ML17304B445
| ML17304B445 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 09/11/1989 |
| From: | Conway W ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 161-02294-WFC-R, 161-2294-WFC-R, NUDOCS 8909150124 | |
| Download: ML17304B445 (54) | |
Text
.gg CELE RATED DJ~BUTION DKMONSTRION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:8909150124 DOC.DATE: 89/09/11 NOTARIZED: NO
'DOCKET FACIL:STN-50-528 Palo Verde Nuclear Station, Unit 1, Arizona Publi 05000528 AUTH.NAME AUTHOR AFFILIATION CONWAY,W'.F.
Arizona Public Service Co. (formerly Arizona Nuclear Power RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Forwards rev to Section 8 of Cycle 3 reload. analysis rept.
DISTRIBUTION CODE:
IE26D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: Startup Report/Refueling Report (per Tec Specs)
.F I
NOTES:
RECIPIENT ID CODE/NAME PD5 LA CHAN,T COPIES LTTR,ENCL 1
0 2
2 RECIPIENT ID CODE/NAME
'PD5 PD DAVIS,M.
COPIES LTTR ENCL 1
1 2'
05000528 j
INTERNAL: ACRS REG FI EXTERNAL: LPDR NSIC
'NOTES'TON 02 5
5 1
1 1
1 1
1 1
1 1
1 IRM TECH ADV NUDOCS-ABSTRACT RGN5 FILE 01 NRC 'PDR 1
1 1
1 1
1 1
1 NOIE 'IO ALL 'RZDS" RECIPIENTS'LEASE HELP US 'IO REZQCE %ASTE!'ONTACT 'IHE DCCUIKMZ CONTROL DESKi RQCM Pl-37 (EXP. 20079)
KO ELIMINATE YOUR?QHE FBCH DISTRIBUTION LISTS PDR IXCUIKHIS'OUDGH'T NEED!
TOTAL NUMBER OF COPIES REQUIRED:
LTTR 20 ENCL 19 I
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Arizona Public Service Company P.O. BOX 53999
~
PMOENIX,ARIZONA85072-3999 WILLIAMF.'CONWAY
,EXECUTIVEVICEPRESIDENT NUCLEAR 161-02294-'WFC/RAB September 11,, 1989 Docket No.
STN 50-528 Document Control Desk U.
S. Nuclear Regulatory Commission Hail Station Pl-37 Washington, D.
C.
20555
Reference:
Letter to the NRC from D. B. Karner,
- APS, dated January 18.,
1989; Subj'ect:
,Reload Analysis Report for Unit 1, Cycle '3 (161-01620)
Dear Sirs:
Subject:
Palo Verde Nuclear Generating Station (PVNGS)
Unit 1 Revision to Section 8 of Reload Analysis Report for Unit 1, Cycle 3'ile:
89-E-056-026 Attached is a revision to Section 8.0 of,the Unit 1, Cycle 3 Reload Analysis
- Report, which was transmitted by the referenced letter.
The revised page replaces page 8-2 of the previously reported transmittal.
Please discard the removed page.
This submittal revises the hot rod burnup from 969 MWD/MTU to 1000 MWD/MTU to be consistent with the values found in Table 8.1-1.
If you have any questions, please call A.
C. Rogers of my staff at (602) 371-4041.
Sincerely, WFC/RAB/jle Attachment cc:
J.
B. Martin T. L. Chan M. J.
Davis T. J. Polich A. C.
Gehr
~I
Burnup depen nt calculations were performed with STRIKIN-II to determine the limiting conditions for the ECCS performance analysis.
The fuel. performance da a was generated with the FATES-3A fuel evaluation model (References 8-6 and 8-7) with the NRC grain size restriction (Reference 8-8).
It was demonstrated that the burnup with the highest initial fuel stored energy was limiting.
This occurred at a hot rod burnup of IKO MMD/NTU.
