ML20129H469
| ML20129H469 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 10/29/1996 |
| From: | Salehi K NRC (Affiliation Not Assigned) |
| To: | Hutchinson C ENTERGY OPERATIONS, INC. |
| References | |
| NUDOCS 9610310311 | |
| Download: ML20129H469 (17) | |
Text
f I
October 29, 1996 Mr. C. Randy Hutchinson Vice President, Operations AN0 Entergy Operations, Inc.
1448 S. R. 333 Russellville, AR 72801
SUBJECT:
REVIEW 0F PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF EVENT AT ARKNASAS NUCLEAR ONE - UNIT 1
Dear Mr. Hutchinson:
Enclosed for your information is a copy of the final Accident Sequence Precursor analysis of the operational event at Arkansas Nuclear One - Unit 1, reported in Licensee Event Report (LER) No. 313/95-005. contains the final analysis prepared by the Oak Ridge National Laboratory, our contractor, based on review and evaluation of your comments on the preliminary analysis and comments received from the Nuclear Regulatory Commission (NRC) staff and Sandia National Laboratories, our contractor. contains our responses to your specific comments. Our review of your comments used the criteria contained in the material which accompanied the preliminary analysis. The results of the final analysis indicate that this event is a precursor for 1995.
Please contact me at (301) 415-1367 if you have any questions regarding the enclosures. We recognize and appreciate the effort expended by you and your staff in reviewing and providing comments on the preliminary analysis.
Sincerely, Original signed by Kombiz Salehi, Acting Project Manager Project Directorate IV-1 i
Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket No.
50-313
Enclosures:
1.
Final Accident Sequence Precursor Analysis 2.
Responses to your comments cc:
See next page DISTRIBUTION:
Docket File ACRS PUBLIC OGC PD4-1 r/f JDyer, RIV JRoe KSalehi SMays, AE00 w/o encl.
EAdensam (EGA1)
CHawes P0'Reilly, AE00 w/o encl.
Document Name: ANO.LTR OI 0FC (A)PM/PD4-1 (A)LA/PD4-1 KSalehik CHawesl!1ntl NAME f
DATE
/0/2f/96 b/96
,1 j n g)
COPY dED/N0 IYESh0 g
%Y OFFICIAL RECORD COPY 9610310311 961029 PDR ADOCK 05000313 P
M atuq UNITED STATES g
g j
NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30seH001 Sg
/
October 29, 1996 Mr. C. Randy Hutchinson Vice President, Operations ANO Entergy Operations, Inc.
1448 S. R. 333 Russellville, AR 72801
SUBJECT:
REVIEW OF PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF EVENT AT ARKNASAS NUCLEAR ONE - UNIT 1
Dear Mr. Hutchinson:
Enclosed for your information is a copy of the final Accident Sequence Precursor analysis of the operational event at Arkansas Nuclear One - Unit 1, reported in Licensee Event Report (LER) No. 313/95-005.
Enclosure I contains the final analysis prepared by the Oak Ridge National Laboratory, our contractor, based on review and evaluation of your comments on the preliminary analysis and comments received from the Nuclear Regulatory Commission (NRC) staff and Sandia National Laboratories, our contractor. contains our responses to your specific comments. Our review of your comments used the criteria contained in the material which accompanied the preliminary analysis.
The results of the final analysis indicate that this event is a precursor for 1995.
Please contact me at (301) 415-1367 if you have any questions regarding the enclosures. We recognize and appreciate the effort expende.: by you and your staff in reviewing and providing comments on the preliminary analysis.
Sincerely, jfgf cW Kombiz ehi, Acting Project Manager Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket No. 50-313
Enclosures:
1.
Final Accident Sequence Precursor Analysis 2.
Responses to your comments cc w/encls: See next page I
=
= _ _
A i
i Mr. C. Randy Hutchinson Entergy Operations, Inc.
Arkansas Nuclear One, Unit 1 i
cc:
Executive Vice President Vice President, Operations Support
& Chief Operating Officer Entergy Operations, Inc.
Entergy Operations, Inc.
