ML052510007
| ML052510007 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 09/07/2005 |
| From: | Casto C Division Reactor Projects II |
| To: | Singer K Tennessee Valley Authority |
| References | |
| EA-05-169 IR-05-013 | |
| Download: ML052510007 (13) | |
See also: IR 05000390/2005013
Text
September 7, 2005
Tennessee Valley Authority
ATTN: Mr. K. W. Singer
Chief Nuclear Officer and
Executive Vice President
6A Lookout Place
1101 Market Street
Chattanooga, TN 37402-2801
SUBJECT:
NRC INSPECTION REPORT NO. 05000390/2005013; PRELIMINARY
GREATER THAN GREEN FINDING; WATTS BAR NUCLEAR POWER PLANT
Dear Mr. Singer:
This letter and the enclosed supporting documentation discuss a finding that appears to have
greater than very low safety significance. As described in Section 1R20.2 of NRC Inspection
Report 05000390, 391/2005002, issued on April 29, 2005, a finding was identified with respect
to procedural non-compliances at Watts Bar. The finding involved a challenge to reactor
coolant system (RCS) integrity by pressurizer power-operated relief valve (PORV) actuations
and a challenge to RCS inventory control by the loss of RCS coolant via the open PORVs. On
February 22, 2005, we determined that your staff made inappropriate operational decisions
during the transition to solid plant operations to return a charging control valve to service
following a design change and before all post-maintenance testing (PMT) was complete. As a
result of the erratic control provided by the valve, operators failed to adequately implement
procedures for solid plant operations, as required by the Watts Bar Technical Specifications,
which resulted in multiple actuations of the pressurizer PORVs in low temperature over
pressure (LTOP) mode.
On February 23, 2005, Problem Evaluation Report (PER) 77176 was initiated in the Watts Bar
Corrective Action Program for the cycling of the pressurizer PORV the previous day. The
operator log entry of the PER initiation was the first log entry that made any mention of the
charging problems and PORV lifts from the previous day. The PER also implied that only a
single lift of the PORV had occurred. The inspectors review of the reactor coolant and
charging system parameters for the period in question determined that the Cold Over-Pressure
Mitigating System was challenged by the actuation of both PORVs multiple times during a
two-hour period. The block valve for one PORV, 1-RFV-63-340A, had been closed to reduce
containment gas problems via leakage from the valve packing and as such this PORV did not
relieve actual pressure during a total of seven actuations. However, the other PORV,
1-RFV-63-334D, actuated a total of five times to reduce pressure in parallel with a group of five
actuations by 1-RFV-63-340A. The inspectors determined that the first single actuation and the
group of five/five actuations of 1-RFV-63-340A/1-RFV-63-334D were due to a failure to comply
2
with procedural requirements contained in General Operating Instruction (GO)-6, Unit
Shutdown from Hot Standby to Cold Shutdown. To transition to solid water operations, Section
5.5, Step [1] [e] states, Slowly RAISE charging to fill Pzr at less than 30 gpm. Contrary to this,
the 30-gpm requirement was exceeded resulting in the PORV actuations. The requirement was
exceeded when the normal charging flow control valve 1-FCV-62-93 exhibited erratic operation
following activities to swap from bypass to normal charging and when operators swapped back
to bypass charging.
Previous erratic control problems with 1-FCV-62-93 had resulted in a precaution and limits
statement in GO-6 stating that it may cycle with RCS pressure below 500 psig when manually
attempting to control low charging flow rates. During the transition to solid plant operations, the
RCS pressure was less than 400 psig. Additionally, the Watts Bar RCS system description
states that when the RCS is operated in the water-solid mode, the charging flow to the RCS is
to be set at a constant value. This was not consistent with the operational decision to place
1-FCV-62-93 in service, before all post-modification testing was done, during the transition to
solid plant operations.
Work Order (WO) 04-825584-000, which implemented the design change stated that the
equipment cannot be declared operable until the modification turnover package was complete.