The temperature and oxidation calculations were performed for the 1.0 Double-Ended Guillotine at Pump Discharge (DEG/PD) break.
This break size is the limiting break size of the reference cycle and, as the hydraulics are identical, is the limiting break size for Cycle 3.
8.1. 3 Results Signi icant core and system parameters for the preference cycle and
'PVNGS-1 Cycle 3 are shown in Table 8.1-3;.
Table 8.1-2 presents the analysis results for the 1.0 DEG/PD break which produces the highest peak clad temperature.
This limi ing case results in a peak clad temperature of 1944 F, which is weil below the accep ance limit of 22QQ F.
The maximum local and core wide zirconium oxidation, as shown in Table 8.i-2, remain well below the acceptance limi-values of 17~ and 1~, respec ively.
These results remain applicable for up to 400 tubes pluoged per s~eam oenera-or and a reduction in sys-em flow rate to 155.8X10 ibm/hr and a reduction in core flow ra:e o
151. 1x10 1 bm/hr.
The reduc ion in delivered low pressure safety injection flow (see Reference 8-11) does not impact the reflooding of the reac or vessel following a large break loss-of-cooian accident as long as there is sufficient flow from the safety injec ion pumps to maintain a full downcomer annulus following discharge of the safety injec ion tanks.
Mith the revised low pressure safe y injection flow, there is sufficient flow to maintain a full downcomer.
4l 0~
4'OCKET NO(S)
STN 50-528 s STN 50-529 and STN 50-530 August 31, "l989 DISTRIBUTION.
Docket: Fi1e PDR-
'PDR PD 5 MDLynch TChan MDavis JLee Mr. >lilliam F.
Conway Executive Vice President Arizona Nuclear Power. Prospect Post Office Box 52034 Phoenix, Arizona 85072-2034
SUBJECT:
ARIZONA PUBLIC SERVICE COMPANY'T AL PALO VERDE NUCLEAR GENERATING STATION The following documents concerning our review of the subject. facility are transmitted for your information.
DESCRIPTION OF DOCUMENT Notice of Receipt of Application Draft/Final Environmental Statement Notice of Availability of Draft/Final Environmental Statement Safety Evaluation Report, or Supplement No.
Environmental Assessment and Finding of No Significant.Impact Notice of Issuance of Environmental Assessment Notice of Consideration of,Issuance of Facility Operating License or'Amendment to Facility. Operating License Biweekly Notice; Applications and Amendments,to Operating Licenses Involvin No. Si nifleant Hazards Conditions See,Page(s)
Exemption
'DATED Construction Permit No. CPPR-Facility Operating License No.
Order Monthly Operating Report for Annual/Semi-Annual Report:
Other Amendment No.
,Amendment No.
transmitted by Letter transmitted by Letter 8/15/89 Office of Nuclear Reactor Regulation
Enclosures:
As Stated cc: See next page OFFICE>
SURNAME>
DATE>
DR
/PD5 e.
8L 89 NRC FORM 31a tto/aol NRCM 0240 OFFICIAL RECORD COPY
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 205SS August 31,.1989 DOCKET ND(S)
STN 50 528 r STN 50-529 and STN 50-530 Mr. William F.'onway Executive Vice President Arizona Nuclear Power, Project Post Office Box 52034
- Phoenix, Arizona 85072-2034
SUBJECT:
ARIZONA PUBLIC SERVICE COMPANY, ET AL PALO VERDE NUCLEAR GENERATING STATION The following documents concerning.our review of the subject facility are transmitted for your information.
DESCRIPTION OF DOCUMENT Notice of Receipt of Application Draft/Final Environmental Statement Notice of Availabilityof Draft/Final Environmental Statement Safety Evaluation Report, or Supplement No.
