specification 4.0.5.

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LA SALLE - INGT 2 8 1/4 4-la Amaruhant No. 41

ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison proposes to remove five of the currently installed Safety / Relief Valves (SRVs), due to the current excess capacity, and to reduce maintenance costs and worker radiation dose, The current requirement for 17 of the 18 installed SRVs to be operable would be changed to require 12 of the 13 installed SRVs to be operable.

Commonwealth Edison has evaluated the proposed Technical Specification Amendment and determined that it does not represent a significant hazards consideration. Based on the criteria for defining a significant hazards consideration established in 10 CFR 50.92, operation of LaSalle County Station Units 1 and 2 in accordance with the proposed amendment will not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated because:

The probability of an accident previously evaluated will not increase as a result of this change, because the change in valve configuration, and the accompanying piping modification does not alter any of the initiators of an accident or cause them to occur more frequently. The piping modifications will be performed consistent with the current piping classifications for the affected components. Removal of the SRVs will not impact the ability of the remaining SRVs to perform their functions, as described below.

The consequences of an ASME Overpressurization Event are not significantly increased and do not exceed the previously accepted licensing criteria for this event. General Electric (GE) has calculated the revised peak vessel pressure for LaSalle Station to be 1341 psig, which is below the 1375 psig criterion of the ASME Code for upset conditions, referenced in Section 5.2.2, Overpressurization Protection, of the Updated Final Safety Analysis Report (UFSAR), and NUREG-0519 (Safety Evaluation Report related to the operation of LaSalle County Station, Units 1 and 2, March 1981), and Section 15.2-4, Closure of Main Steam Isolation Valves (BWR) of NUREG-0800 (Standard Review Plan). The consequences of this event will continue to be verified on a cycle-specific basis, beginning with LaSalle Unit 1 Cycle 9 (L1C9). These analysis results will be approved as part of the normal reload licensing 10CFR50.59 processes.

GE has also performed an analysis of the limiting Anticipated Transient Without Scram (ATWS) event, which is the MSIV Closure Event (MSIVC). This analysis C-1

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calculated the peak vessel pressure to be 1378 psig, which is well below the 1500 psig criterion of the ASME Code for emergency conditions. General Electric has verified that these results will not be impacted with the introduction of Siemens fuel.

The conclusions given in the safety analyses with regards to primary containment dynamic loads, main steam piping loads, Loss-of-Coolant Accident (LOCA) impact, Minimum Critical Power Ratio (MCPR) impact and SRV

' availability also show that current accident and transient analyses are not  !

impacted by this change beyond those reanalyzed by GE and discussed above.

There is no increase in the amount or types of radioactive release for any of the  ;

affected accidents or transients.

Therefore, there is not a significant increase in the consequences of an accident  ;

previously evaluated.

2) Create the possibility of a new or different kind of accident from any accident >

previously evaluated because:

The as-left SRV piping configuration will continue to be consistent with the l current classifications for these piping and supports, and have been evaluated by Sargent and Lundy analyses. This ensures no different types of events may l be caused by piping failures at these locations. This is the only physical L modification proposed by this submittal, and it will not create the possibility of a ,

new or different kind of accident from those previously evaluated. Other systems are not modified with this change and have been shown in this submittal to continue to function as intended with the new system configuration, with the l exception of the abandoned discharge line snubbers which may be replaced with struts, except where they will be retained as snubbers due to thermal expansion  ;

requirements. The changed supports are required to function only as struts with  ;

the revised piping. Consideration and evaluation of this function ensure no new or different accidents are created.

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3) Involve a significant reduction in the margin of safety because:

i While the calculated peak vessel pressures for the ASME Overpressurization  !

Event and the MSIVC ATWS Event are increased due to the proposed SRV  ;

removals, the new peak pressures remain below the respective licensing  ;

acceptance limits associated with these events. i The actual cycle-specific reload analysis of the ASME Overpressurization Event will be verified to be within the licensing acceptance limit for that event prior to each cycle startup, as required in the normal reload 10CFR50.59 process.

These licensing acceptance limits have been previously evaluated as providing l a sufficient margin of safety. For other accidents and transients, including l suppression pool loadings, the SRV removals have a negligible, if any, effect on I the results, so the margin of safety is preserved.