P. O. Box 31995 i
P. O. Box 31995 Jackson, MS 39286-1995 l
Jackson, MS 39286-199 Wise, Carter, Child & Caraway Director, Division of Radiation P. O. Box 651 Control and Emergency Management Jackson, MS 39205 Arkansas Department of Health 4815 West Markham Street, Slot 30 Little Rock, AR 72205-3867 i
Winston & Strawn i
1400 L Street, N.W.
i Washington, DC 20005-3502 Manager, Rockville Nuclear Licensing Framatone Technologies 1700 Rockville Pike, Suite 525 Rockville, MD 20852 i
Senior Resident Inspector U.S. Nuclear Regulatory Commission P. O. Box 310 4
London, AR 72847 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission i
611 Ryan Plaza Drive, Suite 400 2
Arlington, TX 76011-8064 County Judge of Pope County i
Pope County Courthouse Russellville, AR 72801 1
4 d
e LER No. 313/95-005
'LER No. 313/95-005 Event
Description:
Trip with one Emergency Feedwater (EFW) train unavailsble Date ofEvent: April 20,1995 Plant: Arkansas Nuclear One, Unit 1 Event Summan ArkansanNuclear One, Unit I (ANO 1) was operating at 100% power when a spurious trip of the main generator resulted in a main turbine trip, thereby causing an automatic trip of the reactor Multiple equipment malfunctions were expenenced, including failure of both flow control valves associated with the motor-drin:n emergency feedwater pump (MDEFWP) train. The conditional core damage probability (CCDP) estimated for this event is 6.4 x 104 Event Description Arkansas I was operating at full power when a ground fault on the B phase of the current transformer supplying the negative sequence relay (NSR) caused a generator lockout followed by turbine and reactor trips. (The NSR protects the main generator from theimal damage due to negative sequence cu.Te.nt caused by system faults or an open phase cmdition.) During the post-tip response, one main steam safety valve, PSV 2684 (see Fig.1), appeared to remain open longer than operators expected. To reduce the pressure in the B once-through-steam-generator (OTSG), operators opened the B turbine bypass valve to approximately 50%. As pressure in the B steam generator (SG) dropped, PSV-2684 seated and the B turbine bypass valve closed. PSV-2684 reopened and operators again opened the B turbine bypass valve,thereby allowing PSV-2684 to reclose. Subsequent review verified that valve PSV-2684 responded normrlly on blowdown and rescat.
l Both main feedwater pumps (MFPs) were used to maintain 3G levels, runmng back to muumum speed after the reactor i
trip, as expected. After SG levels stabilized, the MFPs should have retumed to automatic level control. The A MFP returned to automatic control as designed, but the D MFP did not. Operators manually adjusted the B MFP flow and returned it to automatic control. The B MFP failed to shift back to automatic control because foreign material (a calibration sticker) on a module co mector prevented a proper electrical connection to a relay coil.
During the first hour after the trip, condenser vacuum gradually decreased to about 20 in. Hg. The decrease was attributed to excessive air in-leakage, coupled with a failure of the B vacuum pump to automatically shift into hogging mode (higher flow rate at reduced vacuum). Operators detemuned that the excessive air in-leakage was occurring through the moisture separator rebester (MSR) relief valves. By increasing the MSR steam seal pressure and switching the B vacuum pump to hoggmg mode, the vacuum in the en wtan_=ar was recovered.
About an hour after the trip, a +5 Volt de power supply for Train A of the emergency feedwater initiation and control (EFIC) system failed. This failure, believed to be caused by component failure in the voltage regulating circuit for the power supply, resulted in a half-trip of the EFIC system. Train A SG level indication was lost, as was remote control of aernaspharic dump valve (ADV) CV-2668 and emergency feedwater valves CV 2646 and CV-2648 (see Fig. 2).
l Additional Event-Related Information 1
To adequately remove heat from the reactor core akr a scram or a trip, only one of two EFW pump trains must be 1
ENCLOSURE 1
(
I LER No. 313/95-005 available to deliver water to at least one of the two OTSGs. Le failure of the +5 Volt power supply resulted in the loss i
of EFW flow control valves in the MDEFWP train (CV 2646 and CV 2648) and ADV CV-2668 control in either
]
mua-te or manual control (local control of the ADV was still possible).