Environmental Assessment and Finding of No Significant Impact Notice of Issuance of Environmental Assessment Notice of Consideration of Issuance of Facility Operating License or Amendment to Facility Operating License Biweekly Notice; Applications and Amendments to Operating Licenses Involvin No Si nificant Hazards Conditions See Page(s)
DATED Exemption Construction Permit No. CPPR-Facility Operating License No.
Order Monthly Operating Report for Annual/Semi Annual Report:
Other
, Amendment No.
,Amendment No.
transmitted by Letter transmitted by Letter 8/15/89 Office of Nuclear Reactor Regulation
'tnclcsures:
As Stated cc:
See next page
~i,
~ '
4
Mr. William F.
Conway Arizona Nuclear Power Project Palo Verde CC:
Arthur C. Gehr, Esq.
Snell 8 Wilmer 3100 Valley Center Phoenix, Arizona 85073 Charles R. Kocher, Esq. Assistant Council James A. Boeletto, Esq.
Southern California Edison Company P. 0.
Box 800
- Rosemead, California 91770 Mr. Tim Polich U.S. Nuclear Regulatory Commission HC-03 Box 293-NR Buckeye, Arizona 85326 Regional Administrator, Region V
U. S. Nuclear Regulatory Commission 1450 Maria Lane Suite 210 Walnut Creek, California 94596 Mr. Charles B. Brinkman Washington Nuclear Operations Combustion Engineering, Inc.
12300 Twinbrook Parkway, Suite 330 Rockvi lie, Maryland 20852 Mr. Charles Tedford, Director Arizona Radiation Regulatory Agency 4814 South 40 Street Phoenix, Arizona 85040 Chairman Maricopa County Board of Supervisors ill South Third, Avenue Phoenix, Arizona 85003
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Docket Nos:
50-361 50-362 50-368
'50-382 50-528 50-529 50-530 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 August 15, 1989 V<EMORANDUfI FOR:
FP,Oli:
SUBJECT:
John N. Hannon, Director Project Directorate III-3 Division of Reactor Projects - III, IV, V and Special Projects M. David Lynch, Senior Project Engineer Project Directorate III-3 Division cf keacxor Projects - III, IV, V and Special'Pro'ects
SUMMARY
OF MEETING WITH THE COMBUSTION EtiGIffEEPING OWNERS GROUP (CEOG)
REGAPDING THE DEFAS DESIGN FEATURES TO BE ItfSTALLED PEP, 10 CFR 50.62 (THE ATWS RULE)
A meeting was helci in Bethesda, Maryland on July 12, 1989, between members of the tlPC staff and representatives of four licensees who form the Combustion Enaineering Owners Group (GEOG).
The four licensees are:
Louisiana Power P.
Light COII,pany '(Waterford); Arkansas Power ff Light Company (AHO-2); Southern Califcrnia Edison Company (San Onofre 2
ff 3',
ana Arizona Public Service Company (Palo Verde 1, 2 5 3).
A.list of attendees is presented in Enclosure 1.
Back round A previous meeting with the GEOG was held on May 1, 1989, tc discuss the general ciesign features of the diverse emergency feeawater actuation system (DEFAS) portiori of the ATWS equipment to be installed per the requirements of 10 CFR 50.62.
The meeting on May 1, 1989, discussed the'overall approach by the CEOG in designing the DEFAS as contained in the report,.
CE NPSD-384, which was docketed on April 30, 1989.
There was a subsequent telephone conference on June 21, 1989, between the tIRC staff and representatives of the GEOG which was focused on six concerns identified by the staff regarding the overall design features of the DEFAS. It was agreed by the parties to this telephone conference that these six concerns would form the agenda for the-meeting to be helci on July 12, 1989.
Contact:
H. Li (SICB/DEST), X-20781 D. Lynch (PD/3-3), X-23023 Y
i, iO
John N. Hannon ~Summa r The staff concluded early in the meeting on,July 12 as a result of the CEOG presentation that there would be such differences in the DEFAS equipment to be installed by the four licensees that.a final NRC acceptance of the DEFAS design features could 'only be giver. following a staff review of the plant specific submittals.