However, the WO also allowed the valve to be returned to operation with testing to be done
later when plant conditions allow. This WO allowance was implemented on February 22, 2005
with remaining tests, including a valve stroke under high differential pressure, not yet complete.
A last PORV actuation was due to RCS heat up and resultant pressure increase from the
closure of the 1A Residual Heat Removal (RHR) heat exchanger outlet valve per System
Operating Instruction (SOI) 74.01, Residual Heat Removal, Section 8.11, Flush of A Train RHR
Heat Exchanger Bypass during Shutdown Cooling. This aspect was not described in the
original PER 77176 problem description. The following action was contained in a procedure
note, The effect on RCS heat up/cool down should be evaluated. This action was not
appropriately implemented in that the performance of Section 8.11 during solid plant operation
allowed sufficient RCS heat up to result in the actuation of the Pressurizer PORV.
The NRC has determined that the procedural noncompliances identified above represent a
performance deficiency that had an impact on safety by affecting the cold over-pressure
mitigation or low temperature over-pressure system required by the Watts Bar Technical
Specifications (TS). Specifically, TS 5.7.1.1 states that written procedures shall be
implemented and maintained covering the activities in the applicable procedures recommended
by RG 1.33, Revision 2, Appendix A, February 1978, of which Part 2.j requires a procedure for
hot standby to cold shutdown and Part 3.c requires a procedure for shutdown cooling system.
GO-6, Unit Shutdown from Hot Standby to Cold Shutdown, Section 5.5, Step [1] [e] states,
Slowly RAISE charging to fill Pressurizer at less than 30 gpm. SOI-74.01, Residual Heat
Removal, Section 8.11, implemented a flush of the A train RHR heat exchanger bypass during
shutdown cooling and contained a note which stated, The effect on RCS heatup/cool down
should be evaluated. Each procedure was not adequately implemented approaching and
during solid plant operations on February 22, 2005. This performance deficiency constitutes an
apparent violation of TS 5.7.1.1, in that, TVA failed to follow approved procedures, which
resulted in a challenge to RCS integrity by pressurizer PORV actuations and a challenge to
RCS inventory control by the loss of RCS coolant via the open PORVs. Accordingly, this
finding is identified as an Apparent Violation (AV)05000390/2005013-01, Failure to Implement
3
and Maintain Shutdown Procedures which Resulted in Pressurizer PORV Actuations. The
finding is being considered for escalated enforcement action in accordance with the "General
Statement of Policy and Procedures for NRC Enforcement Actions" (Enforcement Policy),
NUREG-1600. The current Enforcement Policy is included on the NRCs Web site at
www.nrc.gov; select What We Do, Enforcement, then Enforcement Policy.
This finding was assessed using the applicable Significance Determination Process (SDP) and
was preliminarily determined to be a Greater-Than-Green finding (i.e., a finding with some
increased importance to safety, which may require additional NRC inspection). The finding
appears to have greater than very low safety significance, primarily because the dominant SDP
scenario for this event results from a failure of the PORV to open following a demand on the
overpressure protection system (OPS) and the residual heat removal (RHR) suction relief valve
failing to close after being challenged. Subsequent failure of both RHR isolation valves or
failure of the operator to open a PORV to establish feed and bleed results in core damage. The
results of the NRCs Phase 3 SDP are attached to this letter. We will consider any additional
information you may have that could assist the NRC in making a final significance
determination.
Before we make a final decision on this matter, we are providing you an opportunity to:
(1) present to the NRC your perspectives on the apparent violation and the facts and
assumptions used by the NRC to arrive at the finding and its significance at a Regulatory
Conference or (2) submit your position on the finding to the NRC in writing. If you request a
Regulatory Conference, it should be held within 30 days of your receipt of this letter and we
encourage you to submit supporting documentation at least one week prior to the conference in
an effort to make the conference more efficient and effective. If a Regulatory Conference is
held, it will be open for public observation. The NRC will also issue a press release to
announce the conference. If you decide to submit only a written response, such a submittal
should be sent to the NRC within 30 days of the receipt of this letter.
