branch lioensee ~p'i ~t,licensing projects branch 8'1 division of project management u.s....

REGULATO, INFORMATION DISTRIBUTION TEM (RIDS)

ACCESSION NBR:8102270093 DOCsDATE! 81/02/23 NOTARIZED: NO, 'OCKETFACIL:50 387 Susquehanna Steam Electric Stationi Unit ig Pennsylva 05000387

50-388 Susquehanna Steam Electric Stationi Unit 2i Pennsylva 05000388AUTH'AME AUTHOR AFFILIATION

CURTISEN ~ HE Pennsylvania Power L Light Co'.RECIP ~ NAME RECIPIENT AFFILIATION

YOUNGBLOODiB~ J, Licensing Branch 1

SUBJECT: forwards responses to Lioensee Review Group positions reffacility~ Status of positions should be discussed at NRC ~p'i,810318 meeting. ~t,

DISTRIBUTION CODE: BOOIS COPIES RECEIVED:LTR +ENCL + SIZE:TITLE'. PSAR/FSAR AMDTS and Related Correspondence

NOTES:Send ILE 3 copies FSAR L all amends.Send ILE 3 copies FSAR L all amends,

0500038705000388

RECIPIENTID CODE/NAME.

ACTION: A/D LICENSNGRUSHBROOKiM ~

INTERNAL: ACCID EVAL BR26CHEM ENG BR 08CORE PERF BR 10EMERG PREP 22GEOSC IENCES 10HYD/GEO BR 15ILE 06LIC QUAL BRMECH ENG BR 18NRC PDR 02OP LIC BRPROC/TST REV 20RAD SESS BR22.

F 01BR25

COPIESLTTR ENCL

01 0

1 1

1 1

1, 1

1 0

1 1

2 231

1 1

1 1

1 1

1 1

1 1

1 1

1 1

RECIPIENTID CODE/NAME

YOUNGBLOODRBSTARKgR ~ 00

AUX SYS BR 07CONT SYS BR 09EFF TR SYS BR1?EQUIP QUAL BR13HUM FACT ENG BRILC SYS BR 16LIC GUID BRNATL ENG BR 17MPAOELDPOWER SYS BR 19QA BR 21REAC SYS BR 23SIT ANAL BR 24SYS INTERAC BR

COPIESLTTR ENCL'

01 1

1 1

1 1

1 1

3 3-

1 1

1

1 1

1 1:1 01 01 1

1 1

1 1-1 1

1 1

EXTERNAL: ACRSNSIC

2705

16 161 1

LPDR 03 1 1

SAR 0g )gg)

TOTAL NUMBER OF COPIES REQUIRED: LTTR 57 ENCL'1

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TWO NORTH NINTH STREET, ALLENTOWN, PA. 18101

NORMAN W. CURTISVice President-Engtneering 8, Construction-Nuclear770 5381

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PHONEr (215) 770.51511

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February 23, 1981

Mr. B. J. Youngblood, ChiefLicensing Projects Branch 8'1

Division of Project ManagementU.S. Nuclear Regulatory CommissionWashington, D.C. 20555

Docket No. 50-38750-388

SUSQUEHANNA STEAM ELECTRIC STATIONLRG POSITIONSER 100450 FILE 841-2PLA-642

Dear Mr. Youngblood:

Attached are Pennsylvania Power & Light Company's responses to the LicensingReview Group's positions. We would like to discuss the status of thesepositions with your personnel during the next LRG meeting. This meeting isscheduled for March 18, 1981.

If you have any questions, please call me.

Very truly yours,

N. W. CurtisVice President-Engineering & Construction-Nuclear

CTC/mks

Attachment

cc: R. M. Stark — NRCgoo I

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PENNSYLVANIA POWER 8 LIGHT COMPANY

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'RGPOSITION

ISSUE

RSB-1 Internall Generated Missiles — (3.5.1)

The applicant has not supplied the information to show that all safety-related systems and components within containment, including the con-tainment, are protected from missiles.

With regard to missiles sizes of concern, what is the valve size belowwhich, if failure should occur in a high pressure system, damage toother components within the primary containment would not be significant?State the criteria used to determine this size. Identify all valves inthe primary containment larger than this size and identify the missileprotection provided for each valve (either physical location or barrier).

POSITION (Unique)

Each applicant, on a plant specific basis, will demonstrate accept-ability using one of, or a combination of, the following:

Provide protection from internally generated missiles.

2. Perform Analysis to show that missiles are not generated,or, if generated, have, insufficient energy to cause un-acceptable damage.

This item relates to ACRC generic concern II-8, recirculation pumpoverspeed during a LOCA.

SUS UEHANNA POSITION

Position 2 above has been implemented as described in FSAR Subsections3.5.1.1 and 3.5.1.2. A mathematical analysis and calculation resultswere supplied in response to Question 211.195. Other relevant questionsinclude: 211.2, 211.36, 211.37, 211.38, 211.39, 211.194, 211.196 and211. 197.

JRM/054517

810227P49

LRG POSITION

ISSUE

RSB-2 Control Rod S stem — (4.6.2)

As a result of eliminating the control rod drive system return line, weare reviewing generically with regard to the impact on control rod drivesystem performance. Consequently, we require the applicant submitsystem performance data directly applicable to LaSalle and will requirethe applicant to conform to the conclusion of the generic study asapplicable to LaSalle.

POSITION (Common)

1. Remove the CRD return line and cap the vessel nozzle. Accept-ability is demonstrated by GE analysis of CRD performancecharacteristics.

2. The CRD line is left routed to the vessel. The line is admin-istratively closed by valving it out during operation. CRD

performance must be demonstrated.

3. The CRD return line is rerouted to another vessel entry point(e.g. feedwater line). CRD performance is unchanged, assumingacceptable return path established. IGSCC is corrected by thereroute.

NOTE: 1. The LRG members acknowledge the receipt of the NRC letter"modifications to Boiling Water Reactor Control Rod DriveSystems" in March, 1980 and will factor the concerns of theletter into their documented positions.

2. GE Letter reports dated March 14, 1979 (G.G. Sherwood to V.Stello/R. Mattson) and May 2, 1980 (R. Gridley to 0. Eisonhut)address all NRC concerns for vessel makeup capability.


The acceptability of position 1 above for Susquehanna SES is discussedin the response to Questions 211.43 and 211.192. System performancedata will be provided.