4 Modeling Assuinptions 4
About I bour after the trip, EFIC Train A failed, resulting in a loss of automatic and manual control of EFW flow control valws CV 2646 and -2648. The licensee event report (LER) for this event is not specific regarding the as-failed position 4
of the MDEFWP flow control valves and the impact of the failure on system performance. If the valves failed closed, then the auxiliary feedwater supply from the MDEFWP would be unavailable. If the valves failed full-open, then they would not be capable of regulating flow. His latter condition could eventually require the operators to trip the MDEFWP to prevent s: cam generator overfill. In this case, tripping the MDEFWP would be modeled as a recoverable system failure. Either of the above cases (failed open or failed closed) leads to the unavailability of the MDEFWP; therefore, this event was modeled as a reactor trip with flow from the MDEFWP made unavailable by failure ofits EFW flow control valves. Note that failure of the flow control valves in the open position in conjunction with operator failure i
to control SG level by tripping the MDEFWP could result in failure of the turbine-driven EFW pump (TDEFWP). This potential failure mode was not explored.
Control of EFW flow control valves CV 2646 and CV-2648 was lost when a +5 Volt de power supply in EFIC Train A failed. This failure was apparently caused by a random failure of a voltage regulator within the power supply. No information was provided that specifically indicated an increased potential for common cause failure of the flow control 4
valves in the TDEFWP train, so no increase in common cause failure probability was modeled.
l l
1 To implement the assumed failure of the MDEFWP flow control valves, the set of valves associated with the MDEFWP l
(Basic Event EFW-MOV-CF DISM) were set to TRUE (i e., the valves were failed). This setting caused the motor-driven train of the EFW to be failed in the model. The turbine-driven train was still available and was not subject to the common cause failure (i.e., loss of the +5 Volt de power sapply in EFIC Train A) that rendered the MDEFWP flow control valves inoperable. Basic event probability changes are noted in Table 1.
Analysis Results The CCDP estimated for this event is 6.4 x 104 The dominant sequence, highlighted on the event trees in Figs. 3 and 4, involves the observed trip demand with a failure to trip, and
+
failure of EFW to provide sufficient flow for ATWS mitigation.
+
The assumed moperability of MDEFWP valves increased the failure probability for the MDEFWP train.
Defmitions and probabilities for selected basic events are shown in Table 1. ne conditional probabilities associated with the highest probability sequences are shown in Table 2. Table 3 lists the sequence logic associated with the sequences listed in Table 2. Table 4 describes the system names associated with the dominant sequences. Muumal cut sets associated with the dommant sequences are shown in Table 5.
Acrooyans
]
ADV atmospheric dump valve ANOI Arkansas Nuclear One, Unit 1 2
h
LER No. 313/95-005 ATWS anticipated trsnsient without scram CCDP conditional core damage probability EFIC emergency feedwater isolation control EFW cenergency feedwater FSAR final safety analysis report LER hoensee event report MDEFWP motor-driven emergency feedwater pump MFP main feedwater pump MSR moisture separator rebester NSR negative sequence relay RCS reactor coolant system OTSO once-through steam generator SG steam generator TDEFWP turbine-driven emergency feedwater pump References 1.
LER 313/95-005," Reactor Trip Initiated by Main Turbine Generator Protective Circuitry as a Result of a Logic Circuit Ground Caused by Vibration Induced Insulation Wear," May 19,1995.
)
3
d LER No. 313/95-005 i
4 1
l 1
6 1
j l
i
(
s gIdg ne9-1r a
y OE 8
f!
a e
l a
a q qi ging y l W
/-E a
g y
h 2,2 a:le!
T ge i
.siiy i
U I
I Fig. I ANO 1 Emergency Feedwater System.
4
I i
LER No. 313/95-005 I
C Jf C
Ji a
n
,rs
,rn gvari isl.fu.4 25 25 a
+j N
4
.+
g
-v i*' 4 Z
g Z g
g2
!Z"*gl l 1'
a u
SK!g* I*ig s
- KE*K!
e a
s u
a u a
[u 1%
i;
&:i &:1 s&:
o i
e e
h.,8 l
., s i*'i-I
_i W 'i I iSl!! 'Ji[
e-c!~J[i iS 85 as e<:3, m
l sci-aw vri-me Oil-Ad Vil-eW I
'.>o-x I
B:
- coo.- j iii e
e 8
" i, i, i, a
o n
-ce.--
vi y u
-- o-
_ y, -=_3 6 N i>6"-
i n?
rw in.=
n E4,X5,8 i a
an
..... 1 g g 1,............ I i
i i[s.......
