ML18026A323
| ML18026A323 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 02/23/1981 |
| From: | Curtis N PENNSYLVANIA POWER & LIGHT CO. |
| To: | Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| ER-100450, PLA-642, NUDOCS 8102270493 | |
| Download: ML18026A323 (80) | |
Text
REGULATO, INFORMATION DISTRIBUTION TEM (RIDS)
ACCESSION NBR:8102270093 DOCsDATE! 81/02/23 NOTARIZED: NO, 'OCKET FACIL:50 387 Susquehanna Steam Electric Stationi Unit ig Pennsylva 05000387 50-388 Susquehanna Steam Electric Stationi Unit 2i Pennsylva 05000388 AUTH'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 ref facility~ 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, 05000387 05000388 RECIPIENT ID CODE/NAME.
ACTION:
A/D LICENSNG RUSHBROOKiM ~
INTERNAL: ACCID EVAL BR26 CHEM ENG BR 08 CORE PERF BR 10 EMERG PREP 22 GEOSC IENCES 10 HYD/GEO BR 15 ILE 06 LIC QUAL BR MECH ENG BR 18 NRC PDR 02 OP LIC BR PROC/TST REV 20 RAD SESS BR22.
F 01 BR25 COPIES LTTR ENCL 0
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1 1
1 1
1, 1
1 0
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1 1
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1 1
RECIPIENT ID CODE/NAME YOUNGBLOODRB STARKgR ~
00 AUX SYS BR 07 CONT SYS BR 09 EFF TR SYS BR1?
EQUIP QUAL BR13 HUM FACT ENG BR ILC SYS BR 16 LIC GUID BR NATL ENG BR 17 MPA OELD POWER SYS BR 19 QA BR 21 REAC SYS BR 23 SIT ANAL BR 24 SYS INTERAC BR COPIES LTTR ENCL' 0
1 1
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3 3-1 1
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1 0
1 0
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EXTERNAL: ACRS NSIC 27 05 16 16 1
1 LPDR 03 1
1 SAR 0g )gg)
TOTAL NUMBER OF COPIES REQUIRED:
LTTR 57 ENCL'1
0 l
W L
O'Pe IL TWO NORTH NINTH STREET, ALLENTOWN, PA. 18101 NORMAN W. CURTIS Vice President-Engtneering 8, Construction-Nuclear 770 5381 (J
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February 23, 1981 Mr. B. J. Youngblood, Chief Licensing Projects Branch 8'1 Division of Project Management U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Docket No. 50-387 50-388 SUSQUEHANNA STEAM ELECTRIC STATION LRG POSITIONS ER 100450 FILE 841-2 PLA-642
Dear Mr. Youngblood:
Attached are Pennsylvania Power
& Light Company's responses to the Licensing Review Group's positions.
We would like to discuss the status of these positions with your personnel during the next LRG meeting.
This meeting is scheduled for March 18, 1981.
If you have any questions, please call me.
Very truly yours, N.
W. Curtis Vice President-Engineering
& Construction-Nuclear CTC/mks Attachment cc:
R.
M. Stark NRC goo I i/i PENNSYLVANIA POWER 8
LIGHT COMPANY
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'RG POSITION 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 below which, if failure should occur in a high pressure
- system, damage to other components within the primary containment would not be significant?
State the criteria used to determine this size.
Identify all valves in the primary containment larger than this size and identify the missile protection 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 pump overspeed during a LOCA.
SUS UEHANNA POSITION Position 2 above has been implemented as described in FSAR Subsections 3.5.1.1 and 3.5.1.2.
A mathematical analysis and calculation results were supplied in response to Question 211.195.
Other relevant questions include:
211.2, 211.36, 211.37, 211.38, 211.39, 211.194, 211.196 and 211. 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, we are reviewing generically with regard to the impact on control rod drive system performance.
Consequently, we require the applicant submit system performance data directly applicable to LaSalle and will require the applicant to conform to the conclusion of the generic study as applicable to LaSalle.
POSITION (Common) 1.
Remove the CRD return line and cap the vessel nozzle.
Accept-ability is demonstrated by GE analysis of CRD performance characteristics.
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, assuming acceptable return path established.
IGSCC is corrected by the reroute.
NOTE: 1.
The LRG members acknowledge the receipt of the NRC letter "modifications to Boiling Water Reactor Control Rod Drive Systems" in March, 1980 and will factor the concerns of the letter 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.
SUS UEHANNA POSITION The acceptability of position 1 above for Susquehanna SES is discussed in the response to Questions 211.43 and 211.192.
System performance data 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 and operating experience with the applicant's safety/relief valves.
POSITION (Common) 1.
Provide evaluation and operating history.