JRM/321657

LRG POSITION

ISSUE

RSB-3., Safet /Relief Valves — (5.2.2 and 6.3.2)

Additional information is required both for qualification tests andoperating experience with the applicant's safety/relief valves.

POSITION (Common)

1. Provide evaluation and operating history. See Owners Groupresponse to position 2.1.2 of NUREG-0578.

2. Participate in TMI Qualification Program (II.D.l)


o A detailed evaluation of the qualification tests is provided inresponse to Question 211.70.

o Details for the surveillance and -testing of safety relief valvesare included in the technical specifications and will be includedin the pump and valve in-service inspection submittal.

o Operating history is discussed in response to Question 211.221.

JRM/746719

LRG POSITION

ISSUE

RSB-4 Tri of Recirculation Pum s to Miti ate ATWS - (5.2.2)

We require reperformance of the over pressure protection analysis toconsider the effect of the ATWS RPT.

POSITION-

The over pressure protection report will be resubmitted to demonstratecompliance with the ASME Pressure Vessel Code, considering theaddition of ATWS RPT.

This position was acceptable on the Zimmer docket as noted byofficial NRC meeting minutes for a March 28, 1979 meeting.This item is resolved for Zimmer.


The Susquehanna SES position is stated in response to Question 211.71 and 211.4.

2A

LRG POSITION

ISSUE

RSB-5 'etection of Inters stem Leaka e - (5.2".5)

Ve requested that'he applicant show how it intends to detect leakagefrom the reactor coolant systems into both the low pressure coolantinjection (3 trains) and low pressure core spray systems as required byRegulatory Guide 1.45.

POSITION (Unique)'he

LRG members will demonstrate conformance to Regulatory Guide1.45(4) with reliance on level instrumentation, pressure instruments,and/or radiation monitors..


The Susquehanna SES position is stated in response to question 211.32.

LRG POSITION

ISSUE

RSB-6 Reactor Core Isolation Coolin Pum Suction - (5.4.1)

The applicant must supply further information to determine whether theRCIC pump suction has to be automatically switched from the condensate"storage tank to the suppression pool in the event of .a safe shutdownearthquake and concomitant failu're of the condensate storage tank.

POSITION (Common)

The qualification of RCIC will be resolved by insuring the availabilityof a Seismic Category I water source. This" will be achieved'by:

1. Provide a seismic category 1 CST, or

2. Provide an automatic switchover to the suppression pool, or

3. Provide justification for a manual switchover to the suppressionpool.

NOTE:

a) This position satisfies the requirement identified in NUREG-0660item II.K.3(C.3.22).

b) It is not an OL item in accordance wth NUREG 0694. Compliancewith C.3.22, verify procedures, will be accomplished prior tofuel load on Jan. 1, 1981, whichever is later. Compliancewith design modification will be accomplished by Jan. 1, 1982.

c) Also resolves issue ICSB-7.


Susquehanna SES will be provided with an automatic switchover from thecondensate storage tank to the suppression poolby January 1, 1982.

AAW/3C

LRG POSITION

ISSUE

RSB-7 Shutdown Unintentionall of the Reactor CoreIsolation Coolin S stem — (5.4.1)

"Show how the design of the RCIC protection system prevents uninten-tional shutdown of the system, when the system is required, because ofspurious ambient temperature signals from areas in and around thesystem (especially in the RCIC pump room)."

POSITION (Common)

1. The temperature alarm setpoint will be established by calcu-lating a heat balance for the normal room environment, andthen introducing the heat release cause by an alarm limitleak. Actual ambient temperature settings will be determinedduring startup testing.

NOTES:

1. CECO's notes of LaSalle/NRC Meeting of October ll, 1979 item3.i indicate that LaSalle response to Q212.130 was satisfactory.

2. NUREG-0660 Task II.K.3 item C.3.15 is related to the issuediscussed above. Although this NUREG-0660 is not a require-ment for licensing of the LRG plants (Ref. NUREG-0694), theBVR (TMI) Owners Group is addressing that task and the utili-ties represented by the LRG are participating in that effort.


In response to Question 211.33, Subsection 5.2.5.1.3 was revised to showthe trip setpoints used to prevent unintentional isolation of the system.See also the response to Question 211.227.

JRM/081046

LRG POSITION

ISSUE

RSB-8 Residual Heat Removal S stem (5.4.2)

The applicant must perform tests to show that flow through the safety/relief valves is adequate to provide the necessary fluid relief requiredconsistent with the analyses reported in Section 15.2.9 of the FSAR.

POSITION (Common)

1. Crosby Valves, e.g. LaSalle — Demonstrate by analysis (basedon valve characteristics such as internal dimensions) that thevalve(s) will pass sufficient water to allow a forced cooldownto be maintained.

2. Target Rock (2 stage), e.g. SHOREHAM — Demonstrate by shoptest of a valve that (1) the valve can be opened by lowpressure water and (2) once open is maintained open andpasses sufficient flow to maintain a forced cooldown.

3. Provide a backup to RHR shutdown cooling which does notutilize the SRV in the alternate cooldown mode.

NOTE: NUREG-0660 Task II.D.l is related to the issue discussedabove. This task is identified as a requirement for licensingof LRG plants (Ref: NUREG-0699). The BWR (TMI) Owners Groupis addressing that task and the utilities represented by theLRG are participating in this effort.


This item was addressed in the response to Question 211.8.

JRM/321405

LRG POSITION

ISSUE

RSB-9 ~ Cate orization of Valve which Isolate RHRfrom Reactor Coolant S stem — (5.4.2)

We require that the valves which serve to isolate the residual heatremoval system from the reactor coolant system be classified categoryA/C in accordance with the provisions of Section XI of the ASME code.

RSB-13 Leaka e 8 Testin of Valves Used to IsolateReactor Coolant S stem — (6.3.2)

We require periodic testing arid, establishment of leak rate criteria forthe valves that isolate the reactor coolant system from all the emergencycore cooling system.

POSITION (Common)

Containment isolation valves which also provide isolation betweenhigh and low pressure systems will be tested in accordance withASME Section XI as well as Appendix J to 10CFR50. Exemption fromASME Section XI requirements will be requested if it can be demon-strated the requirement is met on an alternate basis.

NOTE: Exemptions will be requested


These items will be discussed in pump and valve ISI submittal. See alsothe responses to Questions 211.56 and 211.97.

JRM/002065

'IRG POSITION

ISSUE

RSB-10 Available Net Positive Suction Head - (6.3.2)

The applicant must verify that the suction lines in the suppression poolleading to the ECCS pumps are designed to preclude adverse vortex formationand air injection which could effect the pumps performance.

