Information Notice 2002-10, Manual Reactor Trip and Steam Generator Water Level Setpoint Uncertainties
| ML021820008 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 06/28/2002 |
| From: | Beckner W NRC/NRR/DRIP/RORP |
| To: | |
| Dozier J, NRR/RLSB 415-1014 | |
| References | |
| TAC M4812 IN-02-010, Suppl 1 | |
| Download: ML021820008 (15) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, DC 20555-0001 June 28, 2002 NRC INFORMATION NOTICE 2002-10, SUPPLEMENT 1: DIABLO CANYON MANUAL REACTOR TRIP AND STEAM
GENERATOR WATER LEVEL
SETPOINT UNCERTAINTIES
ADDRESSEES
All holders of operating licenses for nuclear power reactors, except those who have
permanently ceased operations and have certified that fuel has been permanently removed
from the reactor.
PURPOSE
The U.S. Nuclear Regulatory Commission (NRC) is issuing this supplement to give addressees
further information about the manual reactor trip of Diablo Canyon Unit 2 which resulted from a
failure of the main feedwater regulating valve, non-conservative steam generator setpoints and
contributing causes, and other licensee actions relating to these events. This supplement
provides information that became available after the issuance of the original information notice
(IN). The NRC expects that recipients will review the information for applicability to their
facilities and consider taking actions, as appropriate. However, this supplement does not
contain any NRC requirements and does not require any specific action or written response.
BACKGROUND
Diablo Canyon Nuclear Power Plant reported a manual reactor trip of Unit 2 which resulted from
a loss of main feedwater to a steam generator (LER 2-2002-002-00) and that the narrow-range
steam generator water level instrumentation did not respond as expected to initiate an
automatic reactor trip and emergency feedwater actuation on low-low water level in the steam
generator (LER 1-2002-001-00). The NRC issued IN 2002-10 on March 7, 2002, to inform
licensees of this event. Following the issuance of the original IN, the NRC staff conducted a
Special Inspection at Diablo Canyon, as well as a public meeting with Westinghouse. In
addition, Diablo Canyon has completed its Licensee Event Reports (LERs), Westinghouse has
issued five Nuclear Safety Advisory Letters (NSALs) relating to this phenomenon or the
presence of the void content of the two phase mixture above the mid-deck plate, and other
facilities have generated reports under Title 10, Section 50.72, of the Code of Federal
Regulations (10 CFR 50.72).
DESCRIPTION OF CIRCUMSTANCES
On February 9, 2002, with Unit 2 at full power, main feedwater regulating valve
(MFRV) FW-2-FCV-540 failed in the closed position, resulting in a rapid decrease in the water
IN 2002-10 Sup 1 level of Steam Generator 2-4. The indicated narrow-range water level decreased to 7.5 percent
and leveled out. Operators tripped the Unit 2 reactor within approximately 1 minute after the
main feedwater regulating valve closed. On February 14, 2002, after the Unit 2 restart but while
still investigating the event, the licensee identified a potentially unanalyzed condition involving
the narrow-range steam generator water level instrumentation. The licensee determined that
during the plant transient, the actual water level in steam generator 2-4 fell below the 7.2- percent narrow-range trip setpoint for engineered safety feature and reactor trip actuations
before operators manually tripped the reactor. The steam generator vendor attributed this
water-level discrepancy to a previously unaccounted for differential pressure created by steam
flow past the mid-deck plate in the moisture separator section of the steam generator. This
differential pressure phenomenon caused the steam generator narrow-range instruments to
indicate a higher-than-actual water level. Thus, the steam generator narrow-range water level
low-low setpoint was non-conservative during the loss of normal feedwater transient.
Physical Phenomenon and System Description
Steam generators designed by Westinghouse incorporate two-stage moisture separation. The
first stage uses centrifugal separators, and the second stage uses chevron-type separators. A
mid-deck divider plate separates the two stages. The steam generator water level
instrumentation uses differential pressure instruments with two ranges, a wide-range non- safety-related instrument and three or four (depending on plant) narrow-range safety-related
instruments. The wide-range instrument spans essentially the entire length of the downcomer
region, while the narrow-range instruments span only the upper 25 percent of the wide-range to
cover the normal operating band. The upper taps for all four instruments are located above the
mid-deck plate, while the lower taps are all located below this plate.