See Owners Group response to position 2.1.2 of NUREG-0578.
2.
Participate in TMI Qualification Program (II.D.l)
SUS UEHANNA POSITION o
A detailed evaluation of the qualification tests is provided in response to Question 211.70.
o Details for the surveillance and -testing of safety relief valves are included in the technical specifications and will be included in 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 to consider the effect of the ATWS RPT.
POSITION-The over pressure protection report will be resubmitted to demonstrate compliance with the ASME Pressure Vessel Code, considering the addition of ATWS RPT.
This position was acceptable on the Zimmer docket as noted by official NRC meeting minutes for a March 28, 1979 meeting.
This item is resolved for Zimmer.
SUS UEHANNA POSITION 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 leakage from the reactor coolant systems into both the low pressure coolant injection (3 trains) and low pressure core spray systems as required by Regulatory Guide 1.45.
POSITION (Unique)
'he LRG members will demonstrate conformance to Regulatory Guide 1.45(4) with reliance on level instrumentation, pressure instruments, and/or radiation monitors..
SUS UEHANNA POSITION 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 the RCIC pump suction has to be automatically switched from the condensate "storage tank to the suppression pool in the event of.a safe shutdown earthquake and concomitant failu're of the condensate storage tank.
POSITION (Common)
The qualification of RCIC will be resolved by insuring the availability of 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 suppression pool.
NOTE:
a)
This position satisfies the requirement identified in NUREG-0660 item II.K.3(C.3.22).
b)
It is not an OL item in accordance wth NUREG 0694.
Compliance with C.3.22, verify procedures, will be accomplished prior to fuel load on Jan.
1, 1981, whichever is later.
Compliance with design modification will be accomplished by Jan.
1, 1982.
c)
Also resolves issue ICSB-7.
SUS UEHANNA POSITION Susquehanna SES will be provided with an automatic switchover from the condensate storage tank to the suppression poolby January 1,
1982.
AAW/3C
LRG POSITION ISSUE RSB-7 Shutdown Unintentionall of the Reactor Core Isolation 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 of spurious ambient temperature signals from areas in and around the system (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, and then introducing the heat release cause by an alarm limit leak.
Actual ambient temperature settings will be determined during startup testing.
NOTES:
1.
CECO's notes of LaSalle/NRC Meeting of October ll, 1979 item 3.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 issue discussed above.
Although this NUREG-0660 is not a require-ment for licensing of the LRG plants (Ref. NUREG-0694),
the BVR (TMI) Owners Group is addressing that task and the utili-ties represented by the LRG are participating in that effort.
SUS UEHANNA POSITION In response to Question 211.33, Subsection 5.2.5.1.3 was revised to show the 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 required consistent with the analyses reported in Section 15.2.9 of the FSAR.
POSITION (Common) 1.
Crosby Valves, e.g. LaSalle Demonstrate by analysis (based on valve characteristics such as internal dimensions) that the valve(s) will pass sufficient water to allow a forced cooldown to be maintained.
2.
Target Rock (2 stage),
e.g.
SHOREHAM Demonstrate by shop test of a valve that (1) the valve can be opened by low pressure water and (2) once open is maintained open and passes sufficient flow to maintain a forced cooldown.
3.
Provide a backup to RHR shutdown cooling which does not utilize the SRV in the alternate cooldown mode.
NOTE:
NUREG-0660 Task II.D.l is related to the issue discussed above.
This task is identified as a requirement for licensing of LRG plants (Ref: NUREG-0699).
The BWR (TMI) Owners Group is addressing that task and the utilities represented by the LRG are participating in this effort.
SUS UEHANNA POSITION This item was addressed in the response to Question 211.8.
JRM/321405
LRG POSITION ISSUE RSB-9
~
Cate orization of Valve which Isolate RHR from Reactor Coolant S stem (5.4.2)
We require that the valves which serve to isolate the residual heat removal system from the reactor coolant system be classified category A/C in accordance with the provisions of Section XI of the ASME code.
RSB-13 Leaka e 8 Testin of Valves Used to Isolate Reactor Coolant S stem (6.3.2)
We require periodic testing arid, establishment of leak rate criteria for the valves that isolate the reactor coolant system from all the emergency core cooling system.
POSITION (Common)
Containment isolation valves which also provide isolation between high and low pressure systems will be tested in accordance with ASME Section XI as well as Appendix J to 10CFR50.
Exemption from ASME 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 SUS UEHANNA POSITION These items will be discussed in pump and valve ISI submittal.
See also the 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 pool leading to the ECCS pumps are designed to preclude adverse vortex formation and air injection which could effect the pumps performance.
POSITION (Unique)
The issue is regarded as plant unique and requires plant specific review of system design details.