POSITION (Unique)

The issue is regarded as plant unique and requires plant specific reviewof system design details.

SUSQUEHANNA POSITION

Adequacy of the design to prevent adverse vortex formation 'is verifiedduring preoperational testing. See Subsection 6.3.6 and question 211.214.

AAM/3D

LRG POSITION

IMUE

RSB-ll Assurance of Filled ECCS Line - (6.3.2)

Instrumentation is not sufficiently sensitive to detect voids at the topof ECCS pipe lines. The applicant must provide adequate instrumentationto assure filled ECCS lines.

POSITION (Unique)

1. Jockey pump system on same division as system being filled.2. Pressure and/or flow switch on pump discharge with control room

annunciation.

3. Tech. Spec. surveillance.

NOTE: This position is applicable to LSCS, ZPS, SNPS, and WPPSS.Because of design differences plant unique review is requiredon Fermi-2 and SSES dockets.


A reliable discharge line fillsystem is provided as described in Subsection6.3.2.2.5. Thzs subsection.has been amended in response to questions211.58, 211.102, 211.212 and 211.218. Adequacy of the system is shownby alarms (as described in Subsection 6.3.2.2.5) and by surveillancetesting (as described 'in the technical specifications).

AAW/3G:1

LRG POSITION

ISSUE

RSB-12 0 erabilit of ADS - (6.3.2, 5.4.2)

The applicant must show that the air supply for the ADS is sufficientfor the extended operating time required and assures us by reliabilitydata that the ADS valves will function as required.

POSITION (Common)

The applicant will review and document that the ADS system designsatisfies the expressed concern.

NOTES:

a) The ADS system for Zimmer which is a typical design hasbeen accepted by the NRC.

b) CECO's La Salle/NRC meeting notes of October ll, 1979item 3L indicate issue as closed. La Salle's ADS systemis identical to Zimmer's.

c) NUREG-0660 Task II.K.3 item C.3.28 is related to theissue discussed above. This task is not identified as arequirement for licensing of LRG plants (Ref: NUREG-0694).The BWR(TMI) owners group is addressing that task and theutilities represented by the LRG are participating inthat effort.


A non-interruptible safety grade source of gas is supplied to the ADSvalves as described in the response to Question 211.67 and in Subsection9.3.1.5.

NLF2I

:iLRG POSITIONS

ISSUE

RSB-14 0 erabilit of ECCS Pum s - (6.3.2)

The applicant must provide assurance that the ECCS pumps can functionfor an extended time (maintenance free) under the most limiting post-LOCAconditions.

POSITION (Common)

This issue has been closed on Zimmer and Shoreham dockets on the basisof information presented in response to NRC questions. Similar informationhas been provided on the rest of the dockets.

NUREG-0660 Task II.B.2 is related to the issue discussed above and isbeing addressed by the BVR TMI Owners Group.


Operating history of ECCS pumps was provided in response to question211.106. This is the same data used to address the question on theLaSalle docket.

AAM/3G:2

-LRG POSITION

ISSUE

RSB-15 Additional LOCA Break S ectrum - (6.3.4.)

The applicant was requested to submit two additional LOCA analyses tocomplete the La Salle break spectrum. These were as follows:

a) The design basis accident with a discharge coefficient of 0.6, and

b) A small break analysis for a recirculation break of 0.02 squarefeet.

POSITION (Common)

Lead plant bounding .approach valid and confirmed by all additionalanalyses performed (Ref LaSalle Q212.125). .No additional plant specificanalysis will be provided other than that already provided by each plantfor the limiting failure/break combination.

'USQUEHANNA POSITION

The Susquehanna SES position is stated in response to Question 211.107.

AAW/3G:3

LRG POSITON

ISSUE

The applicant analyzed a coincident instantaneous closure of a flowcontrol valve during a LOCA which resulted in a 300 F increase in peak0

clad temperature. We requested that this accident be reevaluated con-sidering more realistic valve closure dynamics.

POSITION (Common — BWR/5 Only)

No additional analysis is required because the failure describedrepresents a non-mechanistic failure mode. In reality, for allfailure modes except valve hydraulic system failure, the valvefails as is, which is physically limited to at least 20% open. Ifthe hydraulic system fails, the valve will dri.ft closed, but at adesign rate of no more than 10+ 1% per second.

. SUS UEHANNA POSITION

Susquehanna SES does not use a recirculation system flow control valve.=Thus this question is not applicable.

JRM/058261

LRG POSITION

ISSUE

RSB-17 0 erator Action — (6.3.4)

The applicant must show that adequate time is available for operatoraction to restore core cooling prior to excessive core, heating as aresult of a crack in residual heat removal line.

Anal sis of Crack in The RHR Line — (6.3.4)

In the applicants analysis to evaluate a crack in the residual heatremoval line that was postulated to occur during normal shutdown cooling,operator action is indicated to restore core cooling. We require theapplicant to show that adequate time is available for operation action.

POSITION (Common)

Should the RHR shutdown cooling line crack during the normal shutdown,a total reactor isolation will automatically occur. Subsequentlyvessel water Level 2 will be reached and automatic initiation ofHPCS/HPCI will occur. HPCS/HPCI will cycle on and off betweenLevels 3 and 8 until the operator establishes an alternate watersource.

If HPCS/HPCI is unavailable, representative analyses have beenperformed to demonstrate that operator action would not be requiredbefore 20 to 30 minutes following the pipe crack (depending uponplant design) to assure adequate core cooling in accordance withthe acceptance criteria of lOCFR50.46.



Jj91/954626

LRG POSITION

ISSUE

RSB-18 Diversion of Low Pressure Coolant In ection S stem — (6.3.4)

Low pressure coolant injection flow can be diverted to wetwell anddrywell spray and suppression pool cooling. The applicant must demonstratethat adequate core cooling is maintained when diversion is considered.

POSITION (Common)

The limiting break will be reviewed with LPCI diversion after 600sec to confirm that the peak clad temperature limit is not exceededand adequate core cooling is maintained. Such analyses have beenperformed for Zimmer and LaSalle (BWR-5) and Shoreham (BWR-4).These'analyses confirm the acceptability of LPCI diversion at 600sec. The applicability of these results to the other LRG plantswill be verified.


The Susquehanna SES position is stated in the responses to Questions211.105 and 211.228.