In the event at Diablo Canyon, the holes in the mid-deck, which were designed to allow
moisture removed from the second-stage separators to flow back into the downcomers, acted
as orifices which restricted steam flow and allowed pressure differences with water levels
below the mid-deck region. At higher steam flow rates with a decreasing steam generator
water level, steam exiting the first stage separators along with the moisture being separated
was enough to build up pressure below the plate that was not acting above the plate. Since the
upper steam generator water level instrument taps were connected above the plate, a pressure
difference acted on the four instruments and provided a bias that caused the instruments to
indicate a higher-than-actual level. For the limiting safety setting of the low-low steam
generator water level setpoint, this bias acts in the non-conservative direction. The magnitude
of the bias drops as the steam flow decreases.
Post Trip Analysis
Following this event, the NRC completed an onsite special team inspection at Diablo Canyon
Nuclear Power Plant. The inspection examined the events surrounding the Unit 2 reactor trip
on February 9, 2002, as they relate to safety and compliance with the Commissions rules and
regulations and the conditions of the Diablo Canyon license. The inspection consisted of
examining procedures and records, and interviewing station personnel and staff members, as
well as the reactor plant contractor. The NRCs Special Inspection Team also developed a
detailed sequence of events and organizational response time line which is summarized in the
Overview and Sequence of Events Attachment 2 to this IN.
IN 2002-10 Sup 1 This event provided an unusual challenge to the licensee in that it involved an unrecognized
phenomenon. Many plant events involve equipment behaving in an unexpected manner, but
the failure mechanisms are usually well-understood. However, a well-structured corrective
action process should still be effective under these circumstances by being sufficiently rigorous
to recognize conditions that are adverse to quality and then treating them according to their
safety significance. From a review of the post trip review, the NRCs Special Inspection Team
concluded that the licensees process was narrowly focused on finding, understanding, and
correcting the cause of the trip. While the stations post-trip analysis procedure contained steps
to review plant behavior before, during, and after the event, this was effectively not performed.
The cause of the event was readily apparent without the need to analyze plant parameters.
However, by not performing a methodical review of the plants behavior and comparing it to the
behavior expected under those conditions, the licensee failed to recognize that an automatic
plant trip and ESF actuation of auxiliary feedwater did not occur when required. Had this been
recognized, the licensee would probably have delayed restarting the plant until after the cause
and implications were understood.
The licensees review of the anomalous steam generator water level attempted to explain why
wide-range indication did not track with narrow-range indication, which was thought to have
indicated accurately. The NRCs Special Inspection Team concluded that the licensees theory
and supporting data were not compared with other available but conflicting indications. The
licensee calculated that the event would have resulted in loss of approximately 75 percent of
the initial water mass in the affected steam generator, and should have caused the wide-range
level to be 20 percent of the actual level. The licensee did not note that the bottom of the
narrow-range corresponds to approximately 75 percent wide-range level, so the narrow-range
instruments should have been expected to be reading off scale low. Also, when auxiliary
feedwater actuated, narrow-range level instruments did not show increasing level until after
some delay, confirming that actual level was well below the narrow-range.
In addressing the wide-range instrument question, it was clear that the licensee was not fully
satisfied that the issue was well-understood. However, rather than clarify the issue
immediately, the licensee used a station administrative process that required resolution of the
issue within 30 days, and declared the problem to be an issue needing validation to determine
impact on operability. The NRCs Special Inspection Team concluded that this process was not
integrated with the stations operability determination process, and could permit an issue that
was thought to relate to an operability question to be studied for 30 days before addressing the
operability question. Although this issue was resolved in 4 days, this approach was considered
to be contrary to Generic Letter 91-18, which provided guidance on the need to perform prompt
The following paragraphs present examples of corrective actions from other licensees:
Callaway
Callaway reported (EN 38740) that based on an assessment of its steam generator narrow- range low-low level trip setpoints, the existing low-low level trip at 14.8 percent did not account
for the uncertainties associated with the differential pressure created by the steam flow past the
mid-deck plate in the moisture separator section of the steam generator. A plant power
reduction was commenced at 4:58 p.m. on February 28, 2002, to decrease reactor power to
below 30 percent where engineering calculations indicate that the steam generator mid-deck
plate differential pressure condition will no longer result in a non-conservative setpoint.