~
f6
' n 6 6!
!6 6l l6
'4 b $$
e 4
$$ i $3 Fig. 2 ANO I Emergency Feedwater System.
5
LER No. 313/95 005
$l 55558588855585885588l
.e:easeseses;
-aa g
ll l
lil 1
ill l lilnl g
g.o 811
'l i
- ]f ig llS i
i 11 i s
1 ll 1
i lll i
li 111!
\\
l Fig. 3. Domment core damage sequence for LER 313/95-005.
6 l
i l
i LER No. 313/95-005
\\
5 5
5 8
8 8
g
- aaaaaaaa, i
i j
i siI i
'l l i O o
1 X-L'
<k l
Ii 11 l
18a l' i ill l
,lli i I
Fig. 4. Anticipated transient without scram (ATWS) event tree for Arkansas Nuclear One, Unit 1.
7
LER No. 313/95-005 Table 1. Definitions and Probabluties for Selected Basic Events for LER 313/95-005 Modified Event Base Current for this manne Description probabluty probability Type event IE-LOOP Loss of Offsite Power Initiating 8.5 E 006 0.0 E+000 1GNORE Yes Event IE-STOR Steam Oenerator Tube Rupture 1.6 E-006 0.0 E+000 IGNORE Yes Initiating Event IE-SLOCA SmallIms of Coolant Accident 1.0 E-006 0.0 E+000 IGNORE Yes laitiating Event IE-TRANS Transient Initiating Event 1.3 E 004 1.0 E+000 Yes EFW-MOV CF-DlSM Emergency Feedwater (EFW) 2.6 E-004 1.0 E+000 TRUE Yes Motor-Driven Pump (MDP)
Discharge Valves Fail From Common Cause EFW-lDP-FC 1B Failure of EN Turbine-Driven 3.2 E 002 3.2 E-002 No Pump EN-XHE-NOREC Operator Fails to Recover EN 2.6 E-001 2.6 E-001 No System EFW-XHE-NOTHROT Operator Fails to Throttle EN Flow 5.0 E-003 5.0 E-003 No EN-XHE-XA-CST Operator Fails to Align a Backup 1.0 E-003 1.0 E-003 Wo Water Supply HPI CKV-OO-MST Makeup Storage Tank Suction 3.0 E 003 3.0 E-003 No Isolation Motor-Operated Valve (MOV) Common Cause Failures HPI-MDP-CF-ABC High Pressure Injection (HPI)MDP 1.1 E-004 1.1 E-004 No Common Cause Failures HPI-MOV-CF-SUCT HPI Suetiou Isolation MOV 2.6 E-004 2.6 E 004 No Common Cause Failures HPI-XHE-NOREC Operator Fails to Recover the HP1 8.4 E 001 1.0 E+000 Yes System HPI-XHE XM-HPIC Operator Fails to initiate HPI 1.0 E-002 1.0 E-002 No Cooling MFW-SYS-TRIP Main Feedwater System Trips 2.0 E 001 2.0 E-001 No MFW XHE-NOREC Operator Fails to Recover Main 1.6 E 002 1.6 E 002 No Feedwater PCS-ICC-FA-TT Failure of the Mas Turbine to Trip 1.0 E 003 1.0 E 003 No PPR MOV-OO-BLK Power Operated Relief Valve 4.0 E-003 4.0 E 003 No (PORV) Block Valve Fails to Close 8
LER No. 313/95-005 4
Table 1. Deflaitions nad Probabilities for Selected Basic Esents for LER 313,95405 1
Modined Event Base Current for this masse Description probability probability Type event PPR-SRV-CC-PORV PORV Fails to Open on Demand 6.3 E 003 6.3 E-003 No PPR-SRV CC-RCS Relief Valves Fail to 1 imit Reactor 4.4 E-004 4.4 E-004 No Coolant System Pressure 4
PPR-SRV-CO-TRAN PORV Opens During a Transient 8.0 E-002 8.0 E 002 No PPR-SRV OO PORV PORV Fails to Reclose After 3.0 E 002 3.0 E-002 No Openmg I