JRM/002057

LRG POSITION

ISSUE

RSB-19 Failure of the Feedwater Controller — (15.1)

The applicant's analysis for the failure of the feedwater controllerindicates that the temperature drop is no greater than 100 F. At a0

domestic boiling water reactor an actual feedwater temperature occurred0which demonstrated a temperature difference of 150 F. The applicant

must justify the decrease in temperature drop used'for this event orrecalculate the transient by using a )ustified temperature decrease toassure conformance with applicable criteria.

POSITION (Common)

Analyses have been performed on the. lead plant assuming a hT =,150 F.(which bounds observed operating experience) for the LFWHevent. LaSalle (BWR-5) has completed this calculation (see Amendment49) confirming the conclusion that the results are insensitive toAT assumed and is not a limiting event. The applicability of theseresults to the other LRG plants will be verified.


The Susquehanna SES feedwater system is designed so that no singlefailure can cause the loss of more than one feedwater heater string.Loss of one heater string results in a temperature drop of less than100 F. See response to Question 211.116.

It has been shown in the response to Question 211.148 that based onanalysis for a similar plant a temperature drop of 150 F is acceptable.

JRM/000967

4

4 LRG POSITION

4

4

P

4

4

ISSUE

RSB-20 Use of Nonreliable E ui ment in Antici ated0 erational Transients — (15.1)

In analyzing anticipated operational transients, the applicant tookcredit for equipment which has not been shown to be reliable. Ourposition is that this equipment be identified in the technical specifi-cations with regard to availability, setpoints and surveillance testing.The applicant must submit its plan for implementing this requirementalong with any system modification that may be required to fulfilltherequirement.

POSITION (Common)

This issue is related to a long time concern by the NRC Re: theuse of non-safety grade equipment to mitigate transients. Manyquestions were asked on many dockets relative to:

a) Credit for Non class lE relief function vs. 1E safetyfunctions setpoints.

b) Credit for RPS inputs from the turbine bldg.

c) Credit for Level 8 turbine trip and turbine bypass system.

As a result of this concern GE and the NRC met (Nov. 78) for a comprehen-sive review of all such transients and as a result of that meeting,determined the most limiting "event" which takes credit for non-safetygrade equipment was the excess feedwater transient relying on the L-8turbine trip and turbine bypass. NRC concurred that providing technicalspecifications for the L-8 trip and the turbine bypass valves satisfactorilyresolves this issue. All LRG plants commit to this except Fermi-2.



JRM/056796

LRG POSITION

ISSUE

RSB-21 Use of Non-Safet Grade E ui ment in Shaft SeizureAccident — (15.2)

The applicant included the use of non-safety grade equipment in hisanalyses for shaft seizure and shaft break accidents. We require thatthese accidents be reanalyzed without allowance for the use of non-safety grade equipment.

POSITION (Common)

The evaluation basis for the subject event was reviewed andaccepted on the Zimmer docket (SER 15.2).

2. A qualitative assessment of the event without reliance on nonsafety grade equipment suggests that the applicant's conclusion(i.e. the event is bounded by the DBA LOCA event) would notchange.

3. The non-safety grade equipment in question have been reviewedin item RSB-21 and steps have been taken to improve theirreliability.

4. No further analysis is required.


The Susquehanna SES position is stated in the response to Question211.120.

JRM/056597

LRG POSITION

ISSUE

RSB-22 ATWS — (15.2. 1)

We require that the applicant agrees to implement plant modifications ona scheduled basis in conformance with the Commission's final resolutionof ATWS. In the event that LaSalle starts operation before necessaryplant modifications are implemented, we require some interim actions betaken by LaSalle in order to reduce, further, the risk from ATWS events.The applicant will be required to:

1) 'Develop emergency procedures to train operators to recognize anATWS event, including consideration of scram indicators, rod positionindicators, flux monitors, vessel level and pressure indicators,relief valve and isolation valve indicators, and containment temperature,pressure, and radiation indicators.

2) Train operators to take actions in the event of an ATWS includingconsideration of immediately manual scramming the reactor by usingthe manual scram buttons followed by changing rod scram switches tothe scram position, stripping the feeder breakers on the reactorprotection system power distribution buses, opening the scramdischarge volume drain valve, prompt actuation of the standbyliquid control system, and prompt placement of the RHR in the poolcooling mode to reduce the severity of the containment conditions.

POSITION (Common)

ATWS RPT plus proc'edure will be provided by Fuel Load.

NOTE: All LRG applicants are willing to implement alternate 2on a best effort schedule as the basis for final resolutionof the ATWS issue.



JRM/056598

LRG POSITION

ISSUE

RSB-23 Peach Bottom Turbine Tri Tests — (4.4.1) (4.4.2)

These tests must be evaluated and assessed using the ODYN computer code.We have not completed our review of the ODYN Code.

POSITION (Common)

The LRG members agree to perform ODYN calculation(s) for the limitingpressure transients utilizing option B (GE/NRC generic resolutionin process — estimated resolution summer 1980) of the NRC letterdated 1/23/80. LaSalle (lead plant) schedule for completion ofcalculation is October 1980.

NOTE: REDY is still an acceptable code for other transients.


The Susquehanna SES position is that the limiting pressure transients willbe analyzed by using ODYN.

JRH/056599

LRG POSITION

: ISSUE

RSB-24 MCPR — (4. 4.1) (4 ~ 4.2) (15 1)

After completion of overpressure analysis, the minimum critical powerratio must be recalculated taking into consideration the turbine tripwithout bypass event.

The transient of generator load rejection without bypass results in anMCPR equal to 1.02 which is below the Safety limit of 1.06. The applicantclassified this event an infrequent occurrence which would allow some

fuel damage. We do not concur with this classification for this event,and we require that the operating limit be modified to satisfy the MCPR

limit of 1.06.

POSITION (Common)

The LRG members agree to perform ODYN calculations for the limitingpressure transients utilizing option B (GE/NRC generic resolutionin process — estimated resolution summer 1980) of the NRC letterdated 1/23/80. LaSalle (lead plant) schedule for completion ofcalculation is Oct. 1980.

Note that we understand REDY is still an acceptable code for othertransients.


The Susquehanna SES position is that the limiting pressure transients will be

analyzed using ODYN.

JRM/096718

-LRG POSITION

ISSUE

RSB-25 GEXL Correlation — (4.4.1)

Although we conclude that the GEXL correlation is acceptable for initialcore load, we are concerned that GEXL correlation may not be conservativefor reload operation. The applicant, in a letter dated March 7, 1979,committed to incorporate the latest approved form of GEXL correlation atthe time of reload for LaSalle. License Condition.