On this basis, the staff will net issue a generic SER on the CE report citea above.
However, there was sufficient information presented in the meetings or Viay 1 and July 12, 1989, to permit the staff to make specific comments on the DEFAS design features which would be common to all four licensees'lant specific designs.
The intent of the staff comments was to
.reflect the view that the general design features of the DEFAS concept presented by the GEOG was consistent with the intent of the ATWS rule.
It was clearly noted by the staff, however, that staff acceptance of the DEFAS design was continoent on a review of the plant specific submittals.
A summary of the staff's comments on the informatior presented at the two meetings citea above is presenteo below.
Enclosure 2 i' copy of the slides presented, by the GEOG cr. July 12, 1989.
Staff Comments on the GEOG DEFAS Desi
>> Features The following is the staff's understare'ng of the Diverse Emergcency Feeawater Actuatior. System (DEFAS) as presented in the met:tings held on Yay: and July 12, 1989.
The DEFAS coi sists of sensors, signal ccnditioning, trip recognition, coirc,)dence logic, initiation logic,, and other. circuitry and equipment needed tc monitor plant conditions and init>~te emergency feedwater actuation auring conditio>>s indicative of an ATWS.
The purpose of the DEFAS is to mitigate ATMS event cl nsequences by providirag a diverse mea>>s to initiate emergency aeedwater, thereby minimizing the potential for a common mode failure affecting both the reactor trip system and the existing emerge>>cy feedwater actuatioi system.
The t EFAS inl4iztion signals cause actuation of the auxiliary/emergency feeawaver pumps.and valves only if there is a demand for auxiliary/emergency feedwater actuation system (EFAS) signal and this sianal has not been generated by the plant protection system (PVS).
The occurrence of the EFAS actuation sig>>al by the PPS.
concurrent with the absence of an enable from the diverse scram system (DSSi, indicates that an ATWS condition does not exist and that emergency feedwater actuation by the DEFAS is not necessary.
Under these conditions, the DEFAS actuation will be blocked through logic in the auxiliary relay cabinet.
The staff's understanding of the functional requirements for the DEFAS is that:
DEFAS must initiaie emergency feedwater flow for conditions indicative of an A1WS where the EFAS has failed.
The DEFAS will not be required to provide mitigation of an accident such as isolating feedwater lov to a ruptured steam qenerator.
i5 ik
John N. Hannon 3
DEFAS wiH stop feedvrater flow to the affected steam generator after reaching a pre-determined level setpoint at about 30 minutes after actuation; thereafter, manual operator intervention will control the system.
DEFAS will utilize logic and redundancy to achieve a 2-out-of-2 initiation, as a minimum.
DEFAS wi 11 utilize steam generator water level as the parameter indicative of the need for EFAS actuation.
DEFAS will interface with the actuated components via the existing auxiliary relay cabinet (ARC) relays.
These relays are not used in the reactor trip system.
DEFAS will be blocked by the EFAS to prevent control/safety interactions and to disable DEFAS when the EFAS actuates.
DEFAS will be blocked by the main steam isolation system (MSIS) signal to prevent control/safety interactions and to disable the DEFAS when conditions for HSIS exist.
DEFAS will be enabled by a signal from the DSS indicating DSS actuation.
DEFAS will include capabilities to allow testing at power.
DEFAS wi 11 include features that provide alarms, plant computer data and other operator interfaces to indicate system status and meet operability requirements.
DEFAS setpoints will be coordinated with the existing PPS setpoints so that a competing condition between the PPS and DEFAS will be prevented.
DEFAS wi 11 he interfaced with existing sensors and output devices by a
fiber optic (F.O.) technique which has been approved by the NRC for nuclear plant safety related system application.
The DEFAS is fiber optically isolated via qualified devices and physically and electrically separated from the existing PPS.
It does not degrade the existing separation criteria of the PPS.