Guidance has been provided in " Final Procedures and Standards on No Significant Hazards Considerations," Final Rule, 51 FR 7744, for the application of standards to license change requests for determination of the existence of significant hazards considerations. This document provides examples of amendments which are and are not considered likely to involve significant hazards considerations. These proposed amendments most closely fit the example of a change which may either result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the applicable Standard Review Plan.

This proposed amendment does not involve a significant relaxation of the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings or a significant relaxation of the bases for the limiting conditions for ,

operations. Therefore, based on the guidance provided in the Federal Register and  !

the criteria established in 10 CFR 50.92 (c), the proposed change does not constitute a i significant hazards consideration.  !

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N ATTACHMENT D ENVIRONMENTAL ASSESSMENT STATEMENT APPLICABILITY REVIEW Commonwealth Edison has evaluated the proposed amendment against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. It has been determined that the proposed changes meet the criteria for a categorical exclusion as provided under 10 CFR 51.22 (c)(9). This conclusion has been determined because the changes requested do not pose significant hazards consideration or do not involve a significant increase in the amounts, and no significant changes in the types, of any effluents that may be released offsite. Additionally, this request does not involve a significant increase in individual or cumulative occupational radiation exposure.

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ATTACHMENT G i i

l GENERAL ELECTRIC  ;

REPORT CONFIRMING APPLICABILITY l OF GENERAL ELECTRIC ATWS ANALYSIS TO SIEMENS FUEL l

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OE Nuclear Eastgy - Engineanng O Licenstag Consulting $srvices .

Review of ATWS Analysis for Attemate Core Characteristics for LaSalle 4

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l Esthasted Impact of Alternate Core Charseteristics l forLaSaBe ATWS (ODYN) Analysis l

This memo summarizes the review of the impact on the ATWS analysis for i

LaSalle with respect to altemate core characteristics than the GE9 fuel used in the GE i analyses (Reference 1). The basis of the alternate core characteristics used in this review -  !

i as provided by Comed (Reference 2).  ;

Summarv-  !

l l The changes in core characteristics are generally small and therefore do not have a significant impact on the ATWS analysis results. h major diffennce is in the void i 1

coefficient, where a 25% decitase is calculated. The ved impact of this change in '

void coefficient is a decrease in vessel peak pressure and power, and an increase in suppression pool temperature and containment pressure. However, the affected events, MSIV closure and TTNBP are not limiting with respect to suppression pool temperature  !

l and containment pressure. Therefore, the changed core characteristics do not result in a i significant impact to the ATWS analysis. >

Dimennian- 1 i

The assessment of the impact on the ATWS analysis is made based on the )

differences between the GE9 fuel (in Reference 1) and those of an alternate core and fuel design (from Reference 2). The table shown below lists the variation of key core parameters, which are used in this assessment. The variations in the key parameters noted in the table differ by less than 5%, except for the void coefficient, and fuel' l temperature. The impact of these two parameters outside the range of GE9 fuel is discussed below. The other parameters are comparable with what is norrnally expected for GE9 fuel.

N expected impact on the ATWS analysis results, of the 102*F decrease in fuel temperature, is minimal and therefore does not impact the analysis results.

The expected impact of the 25% decrease in void coefficient is evaluated for the j three limiting events in the generic BWR/5 evaluations identiSed in Reference 3. These l ODYN Based Analysis 1

DRF W B13-01760 l March 1,1996 l

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GE Nuclear Energy . Engineenns & Liceamag Consulting Services -

Review of ATW5 Analysis for Ahernate Core Characteristics fbr LaSalle 1:

events are the Turbine Trip without Bypass (TTNBP), the Main Steam Isolation Valve  !

Closure (MSIVC), and the inadvertent Opening of a Relief Valve (IORV). The TINBP and MSIVC events result in the more limiting vessel pressure and power responses, and i 4

the IORV results in the more limiting suppression pool maximum bulk temperature and t

containment peak pressure. The change in void coefficient is estimated to result in reduced peak vessel pressure and core power, and an increase in suppression pool maximum bulk temperature and containment peak pressure is estimated for the TTNBP and MSIVC events. No significant impact of the void coefficient change is estimated for i

the limiting 10RV event as documented in Reference 3. For the void coefficient change, the estimated increase in suppression pool maximum bulk temperature and containment peak pressure for the TINBP and MSIVC is less than 5'F and 1 psi respectively.