POSITION (Common)

1. For first core cycle, GEXL is conservative. Adequate negativeworth is provided by the control system to assure shutdowncapability.

2. The applicant(s) commit to incorporate the latest approvedform of GEXL correlation at the time of reload.


The Susquehanna SES position is stated in response to Question 221.3.Also Pennsylvania Power and Light Company will incorporate the appropriatecorrelation at the time of reload.

JRM/056602

LRG POSITION

ISSUE

RSB-26 Stabilit Evaluation — (4.4.1)

In order to provide additional margin to stability limits, natural cir-culation operation of LaSalle will be prohibited until our genericreview of hydrodynamic stability characteristics is -completed. Inaddition, the LaSalle stability analysis was performed for firstcycle. "

We will require that a new analysis be submitted and approvedprior to second cycles operation. The applicant, in a letter'datedMarch 7, 1979, committed to perform, as part of the future reload analysesto update the hydrodynamic stability analyses. License condition.

POSITION (Common)

Sufficient documentation of stability margin for Cycle 1 has beenpresented for issuance of an operating license. The stabilitymargin analysis will be updated as required for future reloadapplications.

This issue is addressed under generic NRC task action plan (TAP) B-19.


The stability margin analysis will be updated as required for reloadapplications.

JRM/4783

LRG POSITION

ISSUE

ICSB-2 Ph sical Se aration and Electrical Isolation—(7.1.4, 7.2.3 and 7.6.3)

In the applicant's design, class 1E instrumentation do not adhere toadequate separation criteria, have not been qualified, and do notadhere to separation of Class lE to non-class 1E instrumentation.

POSITION (Unique)

It is judged that this review should be made on a plant specificbasis.

Regulatory Guide 1.75 is not applicable to any of the plants in theLRG. The degree of conformance to this Regulatory Guide has beenaddressed on each docket and has in the case of the Zimmer docketbeen accepted (SER 8.1.2 and 8.3.3)


Section D, "Implementation", of Regulatory Guide 1.75, Revision 2 clearlystates that the guide is not applicable to Susquehanna SES vintageplants. The electrical separation criteria used for the Susquehanna SESdesign is described and compared with the requirements of RegulatoryGuide 1.75 in Sections 3.12 and 8.1.6. See also the response to Question32.49.

JRM/042654

LRG POSITION

ISSUE

ICSB-3 ATWS

POSITION (Common)

See Issue RSB-22


See Issue RSB-22,

JRM/334500

LRG POSITION

ISSUE

ICSB-4 Test Techni ues — (7.1.4)

In order to perform routine surveillance testing, it is necessary forthe applicant to pull fuses. We consider that this design does notsatisfy the requirements of IEEE Std 279-1971 Paragraphs 4.11 and 4.20.

POSITION (Unique)

This issue is considered to be plant specific and will not beaddressed by LEG.


Compliance with these paragraphs of IEEE 279-1971 is discussed in responseto Question 032.33 and in Chapter 7.

JRM/140868

LRG POSITION

ISSUE

ICSB-5 Safet S stem Set pints — (7.1.4)

The range of class lE system sensors may be exceeded in the worst casecombination of setpoint and accuracy.

POSITION ,(Unique)

This issue is considered to be plant specific and will not beaddressed by LRG.

Technical justification for the setpoints and allowable values areexpected to be made available during technical specification review.


The Susquehanna SES technical specifications have been submitted and areunder review.

JIBf/951600

LRG POSITION

ISSUE

ICSB-6 ~Drawdn a — (7.1.4, 7.3.3 and 7.6.3)

The one line drawings and sch'ematics contradict the functional controldrawings and system description which are provided in the PSAR. Further-more, contact utilization charts contradict the actual schematics.

POSITION (Unique)

This issue is plant unique;


Pennsylvania Power 6 Light Co. is in the process of reviewing the Susque-hanna SES drawings to identify the necessary corrections.

LRG POSITION

ISSUE

ICSB-7 RCIC Classification

RCIC should be classified safety grade.

POSITION (Common)

See issue RSB-6.


See RSB-6 and Subsection 7.2.1.2.

JRM/953171

LRG POSITION

ISSUE

ICSB-9 Safet — Related Dis la — (7.5)

The design of the safe shutdown indication does not satisfy the require-ments of IEEE Std 279-1971, Paragraph 4.10.

POSITION

1. Indicators and recorders as currently defined in section 7.5of the individual FSAR dockets to be necessary for "post-accident monitoring" will be qualified to operate following aseismic event.

2. Those indicators and recorders for "post-accident monitoring"functions defined above will have redundant channels with atleast one indicator for one channel and one recorder for theredundant channel.

3. "Safe shutdown instrumentation" shall be required to consistof quality components which display diverse parameters indica-tive of safe shutdown. The level of qualification to beimposed on these monitoring systems will be addressed in theapplicant's response to the Regulatory Guide 1.97 rev. 2requirements which is not imposed as requirement to obtain anoperating license in NUREG-0694.

NOTES:

The applicant's position with respect to both post-accidentmonitoring instrumentation (as defined in section 7.5) andsafe shutdown indication are expected to be upgraded in responseto Regulatory Guide 1.97 rev. 2. However, the regulatoryguide requirements do not impose these changes as prerequisitesto an operating license. A phased approach to the upgradingof equipment within the monitoring systems (as is currentlybeing done for NUREG-0588 requirements for environmentalqualification), will be undertak'en by the members of the LRG.



JRM/954566

LRG POSITION

ISSUE

ICSB-10 Rod Block Monitor — (7.6.3)

The applicant does not agree that the rod block monitor is a protectionsystem.

POSITION (Common)

The NRC has conducted an extensive review of the RMCS including re-fueling interlocks, RBM, RVM, RSCS on various dockets. Plants withopen items having similar designs will be conformed to the Zimmerdesign (i;e., the resolution will be reviewed and resolution basesif applicable will be incorporated).

The Zimmer design review has been completed and the issue resolved(Ref. 2). This closure basis will be relied upon.


The Susquehanna SES RBM design is identical to the Zimmer design. Seeresponse to Question 032.62.

JRM/187999

LRG POSITION

ISSUE

ICSB-ll MSIV Leaka e Control S stem —. (7.6.3)

Me identified a single failure to the MSIV leakage control system whichco ~ld lead to possible failure of the system during testing or operation.