Please contact Stephen Cahill at (404) 562-4520 within ten business days of the date of your
receipt of this letter to notify the NRC of your intentions. If we have not heard from you within
ten days, we will continue with our significance determination decision and you will be advised
by separate correspondence of the results of our deliberations on this matter.
Since the NRC has not made a final determination in this matter, no Notice of Violation is being
issued for these inspection findings at this time. In addition, please be advised that the number
and characterization of the apparent violation described in the enclosed inspection report may
change as a result of further NRC review.
4
In accordance with 10 CFR 2.390 of the NRCs "Rules of Practice," a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of NRCs document system
(ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Charles Casto, Director
Division of Reactor Projects
Docket No.:
50-390
License No.:
Enclosure: SDP Phase III Summary
cc w/encl: (See page 4)
5
cc w/encl:
Ashok S. Bhatnagar
Senior Vice President
Nuclear Operations
Tennessee Valley Authority
Electronic Mail Distribution
Larry S. Bryant, General Manager
Nuclear Engineering
Tennessee Valley Authority
Electronic Mail Distribution
Michael D. Skaggs
Site Vice President
Watts Bar Nuclear Plant
Tennessee Valley Authority
Electronic Mail Distribution
Robert J. Beecken, Vice President
Nuclear Support
Tennessee Valley Authority
Electronic Mail Distribution
General Counsel
Tennessee Valley Authority
Electronic Mail Distribution
John C. Fornicola, Manager
Nuclear Assurance and Licensing
Tennessee Valley Authority
Electronic Mail Distribution
Glenn W. Morris, Manager
Corporate Nuclear Licensing and
Industry Affairs
Tennessee Valley Authority
Electronic Mail Distribution
Paul L. Pace, Manager
Licensing and Industry Affairs
Watts Bar Nuclear Plant
Tennessee Valley Authority
Electronic Mail Distribution
Jay Laughlin, Plant Manager
Watts Bar Nuclear Plant
Tennessee Valley Authority
Electronic Mail Distribution
County Executive
Rhea County Courthouse
375 Church Street, Suite 215
Dayton, TN 37321-1300
County Mayor
P. O. Box 156
Decatur, TN 37322
Lawrence E. Nanney, Director
TN Dept. of Environment & Conservation
Division of Radiological Health
Electronic Mail Distribution
Ann Harris
341 Swing Loop
Rockwood, TN 37854
James H. Bassham, Director
Tennessee Emergency Management
Agency
Electronic Mail Distribution
Distribution w/encl: (See page 6)
_________________________
OFFICE
DRP:RII
DRP:RII
EICS:RII
DRS:RII
SPSB
SIGNATURE
SJC
SJC for
CFE
MTS per email
NAME
SCahill:aws
JBartley
CEvans
RBernhard
MTSchiltz
DATE
08/31/2005
08/31/2005
08/30/2005
08/30/2005
09/06/2005
E-MAIL COPY?
YES
NO YES
NO YES
NO YES
NO YES
NO YES
NO YES
NO
Enclosure
RISK FROM LTOP AND SHUTDOWN LOCAS
DURING WATER SOLID MODE AT WATTS BAR
The Probabilistic Safety Assessment Branch (SPSB) evaluated the risk significance of repeated
challenges to the power operated relief valves (PORVs) at Watts Bar on February 23, 2005,
when the licensee was in a water solid condition. This assessment estimated the delta increase
in core damage frequency (CDF) resulting from the (1) increased likelihood of a
low-temperature over pressure event, and (2) the increased likelihood of a shutdown
loss-of-coolant accident (LOCA) resulting from relief valves sticking open.
The dominant core damage scenario for this event results from a failure of the PORV to open
following a demand on the overpressure protection system (OPS) and the residual heat
removal (RHR) suction relief valve failing to close after being challenged. Subsequent failure of
both RHR isolation valves or failure of the operator to open a PORV to establish feed and bleed
results in core damage. This event was evaluated as having a likelihood of 2E-5.