PPR-XHE-NOREC Operator Fails to Close the Block 1.1E-002 1.1E-002 No Valve RCS-PHN-MODPOOR Moderator Temperature Coefficient 1.4 E-002 1.4 E-002 No is not Negative Enough RPS-NONREC Nonrecowrable Reactor Protection 2.0 E-005 2.0 E 005 No System (RPS) Failures RPS-REC Recoverable RPS Failures 4.0 E-005 4.0 E 005 No RPS-XHE XM-SCRAM Operator Fails to Manually Trip the 1.0 E@2 1.0 E 002 No Reactor i
1 9
LER No. 313/95-005 Table 2. Sequence Conditional Probabilities for LER 313/95-005 1
Condkionalcore Event tree damage Percent j
name Sequence name probability (CCDP)
Contribution IRANS 21-8 5.3 E-006 82.7 TRANS 20 6.3 E-007 9.9 TRANS 21 9 3.1 E-007 4.9 Total (all sequences) 6.4 E-006
^
i f
Table 3. Sequence 14sie for Domlanat Sequences for LER 313/95-005 Event tree name Sequence name logic TRANS 21-8 RT, /RCSPRESS, EFW-ATWS TRANS 20
/RT, EFW, MFW, HPI-COOL TRANS 21 9 RT, RCSPRESS I
Table 4. System Names for LER 313/95-005 System name 14 sic EFW No orInsufficient EFW System Flow EFW-ATWS No orInsufficient EFW System Flow During an ATWS Event HPI No orInsufficient Flow from the HPI System HPI-COOL Failure to Provide HPl Cooling j
t MFW Failure of the Main Feedwater System
^
PORV PORV Opens During Transient
-t PORV-RES PORY Fails to Rescat RCSPRESS Failure to Limit RCS Pressure RT Reactor Fails to Trip During Transient 10
d LER No. 313/95 005 Table 5. Conditional Cut Sets for Higher Probability Sequences for LER 313/95-005 4
Cut set Percent Conditional l
no.
contribution probabiDty*
Cut sets' TRANS Sequence 218 5.3 E-006 Ek
~ '
+
f 1
98.0 5.2 E-006 RPS-NONREC, EFW XHE-NOREC 2
1.9 1.0 Ex <t RPS-XHLXM-SCRAK RPS-REC, EFW XHE-NOREC w*.
TRANS Sequence 20 6.3 E-007 1
41.9 2.6 E-007 EFW-MOV CF-DISM, EFW TDP FC-1B, EFW XHE-NOREC, MFW4YS TRIP, MFW XHE NOREC, HPI-XHLXM-HPlc, HPI-XHE NOREC 2
26.4 1.6 E-007 EFW-MOV-CF-DISM, EFW TDP FC 1B, EFW XHE-NOREC, MFW4YS-TRIP, MFW XHE-NOREC, PPR-SRV CC-PORV 3
12.5 7.9 E-008 EN MOV-CF DISM, EFW TDP-FC-1B, EFW XHE-NOREC, MFW4YS TRIP, MFW XHE-NOREC, HPl4KV OO-MST, HPI-XHE NOREC l
4 6.5 4.1 E-008 EFW XHE-NOT11 ROT,EN XHE-NOREC, MFW SYS-TRIP, MFW XHE-NOREC, HPI XHE-XM-HPlc, HPI XHE-NOREC 5
4.1 2.6 E-008 EFW XHLNOTHROT, EFW XHLNOREC, MFW-SYS-TRIP, MFW-XHLNOREC, PPR SRV CC-PORV 6
1.9 1.2 E-008 EFW XHLNOTHROT, EFW XHE-NOREC, MFW-SYS-TRIP, MN XHLNOREC, HPI-CKV-OO-MST, HPI XHE-NOREC 7
1.3 8.3 E-009 EFW XHLXA-CST, EFW XHE NOREC, MFW SYS-TRIP, MFW XHE-NOREC, HPI XHE XM-HPlc, HPI XHE NOREC 8
1.1 7.0 E-009 EN-MOV CF DISM, EFW TDP-FC 1B, EFW XHLNOREC, MN SYS-TRIP, MFW XHLNOREC, HPI MOV-CF SUCT, HPI-XHE.NOREC TRANS Sequence 21-9 3.1 E-007 I
88.9 2.8 E-007 RPS-NONREC, RCS-PHN-MODPOOR 2
6.3 2.0 E-008 RPS-NONREC, PCS-ICC-FA-TT 3
2.7 8.8 E-009 RfS-NONREC, PPR4RV-CC-RCS 4
1.7 5.6 E-009 RPS XHLXM-SCRAM, RPS-REC, RCS-PHN-MODPOOR Total (all sequences) 6,4 E-006
- 1he a=Aa-1 probakiny for each cut set is deternuned by andtiplying the probabddy of the intiating event by the probabildies of the basic eveses in that annunal cut est 1he probabday of the innaating everns are given in Table I and begin with the designator "IE". The probabilnies for the basic events are also giwa in Table 1.