Therefore, it is conchided that the alternate core characterictico do not havo a significant impact on the ATWS analysis results (Reference 1) for LaSalle.

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i ODYN Based Analysis 2  !

' DRF # B1F01760 March 1,1996

OE Nuclear Eastgy Enghwering & Licensing Canadting Services - -

Review of ATWS Analys5 for Alearna3 Core Characteristics for I mhh Key Parameter Change from GE9 fuel Void Coemeient, cents /% void -25.1 % -

Core Average Void, fraction +2.2 %

Core Average Doppler Coefficient,- -3.0 %

c=tsU Core Average Fuel Temperature,"F -102 7 i Normalized Core Average Axial Power -2.3 % ,

Distribution, PealdNode same node 4 Core N.sure Drop, psi +0.2 psi Boron Reactivity Worth, hot standby, +3.9 %

Ak/k 1571 ppm B .

Core Average Exposure, GWd/MTU 0.8 GWd/MTU

References:

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1) GENE-B13-01760, Rev 2, " Safety Review for LaSalle County Station Units 1 and 2 Safety Relief Valves Reduction and Setpoint Tolerance Relaxation Analyses", February 1996,
2) Ronald J Chin, letter to J Casillas, "ATWS LaSalle Fuel Parameters for Siemens ATRIUM 9B Fuel Assembly Design", February 23,1996.
3) NEDE-24222, " Assessment of BWR Mitigation of ATWS, Volume 2 (NUREG 0460 Altemate No. 3)", December 1979.

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1 ODYN Based Analysis.

3 DRF # B13-01760 March 1,1996 1

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-na as- - 4 .+-weK-.- Aa-aa.a.

ATTACHMENT H SARGENT AND LUNDY EVALUATION FOR SRV DISCHARGE PIPING AND MAIN STEAM PIPING DUE TO SRV REMOVAL

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i NDIT LAS-ENDIT 0134 Upgrade 1 )

Attachment A. Rev. 0 'l Page A 1  !

COMMONWEALTH EDISON COMPANY LASALLE COUNTY STATION - UNITS 1 & 2 Assessment Of Pipina And Supports Due To Removal Of Five SRVs From Main Steam Lines The Main Steam (MS) Lines (Piping Subsystems MS01 thru MSO4) at Units 1 & 2 of LaSalle l County Station currently have eighteen Safety Relief Valves (SRVs) in each unit. To reduce the t

time and effort spent in maintenance of these safety relief valves, a reduction of up to five SRVs is planned to be implemented during future refueling outages. The following five SRVs, as shown in the table below, are planned to be removed in both Units 1 & 2. The existing five SRVs connected to the MS02 subsystem piping will remain in place.

l Subsystem No. of SRVs SRVs Planned to Remaining No.

(Existing) be Removed of SRVs and Tail Pipes '

l MS01 4 B21-F013J 2 B21-F0138 MS02 5 None 5 l

MS03 5 B21-F013N 4 MSO4 4 B21-F013A 2 B21-F013G

( TOTAL 18 5 13 The potential impact on the containment pool hydrodynamics loads due to the reducbon in the number of SRVs from the existing eighteen valves to the future thirteen valves is part of General Electnc's work scope and is therefore not addressed in this report.

  • Descnobon of Proposed Modificabons 1

i As shown in the above table, during the scheduled refueling outages, the five listed SRVs will be removed. A blind flange will De installed at the eight inch diameter pipe end after the corresponding l SRV is removed. The twelve inch diameter SRV tail pipes, downstream of the SRVs which are l planned to be removed, may be left abandoned in place. A blind flange will be installed at each l abandoned tail pipe at the removed SRV end. These tail pipes may be removed during the outages in which the SRVs are removed or may be removed in subsequent refueling outages. The abandoned SRV tail pipes have been designed and supported for the applicable loads, meeting UFSAR j allowables except for the safety relief valve / turbine trip transient loads which are no longer applicable i for the irwettve tail pipes. Upon their final removal, these tait pipes shall be blind flanged above the l l dryweR floor and the porton of the tail pipe in the wetwell will be left in place.