POSITION (Common)

The design of the MSIV - LCS will be modified to eliminate thesingle failure concern. This modification will be equivalent asthat accepted on the Hatch and Zimmer dockets.


The design of the MSIV-LCS has been modified to eliminate the singlefailure concern.

JRM/054519

LRG 7'OSITION

ISSUE

PSB-1 Low or De raded Grid'Volta e (8.2.2)

Electrical system does not meet our requirements for protection underlow or degraded grid voltage conditions.

POSITION (Common)

Either: l) Applicant will commit to implement a second level of undervoltageprotection consistent with the guidance provided by the NRC Staff beforethe start of the second fuel cycle (LSCS, SNPS, EF-2, WNP-2); or

2) Applicant will demonstrate the adequacy of the grid without thesecond level of voltage protection to the satisfaction of the NRC staff(ZPS, SSES).


The Susquehanna SES position is presented in the response to Question~ ~40.6. Detailed design features are still under development.

JRM/042878

LRG POSITION

ISSUE

PSB-2. Test Results for Diesel Generators (8..3.2)

Test results for the diesel generators to indicate margin have not beensubmitted.

POSITION (Unique)

This issue applies to GE prototype tests performed to qualify thesubject diesels- as the emergency power supply for the HPCS system.Test reports for all diesels, developed during the initial testprogram at the site, are available for review by the regional IEoffice at the plant site.

NOTE:

1) GE HPCS diesel test report applicable to most BWR-5's and allBWR-6's, NEDO-10905-3 was submitted Dec. 20, 1979 to 0. D.Parr by J. F. Quirk letter.

e 2)

3)

NRC acceptance provided by O. D. Parr letter to G. G. Sherwoodon April 7, 1980.

r

This NEDO report is applicable for all LSCS diesels and theWNP-2 HPCS diesel.


Susquehanna SES does not have an HPCS system hence does not have thesubject diesels.

JRM/045744

LRG POSITION

ISSUE

PSB-3 Containment Electrical Penetrations (8.4.1)

The reactor containment electrical penetrations do not conform to RegulatoryGuide 1.63 and test results do not demonstrate that the electricalpenetrations can maintain their integrity for maximum fault current.

POSITION (Common)

The penetration design will conform to position Cl of RegulatoryGuide 1.63 (Oct. 1973) and with respect to back-up overcurrentprotection; either:

1) "Incorporating adequate self.-fusing characteristicswithin the penetration conductors themselves constitutean acceptable design approach"; or

2) "Where self-fusing characteristics are not incorporatedthe current overload protection system will conform tothe single failure criteria of IEEE-279 (71) Section 4.2;ANSI-N42.7(72).

NOTE:

1) Position 2 above applies to power circuits only. Control andinstrument circuits are not subject to detrimental high levelfault currents.

2) Regulatory Guide 1.63, Rev. 1 (May 1977) was identified forimplementation on CP applications docketed after Dec. 30,1977. In addition, as listed in NUREG-0427 Table III-13 andIII-14; RG 1.63 is identified as a Category I or a Category IIitem. As such applicants shall be allowed to demonstrate theadequacy of Rev. 0 of the Regulatory Guide.

3) The positions discussed above are not applicable to Fermi-2.The issue is considered closed by NUREG-0314.


Susquehanna SES position is discussed in Section 3.13.

JRM/925399

:LRG POSITION

ISSUE

PSB-4 Ade uac of the 120 Vac RPS Power Su 1 (8.4.7)

The applicant committed to the generic resolution or to expedite theirlicense will commit to the surveillance requirement which were appliedto Hatch 2.

POSITION

The applicants are all committed to implement prior to fuel loadingthe RPS-MG set design modification developed by General Electricfor generic application.

NOTE:

RPS power supply circuit design acceptance provided by letterof Feb. 23, 1979 from R. S. Boyd to G. G. Sherwood.


Susquehanna SES has committed to the GE generic modification in responseto Questions 032.25 and 032.66. See also Subsection 7.2.1.1.2 andFigure 7.2-1.

JRM/026655

LRG POSITION

'ISSUE

PSB-5 Thermal Overload Protection B ass (8.4.9)

We require the applicantwhich was used to selectdevices for valve motorsdevices will be tested.

to provide the detailed analysis and/or criteriathe setpoints for the thermal overload protectionin safety systems and the details as to how these

POSITION (Common)

Although documentation of conformance to R.G.1.106, Rev. 1 is notmandatory based on the classification of this guide as a Cat. I RegulatoryGuide; the LRG will, in order to facilitate the licensing review, implementeither position Cl or C2 of R.G.1.106.

NOTE'he

requirement for main control room indication of bypassesalluded to by the reference to 54.13 of IEEE-279 is judged tobe inapplicable because no "protective action" is involved.This position was judged acceptable on the Zimmer docket(SER 7.1.3).

I

Susquehanna SES meets the requirements of RG 1.106 as documented in Subsection8.1.6.1 and Table 8.1-1. The test program is stated in Technical Specification3/4.8.3.3.

ISSUE

PSB-6 Reliabilit of Diesel Generator

Plant Unique.'


Contention not subject to LRG verificating process.

JRM/920530

LRG POSITION

ISSUE

PSB-7 Shared DG Conformance to Re ulator Guide 1.81

Shared diesel design must meet position 2 of Regulatory Guide 1.81.'I

POSITION

For the two plants involved, it is judged that position 2 of theRegulatory Guide 1.81 is met.


Compliance is stated in Section 3.13.1 on Regulatory Guide 1.81. Seethe response to Question 40.25.

JRM/920529

X,B,G POSnloN

ISSUE

PSB-8 Periodic Diesel Generator Testin

Diesel Generator testing once every 18 months as required by RegulatoryGuide 1.108.

POSITION

Based on the errata to Regulatory Guide 1.108 of September 1977,the 18 month surveillance interval required in PSB-8 is no longerrequired, therefore this open issue should be closed.


The Susquehanna SES position is stated in Section 3.13.1 under RegulatoryGuide 1.108.

JRM/920528

LRG POSITION

ISSUE

CSB-1 Steam B ass of the Su ression Pool (6.2.1.1)

The applicant approach to suppression pool bypass is not consistent withBranch Technical Position CSB 6-5. The applicant must commit to performa low power surveillance leakage test of the containment at each refuelingoutage.