Additionally, the likelihood of having an interfacing systems LOCA (ISLOCA) in the RHR system
from failure of the OPS to work and failure of the RHR suction relief valve was evaluated.
Credit was given to the operators by using the RHR isolation valves to isolate a postulated
rupture in the RHR system. This action reduced the CDF from ISLOCAs to 8E-7.
Description of the Event as Relayed from the Resident Inspector (RI)
On February 23, 2005, the inspectors identified a control room log entry which described the
initiation of problem event report (PER) 77176 for cycling of the pressurizer PORVs. The
repeated cycling occurred as a result of problems associated with (1) charging flow control
valve (1-FCV-62-93) erratic control, and (2) implementation of a design change notice (DCN) to
raise control air pressure on the actuator for 1-FCV-62-93. The DCN was implemented to
eliminate the erratic control of the valve. The inspectors performed a review of the reactor
coolant and charging system parameters for the period in question. The inspectors determined
that the pressurizer PORV (setpoints adjusted for cold over-pressure conditions as required by
Technical Specification 3.4.12) had actuated a total of seven times (2 single actuations and a
group of five actuations) in an approximate 2-hour period. The inspectors determined that first
single and group of five actuations were due to a failure to follow procedures regarding general
operating instruction (GOI) -6, Unit Shutdown from Hot Standby to Cold Shutdown. To
transition to solid water operations section 5.5, step [1] [e] states, Slowly RAISE charging to fill
pressurizer (PZR) at less than 30 gpm. Contrary to this, the licensee exceeded the 30 gpm
requirement and experienced the first PORV actuation. The actuation occurred when
1-FCV-62-93 exhibited erratic operation following activities to swap from bypass to normal
charging. The inspectors noted that while the DCN had been previously implemented while the
plant was on bypass charging, all of the post maintenance testing had not yet been completed.
Since 1-FCV-62-93 operation was still erratic, the licensee swapped back to bypass charging
resulting in the group of five PORV actuations.
2
The inspectors also determined that, contrary to the original PER problem description, the last
pressurizer PORV actuation was due to RCS heatup and resultant pressure increase from the
closure of the 1A RHR heat exchanger outlet valve per system operating instruction (SOI)
74.01, section 8.11, Flush of a Train RHR Heat Exchanger Bypass during Shutdown Cooling.
The inspectors determined that this procedure was not adequately maintained in that the
following action was contained in a procedure note, The effect on reactor coolant system
(RCS) heatup/cooldown should be evaluated. The performance of section 8.11 during solid
plant operation resulted in sufficient RCS heatup to result in another actuation of the PORV.
Plant Mitigation Capability and Event Details
Final safety evaluation report (FSAR) states (page 5.2-36) that one PORV is
sufficient for pressure relief considering one charging pump is charging water
into a water solid reactor coolant system at approximately 485 gpm with the
letdown path isolated. Both PORVs were available; however, one block valve
was shut, so only one PORV was credited with providing automatic over
pressure protection.
As specified in the FSAR (page 5.2-38), the power was locked out to all but one
charging pump when RCS cold leg temperatures are below 350 degrees F.
Based on information from the RI, the handswitch for the disabled charging
pump and both safety injection (SI) pumps were taken to pull-to-lock. The
disabled charging pumps discharge isolation valve and discharge bypass valve
were locked and shut. Each SI pump discharge isolation valve was locked shut
and tagged. Operations reported that it would take approximately 10 minutes to
restore the disabled charging pump and both SI pumps if needed.
The unit was shutdown at 0001 on February 22, 2005, and the first PORV
actuation occurred on 1315 on February 22, 2005. The last PORV actuation
occurred on 1508 on February 22, 2005.
At 1449, on February 22, 2005, a 12-inch diameter containment penetration, a
maintenance port, was opened. The penetration has a blind flange which is
located outside containment. The penetration had emergency closure capability
from outside containment (remove/cut all hose/cables passing through the
penetration and install the outboard blind flange) and was required to be closed
within 15 minutes of loss of shutdown cooling. Since (1) the penetration could be
closed from outside containment, (2) operators would not have to be concerned
with a degraded containment environment immediately following an extended
loss of core cooling, and (3) the penetration was required to be closed within 15
minutes, SPSB assumed containment closure could be established. Therefore,
risk from LERF was not evaluated except for ISLOCA scenarios.