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NDIT 1.AS-ENDIT-0134, Upgra'de 1 -

Attachment A Rev.1 Page A 2 i

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The snubbers left in place to support the abandoned SRV tail pipes will continue to be inspected for their functionality until they are changed to struts. The vertical direction snubbers (except of the two snubbers on 'N' SRV tail pipe for 1MS03 & 2MS03 subsystems) will be changed to rigid struts upon SRV removal. i l

The deflection of the abandoned SRV tail pipes at the cantilever end (i.e., at the removed SRV end) has been determined to vary between 1.5 inches to 4.5 inches for the faulted load condition. The impacted supports have been verified to remain functional for these deflecbons. At the first available opportunity, a walkdown will be performed to assess if such deflections can be accommodated witimt impacting any components in their vicinity. If necessary, either new supports will be added or the cantilever length of the tail pipe will be reduced to minimize the deflection to an acceptable limit.

. Pioino Data and Analysis Result ' '

l The Main Steam Piping at LaSalle Unit 1 and Unit 2 have the same physical properties and similar routing and were originally analyzed for the Unit 1 configuration. Therefore a detailed piping analysis for the Unit 1 piping has been performed and represents both Units except for MS03 subsystem l where a separate analysis has been performed for each Unit to account for as-built support l configuration differences.

l The Main Steam Piping for subsystems 1MS01,1MS03,1MSO4 and 2MS03 have been analyzed by modeling the main header piping, the eight inch diameter piping connected to the SRVs and the twelve inch diameter SRV tail pipes connected to the SRVs which remain in p? ace. The SRVs

  • including their tait pipes which are planned to be removed have been deleted from the piping model.

These abandoned tail pipes are analyzed separately.

l) l The SRVs from a particulars subsystem which are planned to be removed shall be removed dunng  !

the same refueling outage. However, the abandoned tait pipes may be removed in a different outage than the corresponding SRV, one at a time, if so desired. These tail pipes shall be removed in full l length from SRV end to the top of the drywell ficor. The abandoned (non-functional) tail pipe below ,

the drywell floor (in the wetwell) is not currently scheduled to be removed but may be scheduled for l l removal at some future outage.  !

l The following table provides information regarding these lines in Unit 1. The line numbers for the Unit

! 2 Subsystem are similar (e.g., "1" becomes a *2").

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NDIT LAS-ENDIT-0134, Upgrade 1  !!

Attachment A, Rev.1 Page A 3 ,

i Subsystem Line Numbers 1MS01 1MS01 AA - 26* 1MSO4 BJ - 12*

  • 1MSO4 BP - 12* 1MSO4 BB - 12' * .

1MSO4 BV- 12*~ 1MSO4 AB - 8' l 1MSO4 AP - 8' 1MSO4 AJ - 8' 1MSO4 AV- 8* i i

1MS03 1MS01 AC - 26" 1MSO4 AC - 8" l 1MSO4 BC - 12" 1MSO4 AE - 8' 1MSO4 BE - 12" 1MSO4 AR - 8*

1MSO4 BR - 12* 1MSO4 AL - 8' 1MSO4 BL - 12" 1MSO4 AN - 8' 1MSO4 BN-12"

  • I 1MSO4 1MS01 AD - 26* 1MSO4 AA -8*

1MSO4 BH - 12* 1MSO4 AH - 8' 1MSO4 BU - 12" 1MSO4 AG - 8' 1MSO4 BA - 12'

  • 1MSO4 AU - 8*

1MSO4 BG - 12" *

  • These SRV tail pipes will be removed.

l The following information is applicable to the above lines:

Pipe Nominal Wall Thickness Operating Design / ASME Piping Size (inches) (inches) Temperature (*F) Operating Class Pressure (psi) 26 1.075 550 1250/1025 A i