SUSQUEHANNA POSITION Adequacy of the design to prevent adverse vortex formation 'is verified during 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 top of ECCS pipe lines.
The applicant must provide adequate instrumentation to 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 required on Fermi-2 and SSES dockets.
SUSQUEHANNA POSITION A reliable discharge line fillsystem is provided as described in Subsection 6.3.2.2.5.
Thzs subsection.has been amended in response to questions 211.58, 211.102, 211.212 and 211.218.
Adequacy of the system is shown by alarms (as described in Subsection 6.3.2.2.5) and by surveillance testing (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 sufficient for the extended operating time required and assures us by reliability data that the ADS valves will function as required.
POSITION (Common)
The applicant will review and document that the ADS system design satisfies the expressed concern.
NOTES:
a)
The ADS system for Zimmer which is a typical design has been accepted by the NRC.
b)
CECO's La Salle/NRC meeting notes of October ll, 1979 item 3L indicate issue as closed.
La Salle's ADS system is identical to Zimmer's.
c)
NUREG-0660 Task II.K.3 item C.3.28 is related to the issue discussed above.
This task is not identified as a
requirement for licensing of LRG plants (Ref:
The BWR(TMI) owners group is addressing that task and the utilities represented by the LRG are participating in that effort.
SUSQUEHANNA POSITION A non-interruptible safety grade source of gas is supplied to the ADS valves as described in the response to Question 211.67 and in Subsection 9.3.1.5.
NLF 2I
- i LRG POSITIONS ISSUE RSB-14 0 erabilit of ECCS Pum s - (6.3.2)
The applicant must provide assurance that the ECCS pumps can function for an extended time (maintenance free) under the most limiting post-LOCA conditions.
POSITION (Common)
This issue has been closed on Zimmer and Shoreham dockets on the basis of information presented in response to NRC questions.
Similar information has been provided on the rest of the dockets.
NUREG-0660 Task II.B.2 is related to the issue discussed above and is being addressed by the BVR TMI Owners Group.
SUS UEHANNA POSITION Operating history of ECCS pumps was provided in response to question 211.106.
This is the same data used to address the question on the LaSalle 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 to complete 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 square feet.
POSITION (Common)
Lead plant bounding.approach valid and confirmed by all additional analyses performed (Ref LaSalle Q212.125).
.No additional plant specific analysis will be provided other than that already provided by each plant for 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 flow control valve during a LOCA which resulted in a 300 F increase in peak 0
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 described represents a non-mechanistic failure mode.
In reality, for all failure modes except valve hydraulic system failure, the valve fails as is, which is physically limited to at least 20% open.
If the hydraulic system fails, the valve will dri.ft closed, but at a design 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 operator action to restore core cooling prior to excessive core, heating as a
result 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 heat removal line that was postulated to occur during normal shutdown cooling, operator action is indicated to restore core cooling.
We require the applicant 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.
Subsequently vessel water Level 2 will be reached and automatic initiation of HPCS/HPCI will occur.
HPCS/HPCI will cycle on and off between Levels 3 and 8 until the operator establishes an alternate water source.
If HPCS/HPCI is unavailable, representative analyses have been performed to demonstrate that operator action would not be required before 20 to 30 minutes following the pipe crack (depending upon plant design) to assure adequate core cooling in accordance with the acceptance criteria of 10CFR50.46.
SUS UEHANNA POSITION The Susquehanna SES position is stated in response to Question 211.50.
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 and drywell spray and suppression pool cooling.
The applicant must demonstrate that adequate core cooling is maintained when diversion is considered.
POSITION (Common)
The limiting break will be reviewed with LPCI diversion after 600 sec to confirm that the peak clad temperature limit is not exceeded and adequate core cooling is maintained.
Such analyses have been performed for Zimmer and LaSalle (BWR-5) and Shoreham (BWR-4).
These'analyses confirm the acceptability of LPCI diversion at 600 sec.
The applicability of these results to the other LRG plants will be verified.
SUS UEHANNA POSITION The Susquehanna SES position is stated in the responses to Questions 211.105 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 controller indicates that the temperature drop is no greater than 100 F.
At a 0
domestic boiling water reactor an actual feedwater temperature occurred 0
which demonstrated a temperature difference of 150 F.
The applicant must justify the decrease in temperature drop used'for this event or recalculate the transient by using a )ustified temperature decrease to assure 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 LFWH event.
LaSalle (BWR-5) has completed this calculation (see Amendment
- 49) confirming the conclusion that the results are insensitive to AT assumed and is not a limiting event.
The applicability of these results to the other LRG plants will be verified.