Refueling water storage tank had 368,000 gal of inventory at the start of the
event.
3
The RHR Suction Relief valve lifts at 450 psig and has a capacity of 900 gal per
minute (gpm). The required flow rate is 480 gpm at 350 degrees F and 690 gpm
at 200 degrees F which is the combined flow capacity of both charging pumps.
Watts Bar RHR system is arranged with a single loop RHR suction (dual loop
discharge), and therefore has only one relief valve on a 3-inch line connected to
the 14-inch suction piping
RHR Discharge Relief Valves: Setpoints are 600 psig and both valves are on a
2-inch line connected to 8-inch piping. Downstream of the relief valves are the
pressure boundary check valves. The relief capacity of each valve is 20 gpm.
The design pressure of the RHR piping is 600 psig.
The operators had pressurizer level indication, pressurizer level low alarms, and
Based on a time of 13-hours post shutdown (decay heat estimated as 24MW),
the amount of inventory necessary to maintain boiloff was estimated as 174 gals
per minute.
PROBABILISTIC ASSESSMENT
A cold overpressure (COP) event tree was developed for this event. The endstates are:
OK - The RHR function not interrupted and reactor vessel integrity preserved.
RCS BLOWDOWN-RHR-OK - This scenario results when the PORVs lift but fail
to reseat. According to discussions with Reactor Systems, failure of a PORV to
reseat will result in loss of RCS inventory until RCS pressure reaches
atmospheric conditions. However, the inventory loss is not expected to result in
loss of inventory from the RCS hot legs and a loss of RHR pump suction. The
pressurizer is expected to remain above fifty percent full.
Since the RCS pressurizer is expected to remain above fifty percent full and the
core exit thermocouples will indicate that core cooling has been maintained, the
operators are not expected to operate the charging pump in the safety injection
mode, thereby eliminating the potential for increasing RCS pressure.
Since this event does not lead to a loss or interruption of the RHR function, this
scenario does not lead to a shutdown initiating event and was not analyzed
further.
BLOW-DOWN-LOCA - This scenario results from the RHR suction relief valve
failing to reseat after a challenge following the PORV failing to open after a
challenge. Leak path termination requires closure of one of two RHR isolation
valves which causes a loss of the RHR function. Failure of leak path termination
results in a loss of RCS inventory that leads to a loss of RHR pump suction.
This scenario is evaluated using a Watts Bar LOCA event tree to obtain a
conditional core damage frequency (delta CDF).
4
ISLOCA - Failure of the PORV and the RHR suction relief valve is assumed to
result in failure of the RHR system (starting with failure of the RHR pump seals)
once RCS pressure exceeds 1500 -1800 psig. Isolation of the break requires
closure of the RHR isolation valves.
Quantification of Cold Over Pressure Event Tree
Initiating Event - Cold Over Pressure Challenge - The initiating event, RCS pressure increase
that challenges the OPS, was quantified as a frequency. As reported by the RI, there were
seven challenges to the OPS system.
OPS@372 PSIG - Both PORVs were available for cold over pressure protection. However, one
PORV block valve was closed, so only one PORV was credited for automatic cold over
pressure protection. Failure for a PORV to open on demand was estimated as 6E-3 based on
the SPAR model for Watts Bar, Revision 3i.
OPS RESEATS - Failure of the PORV to reseat following a demand after passing water as .1
based on the SPAR model for Watts Bar, Revision 3i.
RHR-SUCTION-RV-LIFTS - Failure of the RHR suction relief valve to lift was estimated using
the same failure rate of a single RCS SRV failing to open on demand (1E-3/demand) based on
the SPAR model for Watts Bar, Revision 3i.
.