DEFAS logic will use two microprocessor based programmable logic controllers (PLC).
Each licensee will perform software verification and validation (VEV).
The record of the VhV process will be available for staff audit during the post-implementation inspection.
DEFAS equipment will be qualified for anticipated operational occurrences.
DEFAS wi 11 be designed under the suitable guality Assurance, procedures consistent with the requirements and clarification of 10 CFR 50.62 contained in Generic Letter 85-06.
~0
~ '
John N.
Mannors DEFAS logic power will be separate and independent fron the existing PPS power.
Each DEFAS logic power supply is capable of providino 120 VAC uninterruptable power for up to one hour following the loss of its power bus.
DEFAS will use a single-board computer with solid state I/O modules as contrasted with the PPS which uses analog bistable trip units.
Therefore, the DEFAS logic is diverse from the PPS.
Based on the revi'ew of information docketed on April 30, 1989 and the meeting presentations on Nay,l and July 12, 1989, the staff comented that the-proposed GEOG aesign for a diverse emergency feedwater actuation system is in general agreement with the ATWS rule and guidance published in Federal Register Vol.
49, Ho. 124, dated June 26, 1984.
However, since there may be differences in hardware equipmert between the various plants, staff acceptarice of the. DEFAS portion of the ATMS implemertation for the affected plants can only be made after receipt of the plant specific designs.
During the meeting, the followir,g technical issues were discussed; the staff positions were s,tated for each issue.
(1)
The interlock from the DSS allows the DEFAS to initiate feedwater flow only if a DSS actuation has occurred.
The staff expressed its concern whether the timing of the DSS actuation is sufficient to allow the actuation of emergency feedwater to perform its mitigation function.
The CEOG provided an analysis deroonstratino the effect of DEFAS timing on peak pressure.
The typical difference in time between the reactor system pressure reaching the RTS setpoint and reaching the DSS setpoint is about 8 seconds.
The tiniing of DEFAS actuation has a
negligible effect on the peak reactor vessel pressure for the limiting ATWS ever t.
Accordirlgly, the staff commented that the design basis of the DSS for interlocking the DEFAS initiation would be appropriate.
(2)
Power sources common for final actuation aevice between the existirg RTS arid the DEFAS.
It is the staff's understanding that the DEFAS cabinet circuitry uses independent power sources which are backed up by batteries for up to one hour.
The DEFAS inputs to the auxiliary relay cabinet are through qualified isolators.
A fault at the DEFAS cabinet will not propagate to the auxiliary relay cabinets.
The staff coirmiented that this is consistent with the intent of the ATMS rule.
However, because some components located in the auxiliary relay cabinets will be shared for both EFAS and DEFAS and hence share RPS power, it is the staff's position that each individual licensee should provide an analysis to demonstrate that power supply faults (e.g.
overvoltage and undervoltge conditions, degraded frequencies, and overcurrent will not compromise. the RTS, the EFAS or the DEFAS equipment.
This analysis should include consideration of alarms for early detection of degraded voltage and frequency conditions to allow for operator corrective actior while the affected circuits/components are still capable of performing their intende8 functions.
This will be reviewed on a plant specific basis.
(Ql
4 bohr. ti. Hannon (3)
Operator actions The DEFAS will secure feeding the affected steam generator after reaching a pre-determined level setpoint (about 30 minutes after actuation);
thereafter, manual operator intervention will control the system.
The staff commented that an operator action after 30 minutes from automatic actuation is consistent with staff policy.
(4)
Separation from existing system The DEFAS final actuation devices are common to existing emergency feedwater system.
The ATWS rule guidance states that the implementation must be such that separation criteria applied to the existing protection system are not violated.
The DEFAS will use qualified F.O. isolators for interfacing with the existing EFAS.
The separation criteria applied to the existing protection system will not be violated.
The staff commented that this is consistent with the intent of the ATWS rule.