SUS UEHANNA POSITION The Susquehanna SES feedwater system is designed so that no single failure can cause the loss of more than one feedwater heater string.
Loss of one heater string results in a temperature drop of less than 100 F.
See response to Question 211.116.
It has been shown in the response to Question 211.148 that based on analysis 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 ated 0 erational Transients (15.1)
In analyzing anticipated operational transients, the applicant took credit for equipment which has not been shown to be reliable.
Our position 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 requirement along with any system modification that may be required to fulfillthe requirement.
POSITION (Common)
This issue is related to a long time concern by the NRC Re:
the use of non-safety grade equipment to mitigate transients.
Many questions were asked on many dockets relative to:
a)
Credit for Non class lE relief function vs.
1E safety functions 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-safety grade equipment was the excess feedwater transient relying on the L-8 turbine trip and turbine bypass.
NRC concurred that providing technical specifications for the L-8 trip and the turbine bypass valves satisfactorily resolves this issue.
All LRG plants commit to this except Fermi-2.
SUS UEHANNA POSITION The Susquehanna SES position is stated in response to Question 211.114.
JRM/056796
LRG POSITION ISSUE RSB-21 Use of Non-Safet Grade E ui ment in Shaft Seizure Accident (15.2)
The applicant included the use of non-safety grade equipment in his analyses for shaft seizure and shaft break accidents.
We require that these accidents be reanalyzed without allowance for the use of non-safety grade equipment.
POSITION (Common)
The evaluation basis for the subject event was reviewed and accepted on the Zimmer docket (SER 15.2).
2.
A qualitative assessment of the event without reliance on non safety grade equipment suggests that the applicant's conclusion (i.e. the event is bounded by the DBA LOCA event) would not change.
3.
The non-safety grade equipment in question have been reviewed in item RSB-21 and steps have been taken to improve their reliability.
4.
No further analysis is required.
SUS UEHANNA POSITION The Susquehanna SES position is stated in the response to Question 211.120.
JRM/056597
LRG POSITION ISSUE RSB-22 ATWS (15.2. 1)
We require that the applicant agrees to implement plant modifications on a scheduled basis in conformance with the Commission's final resolution of ATWS.
In the event that LaSalle starts operation before necessary plant modifications are implemented, we require some interim actions be taken 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 an ATWS event, including consideration of scram indicators, rod position indicators, 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 including consideration of immediately manual scramming the reactor by using the manual scram buttons followed by changing rod scram switches to the scram position, stripping the feeder breakers on the reactor protection system power distribution buses, opening the scram discharge volume drain valve, prompt actuation of the standby liquid control system, and prompt placement of the RHR in the pool cooling 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 2
on a best effort schedule as the basis for final resolution of the ATWS issue.
SUS UEHANNA POSITION The Susquehanna SES position is stated in response to Question 32.87.
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 limiting pressure transients utilizing option B (GE/NRC generic resolution in process estimated resolution summer 1980) of the NRC letter dated 1/23/80.
LaSalle (lead plant) schedule for completion of calculation is October 1980.
NOTE:
REDY is still an acceptable code for other transients.
SUS UEHANNA POSITION The Susquehanna SES position is that the limiting pressure transients will be 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 power ratio must be recalculated taking into consideration the turbine trip without bypass event.
The transient of generator load rejection without bypass results in an MCPR equal to 1.02 which is below the Safety limit of 1.06.
The applicant classified 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 limiting pressure transients utilizing option B (GE/NRC generic resolution in process estimated resolution summer 1980) of the NRC letter dated 1/23/80.
LaSalle (lead plant) schedule for completion of calculation is Oct. 1980.
Note that we understand REDY is still an acceptable code for other transients.
SUS UEHANNA POSITION 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 initial core load, we are concerned that GEXL correlation may not be conservative for reload operation.
The applicant, in a letter dated March 7, 1979, committed to incorporate the latest approved form of GEXL correlation at the time of reload for LaSalle.
License Condition.
POSITION (Common) 1.
For first core cycle, GEXL is conservative.
Adequate negative worth is provided by the control system to assure shutdown capability.
2.
The applicant(s) commit to incorporate the latest approved form of GEXL correlation at the time of reload.
SUS UEHANNA POSITION The Susquehanna SES position is stated in response to Question 221.3.
Also Pennsylvania Power and Light Company will incorporate the appropriate correlation 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 generic review of hydrodynamic stability characteristics is -completed.
In addition, the LaSalle stability analysis was performed for first cycle.
" We will require that a new analysis be submitted and approved prior to second cycles operation.
The applicant, in a letter'dated March 7, 1979, committed to perform, as part of the future reload analyses to update the hydrodynamic stability analyses.
License condition.