POSITION (Common)

The applicants commit to perform a single high pressure test priorto fuel loading and low pressure bypass tests at regular intervalsafter fuel load in conformance with the guidelines contained in theresponses referenced below.

NOTE:

1) LaSalle commits to perform the low pressure test at eachrefueling outage during which the integrated leak rate test isperformed. This is justified by the LaSalle design.

2) Susquehanna will not perform the high pressure test specifiedin the BTP. A low pressure test will be performed at eachrefueling outage during which the integrated leak rate test isperformed. This is justified by an ASME Code design and byfull non-destructive testing of the liner and penetrationwelds.


Because of the unique design of Susquehanna SES (described in responseto Question 021.66) the only potential path for bypass leakage is throughthe containment vacuum breakers. This path will be leak rate tested atleast once per 18 months. The unique design and frequency of leaktesting of the vacuum breakers, justifies the Susquehanna SES programdescribed in Subsection 6.2.6.5.1.

JRM/934757

-LRG POSITION

ISSUE

CSB-2 Pool D namic LOCA and SRV Loads (6.2.1.1)

The staff has completed its review of the short-term program and developedacceptance criteria. We require that the applicant commit to our acceptancecriteria or justify any exceptions taken.

POSITION (Common)

The LRG will not develop detailed positions to address this issue,but will rely on the ongoing generic review directed by the Mark IIOwners Group and monitored by the Division of Safety Technology.

NOTE:

1) The lead plant applicants have committed to meet the acceptancecriteria delineated in the NRC Lead Plant Acceptance Criteriaas documented in their respective DAR's. With the resolutionof questions concerning the load definition for CondensationOscillation and Chugging in August 1980, it is expected thatthe NRC Staff will issue a Lead Plant SER in the fourth quarterof 1980.

2) All exceptions to the Lead Plant Acceptance Criteria have beendocumented in plant unique correspondence.


See the Susquehanna SES Design Assessment Report.

JRM/934756

f

I

-'A*

*' LRG POSITION

ISSUEt

CSB-3', Containment Pur e S stem (6.2.4)I

A 2-inch vent line exists in the purge system to bleed-off excess primarycontainment pressure during operation. 'We require the applicant toevaluate this 2-inch bypass purge system in light of the criteria ofBranch Technical Position'SB 6-4.

POSITION (Unique)

I,a Salle has modified the design by adding a second valve in the2-inch line thus eliminating the single failure potential.


The design of Susquehanna SES meets the criteria of BTPCSB6-4 as statedin response to Question 21.86.

IRG POSITION

ISSUE

CSB-4 'ombustible Gas Control (6.2.5)

Although the proposed combustible gas control system is designed in,accordance with requirements of 10 CFR Part 50.44, we will require theapplicant to commit the following because of certain system characteristics:

a) If the containment pressure is above 15.3 psig and the hydrogenconcentration is 3.3 volume percent, the containment spraysystem must be actuated to reduce the'containment pressure.

b) Following a IOCA, the recombiner system becomes an extensionof the containment boundary. We require the applicant todemonstrate the leak tight integrity of the recombiner system.

POSITION

Those plants for which the recombiner system design pressureis less than the predicted containment design pressure; 'theapplicants commit to actuate the containment spray system asrequest'ed on the individual docket.

2) The applicant's agree to perform sy'tem leak tests. Specifictest's depend upon the details of the system design and areaddressed in the responses referenced below.


The hydrogen recombiners at Susquehanna SES are internal to the primarycontainment as described in Subsection 6.2.5.2. As shown on Table6.2-18, the recombiners are designed for 77 psi which exceeds the peakcalculated containment pressure. Thus this concern is not applicable toSusquehanna SES.

NLF2K

I+

~ „

t

LRG POSITION

ISSUE

CSB-5 .Containment I,eaka e Testin (6.2.6)

Additional information is required relating to containment leakagetesting to show compliance with Appendix J.

POSITION (Unique)


The Susquehanna SES position is described in Subsection 6.2.6 and Table6.2-22. The response to Question 21.87 provided further information onAppendix J testing.

2M

LRG POSITION

ISSUE

MEB-1 As etrical LOCA 8 SSE 8, Annulus Pressurization Loadson Reactor Vessel Internals and Su orts 3.9.2

Document your reevaluation of the safety-related systems and componentsbased upon the load combinations, 'response combination methodology, andacceptance criteria required by us as presented at our meeting of December12, 1978. (Reference letter dated September 18, 1978).

POSITION

The applicants will provide the information requested by NRC forthe vessel and internals/piping and equipment. Specifically:

1) For the load combination - Nt(OBE+(SRV ))AIL

The results of the analysis to UPSET (B) limits will be submitted.

2) For the load combination - N+(LOCA + SSE)7

Results of the analysis to FAULTED (D) 1'imits will be submittedfor piping not subject to functional capability assessment.The results of the analysis" to EMERGENCY (C) .limits will besubmitted for piping subject to functional capability assessment.


The response is contained in the Design Assessment Report. See Section3.9 of the FSAR and the responses to Questions 110.42 and 110.49.

NLF2N

.LRG POSITION

ISSUE

MEB-2 Prep erational Vibration Assurance Pro ram (3.9.2, 3.9.5)

Addition'al information is required concerning the basis for the allowablevibration amplitude derived and clarification of the use of twice thisallowable is acceptable.

POSITION (Unique)

This item is closed.


This topic is discussed in the response to Question 1l0.40 and Subsection3.9.2.lb.

LRG POSITION

ISSUE

MEB-3 namic Res onse Combination usin the SRSS'Techni ue3.9.3., 3.9.5

Me are studying the problem of utilizing the square root'f the sum ofthe squares for determining dynamic responses other than LOCA and SSE asyou have used. By not utilizing the absolute sum method, the review maybe entended if we do not agree that the square root of the sum of thesquares methodology is applicable.

POSITION


SUS(}UEHANNA POSITION

The combinational methods used on Susquehanna SES were discussed andapproved by the NRC at the draft MEB SER meeting in July, 1980.

NIZ2P

LRG POSITION

ISSUE

MEB-4 " Loadin Combinations Desi n Transients and Stress Limits

Clarify your consideration of the cyclic'oadings due to the operatingbasis earthquake and safety/relief valve actuation in your NSSS fatigueanalyses.