(5)
Assumption on control system failure impact to the accident analysis.
The CEOG presented justification to show that the DEFAS design will have minimal impact on the accident analysis.
With the
- DSS, ESAS, and RSIS interlocks, the Owners Group indicated that a single failure would not cause the DEFAS to erroneously actuate such that it could adversely impact FSAR Chapter 6 ard 15 event analysis.
The staff ackrowledged that the Standard Review plan required a consideration of the effects of control sysi.eio action and inaction when assessing the trarsient response of the plant.
The staff agreed that the conceptual design proposed by the GEOG aoequately minimized the potential for improper actuation of the DEFAS during non-ATWS accident conditions.
In the course of the meeting, the CEOG asked the staff to consider reviewing a set of assumptions which would be used in performing plant specific IO CFR 50.59 analyses of modifications to be made when installing the ATWS hardware.
The staff responded that preparation of an analysis pursuant to a 10 CFR 50.59 licensee review was the sole responsibility of each licensee and that the staff would neither do a prior review nor consider approving any such analysis.
However, the staff stated that it would review the pertinent aspects of a design and analysis submitted in compliance with 10 CFR 50.62 (the ATWS rule).
In this regard, the staff indicated that its comments, as documented
- above, on the information submitted at the meeting on Nay I, 1989, and at this meeting, reflects its view that the proposed DEFAS design is in general agreement with the intent of the ATWS rule.
The staff also emphasized that the four licensees should proceed with all aspects of the plant specific des igns and ana lyses.
With regard to implementation of the DEFAS portion of the ATMS design, the staff stated its position that the licensees in attendance should proceed in an expedited manner to design, procure and install the hardware for the DEFAS While the staff will review each of the GEOG plant specific ATWS
.I~
John N. Hannon designs and issue an SER for each submittal, the staff also stated that design, procurement and implementation by the licensees of the DEFAS portion of ATWS should not be delayed pending issuance of these SERs.
The staff noted that IO CFR 50.62(d) required. each licensee to "develop and submit a proposed schedule (for implementation)...Each shall include an explanation of the schedule along with a justification if the schedule -calls for final implementation later than the second refueling outage after July 26, 1984..."
As done in prior reviews of other ATHS submittals,, the staff again stated its position that delays attributable to disagreements over minor technical points is not sufficient basis for a schedular exemption request pursuant to 10 CFR 50.62(d).
This position derives from the staff's comments on the CEOG's ASS discussions on Hay I and July 12, 1989, as documented
- above, thereby clarifying the major technical issues.
In this regard, the staff promised a relatively quick review of plant specific ATVS submittals in recognition of the differences in plant hardware between each of the affected CE plants.
N. David Lynch, Senior Project Engineer Project Directorate III-3 Division of Reactor Projects - III, IV, V
and; Specia.l Projects
iO
ENCLOSURE 1
LIST OF ATTENDEES JULY 12 1989 NRC h.
D. Lynch D. Wigginton T. Garnes V. Thomas J.
t1auck A. Thadani H. Li S. Newberry C. Poslusny W. Hodges L. Tran J'. 'Werriel b.
Hickman J.
Hannor A. llolan (EGSG)
LPEL D.
W. Gamble R.
W. Prados N. Neisner SCE I. Katter D. Hercurio J.
Redmon C. Diamond ACRS S.
Lcng CE tl. Ryan J.
Kapinos NUS t:. Cheok AFOUL H.