- Basic even EFW MOV CF-DISM is a type TRUE evem and these type of evenu are normally not included in the output of fault tree reduction proyarns. This event has been added to aid in w/ - ' ; the esquences to potential cars damage associated with the event 11 Wr
I LER No. 313/95-005 LER No. 313/95-005 Event
Description:
Trip with one EFW train unavailable Date ofEvent: April 20,1995 Plant: Arkansas Nuclear One, Unit 1 Licensee Comments
Reference:
Letter from D. C. Mims, Director, Nuclear Safety, Entergy Operations, Inc. to the U.S. Nuclear Regulaicuy C -
=-:-1 Review ofPreliminary Accident Sequence PrecursorAnalysis, ICAN059606, May 31,1996.
Cominnent 1:
The event description incorrectly states that PSV-2684 " remained open longer than operators expected." The root cause section of the LER states that subsequent resiew verified that the valve responded nonnally on blowdown and rescat.
Response 1:
As the comment itself notes, the ASP analysis event desciiption did not report that PSV-2684 behaved abnormally, only that the operators believed that it was behaving abnormally. This is supported by the statement in the licensee event report (LER),"However, one valve (PSV 2684) appeared to remain open longer than normal. Operators initiated action to reduce the B Once Through Steam Generator (OTSG) pressure to assist the MSSV [ main steam safety valve) in closing." To reflect that the valve operated properly during subsequent tests, the first paragraph in the Event Description section was changed to (changes noted in italics):
Arkansas I was operating at full power when a ground fault on the B phase of the current transformer supplying the negative sequence relay (NSR) caused a generator lockout followed by turbine and reactor trips. (The NSR protects the main generator from thermal damage due to negative sequence current caused by system faults or an open phase condition.) During the post trip response, one main steam stfety valve, PSV 2684 (see Fig.1), appeared to remain open longer than operators expected.
To redme the pressure in the B Once-Through Steam-Generator (OTSG), operators opened the B turbine bypass valve to approximately 50%. As pressure in the B steam generator (SG) dropped, PSV.
2684 seated and the B turbine bypass valve closed. PSV 2684 reopened and operators again opened the B turbine bypass valve, thereby allowing PSV-2684 to reclose. Substguent review venfed that valve PSV-2684 responded nonnally on blowdown and reseat.
Comument 2:
The event description states that control of the atmospheric dump valve was lost. Adding the word
" remote" to the description clarifies that local control was still available. The same clarification may be added in the Additional Event-Related Information section.
Response 2:
This clanfication has been made. The sentence in the Event Description has been changed to (change noted in italics)" Train A SG level indication was lost, as was nmore control of atmospheric dump valve (ADV) CV 2668 and emergency feedwater valves CV 2646 and CV-2648 (see Fig. 2)." The sentence in the Additional Evest Related Inforniation section has beer. changed to "The failure of the +5 Volt power supply resulted in the loss of EFW flow control valves in the MDEFWP train (CV-12 ENCLOSURE 2
LER No. 313/95-005 2646 and CV 2648) and ADV CV 2668 control in either automatic or manual control (localcontrol of the ADVwas stillpossible)."