12 0.406 480 600/540 B 8 0.906 550 1250/1025 A The above piping have been analyzed using PIPSYS computer program for the loads and load combinations in accordance with the UFSAR using PVRC dampings (per Code Case N411). In addition, the piping analyses and support assessments for the subsystems affected by the SRV removal include provisions for the future power uprate changes in pressure, temperature and main steam mass flow. Also, a three percent SRV set point tolerance has been considered in the piping analyses, based on a previous setpoint tolerance change. The piping stress analyses results for these Main Steam Lines are provided in Table 1 and Table 2 for Class 'A' and 'B' piping noted above.

i As shown in these tables, the stresses in the piping are within the allowables for the corresponding l ASME Code Equations.

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) NDIT LAS-ENDIT 0134, Upgraide 1 ,

Attachment A, Rev.1 l

! Page A4 I .

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The Main Steam Isolation Valves and Safety Relief Valves have been rereviewed for the new valve j

. accelerations and valve end loads. The necessary fatigue analyses have also been performed for 1 these valves. I j The welded attachments to the affected subsystem piping have been quaisfied for the increased loads.

Additionally, the new piping stress levels are such that no additional pipe breaks need to be j postulated.

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The increase in the deflecbon of the main header piping has been determined to be insignificant from J

the original design condition.

i Based on the results of the piping stress analyses, and using criteria per Note (1) (b) of 1989 Edition i

of ASME B & PV Code Section XI, Table IWB-2500 (Examination Joints Category B-J) additional inservice inspection (ISI) of the welds will be required at the following fittings at the ISI weld points ,

indicated below.

l

] Subsystems Node Points 181 Drawing Nos. New 181Wald Pts.

j Unit #1 Unit #2 Unit #1 Unit #2 2

1MS01,2MS01 225T MS-1001 MS-2001 9 8 i 1MS03,2MS03 110T,180T,250T MS-1003 MS-2003 14,18,22 13,17,21

l. 1MSO4,2MSO4 100T,225T MS-1004 MS-2004 8,10 7,15

.l j e Pipina Supports / Penetrations and Structural Assessment Results l Tables 3A and 38 provide results of the assessment of the Unit 1 and Unit 2 pipe supports for the load j increases due to reanalysis of the Main Steam Lines (MS01, MS03, MSO4) for the conditions

, discussed above. These tables show the available design rnargins for the auxiliary steel and standard j components for the increased loads. For the pipe supports where the loods have been reduced, the

! design margins have not been tabulated sina the design margins have increased due to SRV j removal. The pipe supports listed in Tables 3A and 3B were assessed based on the stress allowables  ;

in accordance with the UFSAR, and the design margins have been determined to be it 1.0.

i The structural framings (i.e., various galleries in the drywell, sacnficial sheid wall, etc.) have also been ,

reviewed for the increased pipe support loads due to SRV removal and have been determined to meet  !

the UFSAR allowables. Since these structural framings support other piping and components (i.e.,  ;

pipe whip restraints, conduits, etc.) which are not impacted by the SRV removal, they have not been included in the Design Margin Tables 3A and 38.

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NDIT LAS-ENDIT0134 Upgrade 1 l Attachment A. Rev.1  ;

Page A 5 l

  • Conclusion The Main Steam Header with the remaining SRV Tail Pipes, as documented in Tables 1,2,3A and 3B are acceptable from a piping stress stand point. The stresses in the piping, pipe supports and structural elements are within the allowable limits given in the UFSAR, and the design margins in the most stressed element are greater than or equal to 1.0.

The potential impact on the containment pool hydrodynamics loads due to reduction in the number of SRVs from the existing eighteen valves to future thirteen valves is part of General Electnc's w' o rk scope and is therefore not addressed in this report. Additionally, the evaluation of Main Steam Wetwell piping due to future power uprate changes etc. is also part of GE scope of work.