POSITION (Common)
Sufficient documentation of stability margin for Cycle 1 has been presented for issuance of an operating license.
The stability margin analysis will be updated as required for future reload applications.
This issue is addressed under generic NRC task action plan (TAP) B-19.
SUS UEHANNA POSITION The stability margin analysis will be updated as required for reload applications.
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 to adequate separation criteria, have not been qualified, and do not adhere to separation of Class lE to non-class 1E instrumentation.
POSITION (Unique)
It is judged that this review should be made on a plant specific basis.
Regulatory Guide 1.75 is not applicable to any of the plants in the LRG.
The degree of conformance to this Regulatory Guide has been addressed on each docket and has in the case of the Zimmer docket been accepted (SER 8.1.2 and 8.3.3)
SUS UEHANNA POSITION Section D, "Implementation", of Regulatory Guide 1.75, Revision 2 clearly states that the guide is not applicable to Susquehanna SES vintage plants.
The electrical separation criteria used for the Susquehanna SES design is described and compared with the requirements of Regulatory Guide 1.75 in Sections 3.12 and 8.1.6.
See also the response to Question 32.49.
JRM/042654
LRG POSITION ISSUE ICSB-3 ATWS POSITION (Common)
See Issue RSB-22 SUS UEHANNA POSITION 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 for the applicant to pull fuses.
We consider that this design does not satisfy 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 be addressed by LEG.
SUS UEHANNA POSITION Compliance with these paragraphs of IEEE 279-1971 is discussed in response to 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 case combination of setpoint and accuracy.
POSITION
,(Unique)
This issue is considered to be plant specific and will not be addressed by LRG.
Technical justification for the setpoints and allowable values are expected to be made available during technical specification review.
SUS UEHANNA POSITION The Susquehanna SES technical specifications have been submitted and are under 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 control drawings 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; SUS UEHANNA POSITION 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.
SUS UEHANNA POSITION 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 7.5 of the individual FSAR dockets to be necessary for "post-accident monitoring" will be qualified to operate following a seismic event.
2.
Those indicators and recorders for "post-accident monitoring" functions defined above will have redundant channels with at least one indicator for one channel and one recorder for the redundant channel.
3.
"Safe shutdown instrumentation" shall be required to consist of quality components which display diverse parameters indica-tive of safe shutdown.
The level of qualification to be imposed on these monitoring systems will be addressed in the applicant's response to the Regulatory Guide 1.97 rev.
2 requirements which is not imposed as requirement to obtain an operating license in NUREG-0694.
NOTES:
The applicant's position with respect to both post-accident monitoring instrumentation (as defined in section 7.5) and safe shutdown indication are expected to be upgraded in response to Regulatory Guide 1.97 rev.
2.
However, the regulatory guide requirements do not impose these changes as prerequisites to an operating license.
A phased approach to the upgrading of equipment within the monitoring systems (as is currently being done for NUREG-0588 requirements for environmental qualification), will be undertak'en by the members of the LRG.
SUS UEHANNA POSITION The Susquehanna SES position is stated in response to Question 32.29.
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 protection system.
POSITION (Common)
The NRC has conducted an extensive review of the RMCS including re-fueling interlocks, RBM, RVM, RSCS on various dockets.
Plants with open items having similar designs will be conformed to the Zimmer design (i;e., the resolution will be reviewed and resolution bases if applicable will be incorporated).
The Zimmer design review has been completed and the issue resolved (Ref. 2).
This closure basis will be relied upon.
SUS UEHANNA POSITION The Susquehanna SES RBM design is identical to the Zimmer design.
See response 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 which co ~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 the single failure concern.
This modification will be equivalent as that accepted on the Hatch and Zimmer dockets.
SUS UEHANNA POSITION The design of the MSIV-LCS has been modified to eliminate the single failure 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 under low or degraded grid voltage conditions.
POSITION (Common)
Either:
l) Applicant will commit to implement a second level of undervoltage protection consistent with the guidance provided by the NRC Staff before the start of the second fuel cycle (LSCS,
- SNPS, EF-2, WNP-2); or 2)
Applicant will demonstrate the adequacy of the grid without the second level of voltage protection to the satisfaction of the NRC staff (ZPS, SSES).
SUS UEHANNA POSITION The Susquehanna SES position is presented in the response to Question
~
~
40.6.
Detailed design features are still under development.
JRM/042878
0 0
LRG POSITION ISSUE PSB-2.
Test Results for Diesel Generators (8..3.2)
Test results for the diesel generators to indicate margin have not been submitted.
POSITION (Unique)
This issue applies to GE prototype tests performed to qualify the subject diesels-as the emergency power supply for the HPCS system.