W. Tull R. A.,Barnes APS K. L. HcCandless Clark
0 i
PRESENTATION ON THE RESPONSE TO THE NRC REQUEST FOR ADDITIONALINFORMATION ON CE NPSD-384 DESIGN FOR A DIYERSE EMERGENCY FEEDWATER ACTUATION SYSTEM CONSISTENT WITH 10CFR50. 62 GUIDELINES ARIZONA PUBLIC SERYICE COMPANY ARKANSAS POWER AND LIGHT COMPANY LOUISIANA POWER AND LIGHT COMPANY SOUTHERN CALIFORNIA EDISON COMPANY JULY 12, 1989
PRESENTATI OUTLINE STATEMENT OF INFORMATION REQUEST 0
RESPONSE
TO QUESTION 0
DISCUSSION 0
REQUESTED NRC 'POSITIONS
i 0
I
UE TION 1 PROVIDE AN ANALYSIS FOR AN ASS TO ILLUSTRATE THAT THE TIMING OF THE DSS ACTUATION IS SUFFI-CIENT TO.ALLOW THE ACTUATION OF EMERGENCY FEEDWATER FOR MITIGATION E
PON E
CENPD-158, REVISION 1 CONCLUDES THAT AUX. FEED.
DELIVERY HAS NO IMPACT ON THE LIMITING EVENT OR THE PEAK RCS PRESSURE CENPD-263 CONCLUDES THAT THE TIMING OF AUX.
FEED.
DELIVERY HAS A SMALL IMPACT ON THE LIMITING ATWS EVENT SUBSEQUENT ANALYSES PERFORMED TO.DETERMINE THE SENSITIVITY OF DEFAS TIMING ON PEAK PRESSURE SHOWS NEGLIGIBLE EFFECT ON PEAK PRESSURE FOR LIMITING ATWS EVENT
SEQUENCE OF EVENTS LOFM ATMS MITH DSS BUT NO TRIP TIME SE 0.0 37.6 62.0 86.6 90.3 114.7 135.0 159.4 EVElg LOSS OF ALL NORMAL FEEDWATER.
LOM SG LEVEL AUXILIARYFEEDWATER ACTUATION SIGNAL DSS SETPOINT REACHED MAXIMUM RCS PRESSURE AUX. FEED.
DELIVERED FOR SONGS 2
& 3 AUX. FEED DELIVERED FOR WSES-3 DEFAS INITIATED FLOM DELIVERED SONGS 2&3.
DEFAS INITIATED FLOM DELIVERED FOR WSES-3 AUX. FEED DELIVERED FOR WSES-3 AUX. FEED.
DELIVERED FOR ANO-2 DEFAS INITIATED FLOW DELIVERED FOR ANO-2
i
SEQUENCE OF EVENTS LOFW ASS WITH DSS BUT TRIP hVr LA TIME E
0.0 22.8 32.0 68.8 78, R'2.0
/V+ET LOSS OF ALL NORMAL FEEDWATER LOW SG LEVEL AUXILIARYFEEDWATER ACTUATION SIGNAL DSS SETPOINT REACHED AUX. FEED DELIVERED pc.=os MAXIMUM RCS PRESSURE
0
AUXILIARY FEEDWATER TIMING E
ITIVITY ASSUMED PLANT CLASS LLSG SIG.
AFM DELIVERY PEAK PRESSURE (SEC)
(SEC)
(PSIA) 3410 l4tr 3410 Sh 38 38 58*
4250 4290 3800 Rh 3800 Sh 23 23 3800 3820
- NOT ACHIEVABLE.
FOR DEMONSTRATION PURPOSES ONLY.