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l NDI LAS-ENDIT 0134, Upgrade 1 l Attachment A, Rev.1 l Page A 6 Table 1 Unit 1 & 2 Class 'A' Piping (Main Steam Header & 8" ( Tee) Highest Stress Summary l

Subsystem Piping Stress Actual Max. Allowable interaction Ratio Highest Usage identification Equation No. Piping Stress Stress Factor (psi) Limits (psi) Actual / Allow 1MS01 EQ. 9 21,500 26,550 0.810 0.357 2MS01 EQ. 9C 32,300 39,825 0.811 i EQ.10* 109,000 54,600 1.996 l

l EQ.12 48,200 54,600 0.883 l EQ.13 52,500 54,600 0.962 I

1MS03" EQ.9 26,300 27,300 0.963 0.701 2MS03" EQ.9C 35,600 40,950 0.869 EQ.10

  • 135,000 52,100 54,600 54,600 2.473  !:

EQ.12 0.954 EQ.13 54,177 54,600 0.992

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l 1MSO4 EQ. 9 22,000 26,550 0.829 0.173 2MSO4 EQ.9C 28,600 39,825 0.718 EQ.10* 103,000 54,600 1.886 EQ.12 36,300 54,600 0.665 EQ.13 45,900 54,600 0.841

  • lf actual stress for EON 10 is greater than 3 Sm EON 12 and 13 are evaluated.

" The actual stresses are the envelope of 1MS03 and 2MS03 Subsystem Analysis l

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NDIT LAS ENDIT-0134, Upgrade 1 Attachment A, Rev.1

, Page A 7 i !

i Table 2 l'

Unit 1 & 2 Class 'B' Piping (SRV Tall Pipes) Highest Stress Summary i

Subsystem Piping Stress Actual Max. Piping Allowsole Stress interaction Ratio r i i

identification Equation No. Stress (psi) Limits (psi) Actual / Allow i{

l 1MS01 EQ. 8 8,330 15,000 0.555 2MS01 EQ.9B 16,200 18.000 0.900 EQ.9C 17,500 27,000 0.648 .

EQ.10

  • 16,600 22,500 0.738 EQ.11 22.600 37,500 0.603 il 1MS03" EQ. 8 11,700 15,000 0.780 2MS03" EQ.98 18,000 18,000 1.000 EQ.9C 20,200 27,000 0.748 EQ.10
  • 25,500 22,500 1.133 EQ.11 31,800 37,500 0.848 1MSO4 EQ. 8 6,800 15,000 0.453 2MSO4 EQ.9B 14,600 18,000 0.811-EQ.9C 18,700 27,000 0.693 EQ.10
  • 19,400 22,500 0.862 EQ.11 24,600 37,500 0.656
  • If EON 10 actual stress is greater than allowables, EQN 11 is evaluated

" The actual stresses are the envelope of 1MS03 and 2MS03 Subsystem Analysis l j.l l

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NDIT LAS-ENDIT-0134, Upgrade 1 l Attachment A. Rev.1 Page A 8  ;

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Table 3A i  !

Unit 1-Pipe Support Load increase Assessment Results  !

Support References Piping References Structural Components Evaluaton Mechanical Components Evaluaton  !

{ ltem No. Pipe Support Number Subsystem Piping Auxiliary Limshng Aux. Steel

  • Standard Limiting Component i Number Analysis Steel Calc. Struct. Design Component Standard Design  !