POSITION '

A BVR NSSS model subjected to 3 different recorded time hi'storiesand modal responses truncated to study the response of three differentfrequency bandwidths (0-10 Hz, 10-20 Hz and 20-50 Hz) was analyzed.This study showed that during a 40 year life, the probability ofone OBE with 50$ of the SSE intensity is extremely remote. Ittakes 20 quarter-SSE's, which are more realistic to produce thesame level of stress of one OBE. Therefore, to cover the combinedeffects of these earthquakes and the cummulative effects of evenlesser earthquakes, one OBE intensity earthquake is postulated forfatigue evaluation. In addition, the number of stress cyclesbetween one-half peak stress and full peak stress is less than 4g.Therefore, the assumption of 10 peak stress cycles provides an

'ddedmargin of conservatism.


The cyclic loadings used on Susquehanna SES were discussed and approvedby the NRC at the Draft MEB SER meeting in July 1980.

ISSUE

MEB-5

POSITION

LRG POSITION

Stress Corrosion Crackin of Stainless Steel Co onents-Desi n Modification 3.9.3

This item is closed


This item has been deleted from the LRG active list.

LRG POSITION

ISSUE

MEB-6 Pum 8 Valve 0 erabilit Assurance .Pro ram (3.9.3)

Additional information has been 'requested regarding your analytical andtesting methods for your pump and valve operability assurance program.

POSITION

,This item 'is closed.


See Subsection 3.9.3.2 a and b.

NIZ2R

LRG POSITION

ISSUE

MEB-7 Bolted Connections for Su orts (3.9.3)

. You have not provided the allowable limits for buckling for the reactorvessel support skirt'subjected to faulted conditions. In addition, we

requested information concerning the design of support bolts and boltedconnections.

POSITION



See response to question 110.43, which was discussed and approved at theDraft MEB SER meeting in July 1980.

LRG POSITION

ISSUE

MES-8 Pum 8 Valve Inservice Lists er 10CFR50.55a ( )

You have not submitted your proposed program for the inservice testingof pumps and valves as required by 10 CFR 55.55a (g);

POSITION (Unique)


The pump and valve inservice testing program will be submitted the firstquarter 1981.

AAW/3G:4

LRG POSITION

ISSUE

MEB-9 Review of in situ Test Pro ram of the Safet /Relief Valve

POSITION (Unique)

NOTE:- Both IaSalle and Zimmer will perform such tests. The detailsof the tests have been submitted for NRC review. At this time

~ no further testing is judged to be necessary by the other MK-IIplants.


No special test program is planned for Susquehanna SES.

AAW/3G:5

'LRG POSITION

ISSUE

MEB-ll Control Rod Drive Return Line

We have not completed our review of GE Topical Report NEDE-21821-2Aaddressing reactor feedwater nozzle/sparger design modification forcracks nor have we completed GE's generic modification to the controlrod drive return nozzle. This may required additional request forinformation.

POSITION

Refer to RSB-2


Refer to verification given in RSB-2.

AAW/3G:6

LRG POSITION

ISSUE

NEB-12 Confirmator Pi in Anal sis

Document your test program for all non-class 1, 2 and 3 high energypiping systems outside containment and all seismic Category I portionsof moderate energy piping systems outside containment.

POSITION



The Susquehanna SES testing program is contained in Sections 3.9 and14.2.

CTC:cvc5973

'LRG POSITION

ISSUE

MTEB-1 Preservice and Inservice Ins ection of Class 1 2 and 3Com onents er 10 CFR 50.55a

Preservice and inservice inspection of Class 1, 2 and 3 components havenot been submitted.

POSITION (Unique)


The inservice inspection program for Class 1, 2 and 3 components wassubmitted in January, 1981.

AAW/3G:8

= LRG POSITION

ISSUE

MTEB-2 Exem tions from A endix G to 10 CFR 50 (5.1.4.)

La Salle Station reactor vessels do not meet the specific requirementsof Appendix G of 10 CFR Part 50. Identify and justify your exemptions.(Reference Letter dated January 27, 1977).

POSITION (Unique)

Exemptions for Appendices G and H are identified and justified in Chapter5 of the CESAR's. Where requested, further justification was stated inresponse to the referenced questions.


The Susquehanna SES position on exemptions for Appendices G and H arestated in the responses to Questions 121.1 and 121.2 and in Tables5.3-la, 5.3-lb, 5.3-2a and 5.3-2b.

AAW/36:9

LRG POSITION

ISSUE

KZEB-3 Exem tions from A endix H to 10 CFR 50 (5.3.2. 5.3.3.)

Ia Salle Station surveillance program does not comply with Appendix H of10 CFR Part 50. Identify and justify your exemptions. (Referenceletter dated January 29, 1979)

POSITION (Unique)

Exemptions for Appendices G Sr H are'dentified and justified in Chapter5 of the FSAR's. Where requested, further justification was'stated inresponse to the referenced questions.


For the Susquehanna SES position, see MTEB-2.

AAW/3G:10

,LRG POSITION

, ISSUE

MTEB-4 Reactor Testin and Cooldown Limits4"

Insufficient information has been submitted for us to assess that themethods used to provide stress intensity values are equivalent to thoseobtained from Appendix G of ASME Code. Clarification and justificationof the methods used to construct the operating pressure temperaturelimits should be provided. (Reference 121.1)

POSITION

The reactors will be operated 'in a manner that will minimize the possibilityof rapidly propagating failure. The pressure-temperature limit curves,for all phases of plants operation, were established using the availableimpact test data and'conservative nil-ductility transition referencetemperature estimates to perform a fracture toughness calculation by themethods of the American Society of Mechanical Engineers Code, SectionIII, Appendix G (Summer 1972 Addenda) for all areas of the vessel remotefrom discontinuities. These calculations were based on a postulatedsurface flaw equal to one quarter of the material thickness. All vesselshell and head areas remote from discontinuities were'onsidered and theoperating curves we'e developed based on the limiting, area. The maximumthrough-wall temperature difference r'esulting in continuous heating orcooling at 100~ Fahrenheit per hour was considered. The safety factorsapplied were in accordance with American Society of Mechanical EngineersCode Section III, Appendix G, 10 CFR, Part 50,. Appendix G, paragraph IU.A.2.c and General Electric Company Topical Report, NED0-21778A, TransientPressure Rises Affecting Fracture Toughness Requirements for BWR's."

NOTE: Ia Salle has provided sufficient technical justification anddata in Chapter 5 to demonstrate that the estimate of initialNil-Ductilitytransition reference temperature is acceptable.Where requested, further justification was stated in responseto the reference questions.


The Susquehanna SES position is stated in Subsection 5.3.2.

AAW/3G:ll

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