Test reports for all diesels, developed during the initial test program at the site, are available for review by the regional IE office at the plant site.
NOTE:
1)
GE HPCS diesel test report applicable to most BWR-5's and all BWR-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.
Sherwood on April 7, 1980.
r This NEDO report is applicable for all LSCS diesels and the WNP-2 HPCS diesel.
SUS UEHANNA POSITION Susquehanna SES does not have an HPCS system hence does not have the subject diesels.
JRM/045744
LRG POSITION ISSUE PSB-3 Containment Electrical Penetrations (8.4.1)
The reactor containment electrical penetrations do not conform to Regulatory Guide 1.63 and test results do not demonstrate that the electrical penetrations can maintain their integrity for maximum fault current.
POSITION (Common)
The penetration design will conform to position Cl of Regulatory Guide 1.63 (Oct.
1973) and with respect to back-up overcurrent protection; either:
1)
"Incorporating adequate self.-fusing characteristics within the penetration conductors themselves constitute an acceptable design approach";
or 2)
"Where self-fusing characteristics are not incorporated the current overload protection system will conform to the 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 and instrument circuits are not subject to detrimental high level fault currents.
2)
Regulatory Guide 1.63, Rev.
1 (May 1977) was identified for implementation on CP applications docketed after Dec.
30, 1977.
In addition, as listed in NUREG-0427 Table III-13 and III-14; RG 1.63 is identified as a Category I or a Category II item.
As such applicants shall be allowed to demonstrate the adequacy 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.
SUS UEHANNA POSITION 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 their license will commit to the surveillance requirement which were applied to Hatch 2.
POSITION The applicants are all committed to implement prior to fuel loading the RPS-MG set design modification developed by General Electric for generic application.
NOTE:
RPS power supply circuit design acceptance provided by letter of Feb.
23, 1979 from R.
S.
Boyd to G.
G. Sherwood.
SUS UEHANNA POSITION Susquehanna SES has committed to the GE generic modification in response to Questions 032.25 and 032.66.
See also Subsection 7.2.1.1.2 and Figure 7.2-1.
JRM/026655
LRG POSITION
'ISSUE PSB-5 Thermal Overload Protection B
ass (8.4.9)
We require the applicant which was used to select devices for valve motors devices will be tested.
to provide the detailed analysis and/or criteria the setpoints for the thermal overload protection in safety systems and the details as to how these POSITION (Common)
Although documentation of conformance to R.G.1.106, Rev.
1 is not mandatory based on the classification of this guide as a Cat. I Regulatory Guide; the LRG will, in order to facilitate the licensing review, implement either position Cl or C2 of R.G.1.106.
NOTE'he requirement for main control room indication of bypasses alluded to by the reference to 54.13 of IEEE-279 is judged to be 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 Subsection 8.1.6.1 and Table 8.1-1.
The test program is stated in Technical Specification 3/4.8.3.3.
ISSUE PSB-6 Reliabilit of Diesel Generator Plant Unique.
SUS UEHANNA POSITION 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 the Regulatory Guide 1.81 is met.
SUS UEHANNA POSITION Compliance is stated in Section 3.13.1 on Regulatory Guide 1.81.
See the 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 Regulatory Guide 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 longer
- required, therefore this open issue should be closed.
SUS UEHANNA POSITION The Susquehanna SES position is stated in Section 3.13.1 under Regulatory Guide 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 with Branch Technical Position CSB 6-5.
The applicant must commit to perform a low power surveillance leakage test of the containment at each refueling outage.
POSITION (Common)
The applicants commit to perform a single high pressure test prior to fuel loading and low pressure bypass tests at regular intervals after fuel load in conformance with the guidelines contained in the responses referenced below.
NOTE:
1)
LaSalle commits to perform the low pressure test at each refueling outage during which the integrated leak rate test is performed.
This is justified by the LaSalle design.
2)
Susquehanna will not perform the high pressure test specified in the BTP.
A low pressure test will be performed at each refueling outage during which the integrated leak rate test is performed.
This is justified by an ASME Code design and by full non-destructive testing of the liner and penetration welds.
SUS UEHANNA POSITION Because of the unique design of Susquehanna SES (described in response to Question 021.66) the only potential path for bypass leakage is through the containment vacuum breakers.
This path will be leak rate tested at least once per 18 months.
The unique design and frequency of leak testing of the vacuum breakers, justifies the Susquehanna SES program described 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 developed acceptance criteria.
We require that the applicant commit to our acceptance criteria 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 II Owners Group and monitored by the Division of Safety Technology.
NOTE:
1)
The lead plant applicants have committed to meet the acceptance criteria delineated in the NRC Lead Plant Acceptance Criteria as documented in their respective DAR's.