Comassent 3:
The Modeling Assumptions section states that the EFW control valves were not declared unavailable until about one hour after the trip, leaving the impression that they were unavailable previous to that time. A clarification that the valves actually became unavailable at that time would seem appropriate.
Response 3:
This has been clarified by deleting the sentence in question.
Comument 4:
Upon evaluating the event description, and reviewing the Event Tree and resulting cut sets, it was identified that credit was not given for use of ANO l's Auxiliary (Startup) Feedwater Pump (P 75).
The Auxiliary Feedwater pump is an electric motor-driven centrifugal pump that is normally used during startup and shutdcwn conditions when there is insufficient steam available to run the main feed pumps. This pump is credited in the ANO-1 PSA as a recovery.
The auxiliary feedwater pump can be credited in all transient sequences except for Loss of Offsite Power and Loss of Power Conversion System. The probability of failure for this recovery includes a mechanical failure probability as well as an operator failure. ANO-l's success criteria regarding the transient sequences involving loss of all feedwater defmes an available time of 36 minutes for Loss of Power Conversion System sequences with reactor coolant pumps still rumung.
Response 4:
The event has been reanalyzed, giving credit for recovering the main feedwater supply using the motor-driven startup feedwater pump. The nonrecovery probability for the main feedwater system (MFW-XHE-NOREC) was changed to reflect this. The procedures supplied with the comment letter indicate that two valves must close, at least one of these valves must re-open, and the motor-driven pump must start and run for the recovery to be successful. In addition, adequate suction supply must be provided and operators must manually (remotely) align and start the system. The system failure probability given in Attachment I to the comment letter (1.6E-02) was used in the reanalysis.
Conassent 5:
TRANS Sequence 20: 'lhe probability of failure for the Aux. Feed pump as a recovery can be applied to all cut sets in this sequence.
TRANS Sequence 21-8: The cut set for this sequence includes a failure of the EFW system.
Therefore, the Aux. Feedwater recovery can be applied.
TRANS Sequence 08: The common cause failure probability assumed for the HPI-CKV-OO MST is conservatively high based upon the ANO-1 Common Cause failure probability for these valves.
The ANO 1 Common Cause probability for these valves is 7.38E-04 based upon a beta factor of.08
.. and a probability of failure for the MOV of 9.23E-03.
For HPI-MOV-CF SUCT, the CCF probability is also higher than is representative for ANO-1. The HPI suction valves are stop check valves.. the beta factor for check valves is.06. Using ANO-l's 13
, ~....
. ~-
d LER No. 313/95-005 i
failure to open probability for these valves of 4.93E-04, yields a common cause failure probability of 2.%E-05 when the beta factor is applied..
i
]
i Response 5:
Credit for the MDSFWP is accounted for in TRANS Sequence 20 through the recovery of the MFW syssesn (MFW-XHE-NOREC). The nonrecovery probability provided by ANO 1 (0.016-see Cnmment 4) was used rather than a typical recovery class R2 nonrecovery probability (0.34), as given in Appendix A in NUREGER 4674, Vol. 21.
ATWS sequences, such as TRANS Sequence 21-8, require emergency feedwater in a very short amount of time. The procedures supplied with the enmmmt letter indicate that two valves must close, at least one of these valves must re-open, and the motor-driven pump must start and run for the recove y to be successful In addition, adequate suction supply must be provided and operators must i
manually (remotely) align and start the system Because of the time available for recove y actions, credit for the MDSFWP was given in the recovery of MFW, but not in the recovery of EFW.
1 With respect to the common cause failure probabilities (CCF) for basic events HPI-CKV-OO-MST and HPI MOV-CF-SUCT, data from many plants must be combined to estunate the probability of j
low-and moderate-frequency events because of the sparseness of data. Because of this, the model l
values will tend toward an average response for a group of plants. Regardless, because basic events HPI CKV-OO-MST and HPI-MOV-CF-SUCT do not appear in TRANS Sequences 21-8 and 219
)
(which contribute to almost 90% of the total CCDP), any change in the CCF for these basic events will not significantly affect the overall CCDP. In fact, based on IRANS Sequence 21-8 alone, this event 4
qualifies as an ASP event (i e., CCDP 2 104).
j i
I L
4 i
l J
l 2
a f
1 i
4 14 3
4
..v