Report No. No. Component Margin Calc. No. . Component Margen l 1 M09-MSO4-1508X 1MS01 L-000542 L-000540 Emb. Pl. 2.5 1MS01-37 Strut 3.81 2 M09-MSO4-1163X 1MS01 L-000542 L-000540 R. Sk. Wold 1.83 1MS01-28 Rear Bkt. 1.41 3 M09-MSO4-1546X 1MS01 L-000542 L-000540 Weld 1.2 L-000580 M. L. Clamp 1.07 4 M09-MS00-1006S 1MS01 L-000542 L-000540 Base Pl. 1.03 L-000542 Clamp 1.46 5 M09-MS00-1024S 1MS01 L-000542 L-000540 W eld 5.87 L-000542 Clamp 2.79 6 PENET -1DF-1 1MS01 L-000542 L-000540 W eld 1.1 N/A N/A N/A i 7 PENET -1DF-6 1MS01 L-000542 L-000540 W eld 1.38 N/A N/A N/A 8 PENET - 1M-1 1MS01 L-000542 L-000540 Lugs 5.3 L-000573 Fl. Hd. Weld 1.006 9 M09-MSO4-1298X 1MSO4 L-000521 L-000540 Channel 1.26 EMD-062407 Clamp 1.09 10 M09-MSO4-1302X 1MSO4 L-000521 L-000540 W eld 3.11 1MSO4-38 Strut 1.53 11 M09-MS00-1017S 1MSO4 L-000521 L-000540 1.5" Plate 1.08 L-000521 Clamp 1.33 12 M09-MS00-1055S 1MSO4 L-000521 L-000540 Angle 4x4 1.27 L-000521 Snubber 1.36 13 M09-MS00-1054S 1MSO4 L-000521 L-000540 Weld 4.64 L-000521 Clamp 1.73 14 PENET 1DF-10 1MSO4 L-000521 L-000540 Weld 1.97 N/A N/A N/A-15 PENET 1DF-15 1MSO4 L-000521 L-000540 W eld >1.0 N/A N/A N/A 16 PENET 1M-4 1MSO4 L-000521 L-000540 Lug 7.8 L-000574 Fl. Hd. Weld 1.103 17 M09-MS00-1040X 1MS03 L-000541 L-000540 R. BK. Weld 4.61 L-000541 - Strut 2.18 18 M09-MSO4-1235X 1MSO3 L-000541 L-000540 W eld 4.95 1MS03-19 Strut 1.17 19 M09-MSO4-1366X 1MS03 L-000541 L-000540 Channel 1.67 1MS03-68 R. Bkt. 1.22 20 M09-MS00-1016X 1MS03 L-000541 L-000540 2" Pl. 1.34 L-000541 Strut 1.08 21 M09-MSO4-1251X - 1MS03 L-000541 L-000540 W eld 1.1 L-000580 M. L Clamp 1.67

_._.__.____._m _ _ _..__. ___m.mm.-__m.____. ___m.__,._._-___z_m____m_. _ _ _ _ _ _ _ _ . _ _ _ _ . _ _____M-- -_ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . - - _ < _ _ _ M

NDIT LAS-ENDIT-0134. Upgrade 1 Attachment A, Rev.1 ,

Page A 9 Table 3A Unit 1-Pipe Support Load increase Assessment Results Support References Piping References Structural Components Evaluation Mechanical Components Evaluatien item No. Pipe Support Number Subsystem Piping Auxiliary Limiting Aux. Steel

  • Design Margin =

Actual Stress Y N-

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NDIT LAS-ENDIT-0134, Upgrade 1 Attachment A, Rev.1 Page A 10 Table 3B Unit 2-Pipe Support Load increase Assessment Results Support References Piping References Structural Components Evaluation Mechamcal Components Evaluation S.No. Pipe Support Number Subsystem Piping Auxiliary Linuting Aux.* Standard Limiting Component Number Analysis Steel Calc. Struct. Steel Component Standard Design Report No. No. Component Design ' Calc. No. Component Margin Margin 1 M09-MSO4-2639X 2MS01 L-000594 L-000584 W eld 1.0 L-000617 Rear Bkt. 1.37 2 M09-MSO4-2627X 2MS01 L-000594 L-000584 R. BK. Wold 1.51 L-000617 Rear Bkt. 1.41 3 M09-MSO4-2645X 2MS01 L-000594 L-000584 W eld 1.20 L-000580 M. L Clamp 3.21 4 M09-MS00-2006S 2MS01 L-000594 L-000584 Base Pl. 1.03 L-000594 Clamp 1.46 5 M09-MS00-2024S 2MS01 L-000594 L-000584 W eld 5.90 L-000594 Clamp 2.79 6 M09-MS00-2009S 2MS01 L-000594 L-000584 TS 6x6 1.19 L-000594 - Clamp 2.27 7 PENET 2DF-1 2MS01 L-000594 L-000584 Weld 1.1 N/A N/A N/A 8 PENET 2DF-6 2MS01 L-000594 L-000584 Weld 1.38 N/A N/A N/A 9 PENET 2M-1 2MS01 L-000594 L-000584 Lugs 5.3 L-000585 Fl. Hd. Weld 1.006 10 M09-MSO4-2852X 2MSO4 L-000596 L-000584 R. BK. Wold 2.43 2MSO4-38 Rear Bkt. 1.02 11 M09-MS00-2017S 2MSO4 L-000596 L-000584 W eld 1.06 L-000596 Clamp 1.33 12 M09-MS00-2055S 2MSO4 L-000596 L-000584 Angle 4x4 1.13 L-000596 Snubber 1.36 13 M09-MS00-2054S 2MSO4 L-000596 L-000584 R. BK. Wold 4.89 L-000596 Snubber 1.73 14 M09-MSO4-2848X 2MSO4 L-000596 L-000584 Channel 1.25 L-000596 Strut 1.23 15 PENET 2DF-15 2MSO4 L-000596 L-000584 Weld >1.00 N/A N/A N/A 16 PENET 2DF-10 2MSO4 L-000596 L-000584 W eld 1.97 N/A N/A N/A 17 PENET 2M-4 2MSO4 L-000596 L-000584 Lug 5.3 L-000586 Fl. Hd. Weld 1.12 18 M09-MS00-2039X 2MS03 L-000595 L-000584 Studs ~ 1.67 L-000595 Strut 1.51 19 M09-MS00-2040X 2MS03 L-000595 L-000584 R. BK. Weld 4.25 L-000595 Strut 1.02 20 M09-MSO4-2775X 2MS03 L-000595 L-000584 R. BK, Wold 1.64 L-000616 Strut 1.13 21 M09-MSO4-2728X 2MS03 L-000595 L-000584 Channel 1.06 L-000616 Clamp 1.83 .