With the resolution of questions concerning the load definition for Condensation Oscillation and Chugging in August 1980, it is expected that the NRC Staff will issue a Lead Plant SER in the fourth quarter of 1980.
2)
All exceptions to the Lead Plant Acceptance Criteria have been documented in plant unique correspondence.
SUS UEHANNA POSITION See the Susquehanna SES Design Assessment Report.
JRM/934756
f I
-'A LRG POSITION ISSUE t
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 primary containment pressure during operation.
'We require the applicant to evaluate this 2-inch bypass purge system in light of the criteria of Branch Technical Position'SB 6-4.
POSITION (Unique)
I,a Salle has modified the design by adding a second valve in the 2-inch line thus eliminating the single failure potential.
SUS UEHANNA POSITION The design of Susquehanna SES meets the criteria of BTPCSB6-4 as stated in 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 the applicant to commit the following because of certain system characteristics:
a)
If the containment pressure is above 15.3 psig and the hydrogen concentration is 3.3 volume percent, the containment spray system must be actuated to reduce the'containment pressure.
b)
Following a IOCA, the recombiner system becomes an extension of the containment boundary.
We require the applicant to demonstrate the leak tight integrity of the recombiner system.
POSITION Those plants for which the recombiner system design pressure is less than the predicted containment design pressure;
'the applicants commit to actuate the containment spray system as request'ed on the individual docket.
2)
The applicant's agree to perform sy'tem leak tests.
Specific test's depend upon the details of the system design and are addressed in the responses referenced below.
SUS UEHANNA POSITION The hydrogen recombiners at Susquehanna SES are internal to the primary containment as described in Subsection 6.2.5.2.
As shown on Table 6.2-18, the recombiners are designed for 77 psi which exceeds the peak calculated containment pressure.
Thus this concern is not applicable to Susquehanna SES.
NLF 2K
I+
~
t LRG POSITION ISSUE CSB-5
.Containment I,eaka e Testin (6.2.6)
Additional information is required relating to containment leakage testing to show compliance with Appendix J.
POSITION (Unique)
SUS UEHANNA POSITION The Susquehanna SES position is described in Subsection 6.2.6 and Table 6.2-22.
The response to Question 21.87 provided further information on Appendix J testing.
2M
LRG POSITION ISSUE MEB-1 As etrical LOCA 8 SSE 8, Annulus Pressurization Loads on Reactor Vessel Internals and Su orts 3.9.2 Document your reevaluation of the safety-related systems and components based upon the load combinations, 'response combination methodology, and acceptance criteria required by us as presented at our meeting of December 12, 1978.
(Reference letter dated September 18, 1978).
POSITION The applicants will provide the information requested by NRC for the 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 submitted for piping not subject to functional capability assessment.
The results of the analysis" to EMERGENCY (C).limits will be submitted for piping subject to functional capability assessment.
SUSQUEHANNA POSITION The response is contained in the Design Assessment Report.
See Section 3.9 of the FSAR and the responses to Questions 110.42 and 110.49.
NLF 2N
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 allowable vibration amplitude derived and clarification of the use of twice this allowable is acceptable.
POSITION (Unique)
This item is closed.
SUS UEHANNA POSITION This topic is discussed in the response to Question 1l0.40 and Subsection 3.9.2.lb.
LRG POSITION ISSUE MEB-3 namic Res onse Combination usin the SRSS'Techni ue 3.9.3., 3.9.5 Me are studying the problem of utilizing the square root'f the sum of the squares for determining dynamic responses other than LOCA and SSE as you have used.
By not utilizing the absolute sum method, the review may be entended if we do not agree that the square root of the sum of the squares methodology is applicable.
POSITION This item is closed.
SUS(}UEHANNA POSITION The combinational methods used on Susquehanna SES were discussed and approved by the NRC at the draft MEB SER meeting in July, 1980.
NIZ 2P
LRG POSITION ISSUE MEB-4 Loadin Combinations Desi n Transients and Stress Limits Clarify your consideration of the cyclic'oadings due to the operating basis earthquake and safety/relief valve actuation in your NSSS fatigue analyses.
POSITION '
A BVR NSSS model subjected to 3 different recorded time hi'stories and modal responses truncated to study the response of three different frequency bandwidths (0-10 Hz, 10-20 Hz and 20-50 Hz) was analyzed.
This study showed that during a 40 year life, the probability of one OBE with 50$ of the SSE intensity is extremely remote. It takes 20 quarter-SSE's, which are more realistic to produce the same level of stress of one OBE.
Therefore, to cover the combined effects of these earthquakes and the cummulative effects of even lesser earthquakes, one OBE intensity earthquake is postulated for fatigue evaluation.