RHR-SUCTION-RV-CLOSES - Failure of the RHR suction relief valve to close following a
demand was estimated using the failure rate of a PORV failing to reseat after a challenge since
the RHR suction relief valve is designed to pass water. This failure of 3E-2 was based on the
SPAR model for Watts Bar, Revision 3i.
OP STOPS PUMPs or OPENS PORV - Failure of the operator to terminate the pressure
excursion by: (1) stopping charging flow, or (2) opening the alternate PORV and associated
block valve following failure of the RHR suction relief valve to lift was estimated as 1.0. Based
on discussions with the Division of Engineering and the Reactor Systems Branch, once a
pressure excursion has been initiated, the operator will not have enough time to respond before
the RHR system failure is expected to occur (around 1500 to 1800 psig - 2.5 times the design
pressure). Failure of the RHR system was believed to occur before failure of the reactor vessel
based on preliminary materials information from the Division of Engineering.
Using the top event values discussed earlier, the likelihood of scenarios 4 and 6 given the event
are:
Scenario 4: The likelihood of having a LOCA through the open RHR suction relief valve
was estimated as 1E-3. This scenario is further evaluated in the shutdown SDP phase 2
event tree.
Scenario 6: The likelihood of having an interfacing systems LOCA resulting from
rupture of the RHR piping was assessed as 4E-5.
5
Scenarios 4 and 6 were then analyzed using the Watts Bar LOCA event tree. Each scenario
was analyzed separately because the event tree top event probabilities were found to be
different for each scenario.
Quantification of COP Scenario 4 Using the Loss of Inventory PWR Event Trees POS 2
LOI - The likelihood of a loss of inventory from a stuck open RHR suction relief valve, COP
scenario 4, was estimated as 1E-3.
RCS injection before core damage-based on the licensees mitigation capability, both charging
pumps and both safety injection pumps were credited as being able to keep the reactor core
covered. Assuming a multi-train failure rate of 1E-3 for charging and safety injection, the failure
probability of RCS injection is driven by operator error rather than equipment failure. The
probability of operators failing to inject via available sources following a loss of the operating
train of residual heat removal system and prior to core damage is on the order of 1 x 10-4.
(Inspection Manual Chapter Appendix G Phase 2 Worksheet 2, SDP for a Westinghouse
4-Loop Plant - Loss of Level Control in POS 2). Thus, the failure probability of RCS injection
before core damage is estimated as 1E-4.
Isolate RHR and Open PORV - If the RHR suction relief valve were to stick open, it was
assumed that the operators would attempt to (1) close the RHR isolation valves, and (2) open
the alternate PORV to initiate feed and bleed. Conservatively assuming the design flow rate of
900 gpm, the operators have more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to close the valve and initiate RCS injection and
bleed through a PORV. (The remaining RWST inventory would last over 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> assuming a
charging flow rate of 300 gpm.)
To complete the recovery, the operators must close one of two RHR isolation valves and open
the alternate PORV and its block valve. (It was assumed that the PORV that responded to the
initial challenge is failed). Thus, failure for this top event is both RHR isolation valves failing to
close or the alternate PORV or its block valve failing to open. This failure likelihood was
estimated as;
(3E-3 failure for RHR isolation valve to close - Watts Bar SPAR model, revision
3i)*(.1beta factor for both RHR isolation valves failing to close due to common
cause) + (PORV block valve failing top open (3E-3 Watts BAR SPAR model,
revision 3i) + (PORV failing to open on demand 6E-3) = 9 E-3
The likelihood of the operator failing to close the RHR isolation valves and open a PORV was
estimated using HRA Worksheets for LP&SD contained in the SPAR-H methodology page B-3.
To simplify this analysis, the diagnosis probability defines the operator recovery. The inferred
definition of diagnosis is any cognitive decision making that is necessary to perform a task.