+* AUXILIARYFEEDWATER INITIATED AFTER THE TIME OF PEAK PRESSURE
, ~"
GUE TIO 2
PROVIDE A DISCUSSION OF SGLL AS AN ALTERNATIVE TO THE DSS INTERLOCK REAL ISSUE MILL EARLIER AUX. FEED ACTUATION MITIGATE AN ATWS EVENT FOR LATER TIMES IN THE CYCLE RESPQN E
FOR LIMITING ATWS SCENARIO, AUX. FEED TIMING HAS LITTLE IMPACT ON PEAK PRESSURE FOR THE 3410 MMr CLASS THERE IS NO TIME IN THE CYCLE WHICH YIELDS ATWS PEAK PRESSURES BELOM LEVEL C STRESS LIMITS (CENPD-263)
FOR THE 3800 Nh CLASS THERE MAY BE A SMALL IMPACT ON PEAK PRESSURE FOR LATER TIMES IN CORE CYCLE, I.E.,
BELOW LEVEL C STRESS LIMITS (CENPD-263)
0
,l
3410 %fr PLANT CLASS LOFM ASS PEAK PRESSURE YERSUS HOOERATOR TEMPERATURE COEFFICIEN 4000
~
3500
~ 3I l
HTC 9 504 CYCLE LIFE RCS PRESSURE PZR PRESSURE FLANGE LEAKAGE LEYEL C STRESS LIMIT 2500 0
-1.0
-2.0
-3.0 HODERATOR TEMPERATURE COEFFICIENT, 10-4 DRHO/F
lh
3800 Sir PLANT CLASS LOR4 ATWS PEAK PRESSURE VERSUS HOOERATOR TEMPERATURE COEFFICIENT 5000 4500
~ 4000
~~ 3500 t-j P 3000L FLANGE LEAKAGE MTC 9 50 o CYCL'E LIFE
~A RCS,PRESSURE LEVEL C STRESS LIMIT,,
PZR PRESSURE 2500 I
I I
I I
-0.5
-1.0
-1.5
-2.0
-2.5
-3.0 MODERATOR TEMPERATURE COEFFICIENT, 10-4 ORHO/F 9
f.
QUESTION 3 TESTING CAPABILITIES
RESPONSE
TEST PROCEDURES MILL BE DETERMINED ONCE THE FINAL DESIGN IS ESTABLISHED ON.A PLANT SPECIFIC BASIS 10
QUESTION 4 VRV PROGRAM. FOR PROGRAMMABLE LOGIC CONTROLLERS RE PO E
WSES DESIGN DOES NOT USE PLCs
'VIV PROGRAM MILL BE ESTABLISHED ON A PLANT SPECIFIC BASIS AT AN APPROPRIATE LEVEL FOR NON-SAFETY SYSTEMS
UE TIO 5
CURRENT PLANS AND PROCEDURES FOR AMSAC (DEFAS)
PLANS UNDER CONSIDERATION:
IF FEASIBLE, REPAIR AT POWER ON A SCHEDULE CONSISTENT WITH SAFETY SIGNIFICANCE IF NOT FEASIBLE, REPAIR AND PLACE IN SERVICE UPON ENTERING MODE 1 AFTER NEXT REFUELING OUTAGE IF NOT REPAIRABLE DURING THE OUTAGE, DETERMINE LONG-TERM CORRECTIVE ACTIONS
0 E
QUESTIO ASSUMPTIONS REGARDING CONTROL SYSTEM FAILURES AND IMPACT ON 10CFR50.59 NEGATIYE FINDING FOR INSTALLATION RESP E
IMPACT ON CHAPTER 15 EVENTS COMMON MODE FAILURE POSTULATED BY ATWS RULE NOT ASSUMED A SINGLE FAILURE MILL NOT CAUSE THE DEFAS TO ADVERSELY IMPACT CHAPTER 6 AND 15 EVENTS 13
i~
II
EQUEST F
R NR POSITI S
0 CE NPSD-384, SECTION 5 CONCERNS:
APPLICATION OF 10CFR50.59 VERSUS SRP SECTION 7.7 POWER SOURCES COtOOM FOR FINAL ACTUATION DEVICE BETWEEN EXISTING RTS AND DEFAS SEPARATION FROM EXISTING SYSTEM-DEFAS FINAL ACTUATION DEViCE IS COMMON TO EXISTING AUX. FEED SYSTEM OPERATOR ACTION REQUIRED AFTER DEFAS HIGH SG LEVEL SETPOINT REACHED DOCUMENTED NRC POSITIONS TO FACILITATE DESIGN AND 'IMPLEMENTATION 14
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