In addition, the number of stress cycles between one-half peak stress and full peak stress is less than 4g.
Therefore, the assumption of 10 peak stress cycles provides an
'dded margin of conservatism.
SUS UEHANNA POSITION The cyclic loadings used on Susquehanna SES were discussed and approved by 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 SUS(}UEHANNA POSITION 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 and testing methods for your pump and valve operability assurance program.
POSITION
,This item 'is closed.
SUSQUEHANNA POSITION See Subsection 3.9.3.2 a and b.
NIZ 2R
LRG POSITION ISSUE MEB-7 Bolted Connections for Su orts (3.9.3)
. You have not provided the allowable limits for buckling for the reactor vessel support skirt'subjected to faulted conditions.
In addition, we requested information concerning the design of support bolts and bolted connections.
POSITION This item is closed.
SUS UEHANNA POSITION See response to question 110.43, which was discussed and approved at the Draft 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 testing of pumps and valves as required by 10 CFR 55.55a (g);
POSITION (Unique)
SUS UEHANNA POSITION The pump and valve inservice testing program will be submitted the first quarter 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 details of the tests have been submitted for NRC review.
At this time
~
no further testing is judged to be necessary by the other MK-II plants.
SUS(}UEHANNA POSITION 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-21821-2A addressing reactor feedwater nozzle/sparger design modification for cracks nor have we completed GE's generic modification to the control rod drive return nozzle.
This may required additional request for information.
POSITION Refer to RSB-2 SUS UEHANNA POSITION 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 energy piping systems outside containment and all seismic Category I portions of moderate energy piping systems outside containment.
POSITION This item is closed.
SUS UEHANNA POSITION The Susquehanna SES testing program is contained in Sections 3.9 and 14.2.
CTC:cvc 5973
LRG POSITION ISSUE MTEB-1 Preservice and Inservice Ins ection of Class 1 2 and 3
Com onents er 10 CFR 50.55a Preservice and inservice inspection of Class 1,
2 and 3 components have not been submitted.
POSITION (Unique)
SUS UEHANNA POSITION The inservice inspection program for Class 1, 2 and 3 components was submitted 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 requirements of 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 Chapter 5 of the CESAR's.
Where requested, further justification was stated in response to the referenced questions.
SUSQUEHANNA POSITION The Susquehanna SES position on exemptions for Appendices G and H are stated in the responses to Questions 121.1 and 121.2 and in Tables 5.3-la, 5.3-lb, 5.3-2a and 5.3-2b.
AAW/36:9
1
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 of 10 CFR Part 50.
Identify and justify your exemptions.
(Reference letter dated January 29, 1979)
POSITION (Unique)
Exemptions for Appendices G
Sr H are'dentified and justified in Chapter 5 of the FSAR's.
Where requested, further justification was'stated in response to the referenced questions.
SUS UEHANNA POSITION For the Susquehanna SES position, see MTEB-2.
AAW/3G:10
,LRG POSITION
, ISSUE MTEB-4 Reactor Testin and Cooldown Limits 4"
Insufficient information has been submitted for us to assess that the methods used to provide stress intensity values are equivalent to those obtained from Appendix G of ASME Code.
Clarification and justification of the methods used to construct the operating pressure temperature limits should be provided.
(Reference 121.1)
POSITION The reactors will be operated 'in a manner that will minimize the possibility of rapidly propagating failure.
The pressure-temperature limit curves, for all phases of plants operation, were established using the available impact test data and'conservative nil-ductility transition reference temperature estimates to perform a fracture toughness calculation by the methods of the American Society of Mechanical Engineers Code,Section III, Appendix G (Summer 1972 Addenda) for all areas of the vessel remote from discontinuities.
These calculations were based on a postulated surface flaw equal to one quarter of the material thickness.
All vessel shell and head areas remote from discontinuities were'onsidered and the operating curves we'e developed based on the limiting, area.
The maximum through-wall temperature difference r'esulting in continuous heating or cooling at 100~ Fahrenheit per hour was considered.
The safety factors applied were in accordance with American Society of Mechanical Engineers Code Section III, Appendix G, 10 CFR, Part 50,. Appendix G, paragraph IU.
A.2.c and General Electric Company Topical Report, NED0-21778A, Transient Pressure Rises Affecting Fracture Toughness Requirements for BWR's."
NOTE:
Ia Salle has provided sufficient technical justification and data in Chapter 5 to demonstrate that the estimate of initial Nil-Ductilitytransition reference temperature is acceptable.
Where requested, further justification was stated in response to the reference questions.
SUS UEHANNA POSITION The Susquehanna SES position is stated in Subsection 5.3.2.
AAW/3G:ll
-e
p r/
Jt