The performance shaping factors (PSFs) for this operator recovery were assumed to be:
expansive time, extreme stress, moderately complex diagnoses, and nominal procedures. It
was assumed that shutdown loss of inventory procedures would direct the operators to search
for the source of a leak, and procedures exist for using RCS injection and RCS bleed through a
PORV. All other PSFs were assumed to be nominal. Using the SPAR-H methodology, the
operator failing to isolate the RHR suction valves and initiate RCS bleed was assumed to be
1E-3. Thus, failure of this top event is assumed to be driven by equipment error and was
6
estimated as 9E-3 +1E-3 = 1E-2. Failure of this top event is assumed to lead to core damage
since it is dominated by failure of alternate PORV and its block valve failing to open.
RHR Recovery before RWST Depletion - To isolate the stuck open RHR relief valve, the RHR
system must be isolated. If the RHR system was isolated, the relief valve may reseat once
pressure in the RHR system is reduced significantly below the lift setpoint of 450 psi. If the
valve were to reseat, the RHR isolation valves could be opened, and restoration of a train of
RHR could begin. Failure of the valve not to reseat after RHR system pressure was reduced
was given a value of screening value of .5 due to lack of data. Failure of RHR recovery before
RWST depletion is assumed to be driven by failure of the stuck open relief valve to reseat.
RWST Makeup before Core Damage - One train of RWST makeup was assumed to be
available and low a RWST inventory alarm. It was assumed that the RWST makeup rate could
keep up with the RCS boiloff rate 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> into the event after the RHR system was isolated.
The failure of RWST makeup before core damage has been assigned the nominal value used
in the Shutdown SDP worksheets of 1E-2.
Referring to the Watts Bar LOCA event tree, three core damage scenarios were quantified,
sequences 3, 4, and 5. Using the top event values discussed above, the likelihood of core
damage from COP scenario 4 is:
1E-3 ((.5)*(1E-2) + (1E-2) + (1E-4)) = 1.5E-5
Quantification of COP Scenario 6 Using the Loss of Inventory PWR Event Trees POS 2
LOI - The likelihood of a loss of inventory from a interfacing systems LOCA, COP scenario 6,
was estimated as 4E-5.
RCS Injection before Core Damage - The same failure likelihood from COP Scenario 2 was
except modified for extreme stress using the SPAR-H methodology 5*(1E-4) = 5E-4.
Leak Path Terminated before RWST Depletion - If the RHR system had a rupture, it was
assumed that the operators would attempt to close the RHR isolation valves to try to isolate the
leak and open a PORV to initiate feed and bleed. Based on quantification of this top event in
COP scenario 4, this event is driven by equipment failure rather than operator error. This top
event was quantified in scenario 4 as 1E-2. Failure of this top event by failing to isolate the
RHR system or failing to establish a RCS bleed path is assumed to lead to core damage.
RHR Recovery before RWST Depletion - Since there is a rupture of the RHR system, no credit
is given for RHR recovery.
RWST Makeup before Core Damage - One train of RWST makeup was assumed to be
available and low a RWST inventory alarm. It was assumed that the RWST makeup rate could
keep up with the RCS boiloff rate 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> into the event after the RHR system was isolated.
The failure of RWST makeup before core damage has been assigned the nominal value used
in the Shutdown SDP worksheets of 1E-2.
7
Referring to the Watts Bar LOCA event tree, three core damage scenarios were quantified,
sequences 3, 4, and 5. Using the top event values discussed above, the likelihood of core
damage from COP scenario 6 is:
4E-5 ((1.0)*(1E-2) + (1E-2) +(5E-4 )) = 8E-7
CONCLUSION
The likelihood of having an interfacing systems LOCA (ISLOCA) in the RHR system from failure
of the OPS to function and failure of the RHR suction relief valve was evaluated. Credit was
given to the operators by using the RHR isolation valves to isolate a postulated rupture in the
RHR system. This action reduced the CDF from ISLOCAs to 8E-7.
The dominant core damage scenario for this event results from a failure of the PORV to open
following a demand on the OPS and the RHR suction relief valve failing to close after being
challenged. Subsequent failure of both RHR isolation valves or failure of the operator to open a
PORV to establish feed and bleed results in core damage.
The overall calculated CDF from repeated challenges to the PORVs during solid water
operations was 2E-5/year.