NDIT LAS-ENDIT-0134, Upgrade 1 Attachment A. Rev.1 Page A 11 Table 3B Unit 2-Pipe Support Load increase Assessment Results Suppcit References Piping References Structural Components Evaluation Mechanical Components Evaluation l S.No. Pipe Support Number Subsystem Piping Auxiliary Limiting Aux.* Standard Limiting C0rngrent Number Analysis Steel Calc. StrucL Steel Component Standard Design

Report No. No. Component Design Calc. No. Component Margin

! Margin l

l 22 M09-MS00-2016X 2MS03 L-000595 L-000584 Cover Pl. 1.16 L-000595 Strut 1.09 23 M09-MSO4-2801X 2MS03 L-000595 L-000584 Weld 2.30 L-000616 Clamp 1.24 24 M09-MSO4-2750X 2MS03 L-000595 L-000584 R. BK. Weld 1.16*" L-000595 Pin 1.01 "

25 M09-MS00-2043X 2MS03 L-000595 L-000584 R. BK. Weld 6.49 L-000595 Clamp 2.35 26 M09-MS00-2014S 2MS03 L-000595 L-000584 Weld 2.0 L-000595 Clamp 4.97 27 M09-MS00-2015S 2MS03 L-000595 L-000584 1.5" Plate 3.20 L-000595 Clamp 2.16 28 M09-MS00-2045S 2MS03 L-000595 L-000584 E m b. Pl. 1.18 L-000595 Clamp 1.76 29 M09-MSO4-2749S 2MS03 L-000595 L-000584 Channel 1.95 L-000615 M. L. Clamp 1.07" 30 M09-MSO4-2804S 2MS03 L-000595 L-000584 1.5" Plate 1.18 L-000615 M. L. Clamp 1.08 31 PENET 2M-3 2MS03 L-000595 L-000584 Lug 10.54 L-000592 Fl. Hd. Weld 1.10 32 PENET 2DF-17 2MS03 L-000595 L-000584 W eld 1.12 N/A N/A N/A 33 PENET 2DF-11 2MS03 L-000595 L-000584 Weld 1.12 N/A N/A N/A 34 PENET 2DF-13 2MS03 L-000595 L-000584 Weld 1.12 N/A N/A N/A 35 PENET 2DF-12 2MS03 L-000595 L-000584 Weld 1.12 N/A N/A N/A Allowable Stress

' Design Margin ==

Actual Stress

" Reinforcing of standard component, will be required to achieve the noted Design Margin.

"* Weld size increased to 1/2" to achieve the noted Design Margin.

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