IN 2002-10 Sup 1 Salem
Similar to Callaway, the low-low level trip at 9 percent did not account for the uncertainties and
Salem reduced power to 38 percent.
Sequoyah
Sequoyah personnel also performed an assessment and determined that the existing 10.7 percent low-low level trip setpoint did not account for the uncertainties associated with the
differential pressure created by the steam flow past the mid-deck plate in the moisture
separator section of the steam generator. As a conservative measure, after Westinghouse
identified this issue via NSAL 02-3, Sequoyah actuated the environmental allowance monitor
(EAM) on Friday, February 15, 2002. This feature changed the steam generator low-low level
reactor trip setpoint from 10.7 percent to 15 percent. Since the known steam generator channel
uncertainty was 8.3 percent with a narrow-range span uncertainty of 5.3 percent, Sequoyah
determined that operating with the EAMs continuously actuated would allow continued
operation.
Callaway, Salem, and Sequoyah have since recalibrated the low-low water level setpoints for
the steam generator with the additional margin to account for this newly identified error. The
licensees completed this instrument recalibration before increasing the plants power level to full
reactor power.
Conclusion
The event described in this IN highlights the potential impact of steam generator water level
setpoint errors. These errors could delay the expected automatic reactor trip and emergency
feedwater actuation. The IN identifies additional accident analyses and systems associated
with the mid-deck plate phenomenon and highlights the importance of a thorough post-trip
analysis prior to restart. The IN also provides some of the corrective actions taken because of
this event and provides information sources for further investigation.
IN 2002-10 Sup 1 This information notice does not require any specific action or written response. If you have
any questions about the information in this notice, please contact any of the technical contacts
listed below or the appropriate project manager from the NRCs Office of Nuclear Reactor
Regulation (NRR).
/RA/
William D. Beckner, Program Director
Operating Reactor Improvements Program
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
Technical contacts: Jerry Dozier, NRR Neil OKeefe, Region IV
(301) 415-1014 (361) 972-2507 Email: jxd@nrc.gov Email: nfo@nrc.gov
Hukam Garg, NRR
(301) 415-2929 Email: hcg@nrc.gov
Attachments:
1. List of References
2. Overview and Sequence of Events
3. List of Recently Issued NRC Information Notices
IN 2002-10 Sup 1 This information notice does not require any specific action or written response. If you have
any questions about the information in this notice, please contact any of the technical contacts
listed below or the appropriate project manager from the NRCs Office of Nuclear Reactor
Regulation (NRR).
/RA/
William D. Beckner, Program Director
Operating Reactor Improvements Program
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
Technical contacts: Jerry Dozier, NRR Neil OKeefe, Region IV
(301) 415-1014 (361) 972-2507 Email: jxd@nrc.gov Email: nfo@nrc.gov
Hukam Garg, NRR
(301) 415-2929 Email: hcg@nrc.gov
Attachments:
1. List of References
2. Overview and Sequence of Events
3. List of Recently Issued NRC Information Notices
DISTRIBUTION:
IN File
ADAMS ACCESSION #: *See previous concurrence
DOCUMENT NAME: G:RORP\OES\Dozier\in2002-10s1shortversion.wpd
OFFICE RSE:RORP:DRIP TECH EDITOR RSE:RIII RSE:DE:EEIB
NAME IJDozier NOKeefe* HGarg*
DATE 06 /03/2002 06/03/2002 06/03/2002 06/25/2002 OFFICE BC:DE:EEIB SC:OES:RORP:DRIP PD:RORP:DRIP
NAME JCalvo* TReis WDBeckner
DATE 06/25/2002 06/27/2002 06/28/2002 OFFICIAL RECORD COPY
Attachment 1 IN 2002-10 Sup 1 REFERENCES
LER 1-2002-001-00, Technical Specification Violation Due to Non-Conservative Steam
Generator Narrow-Range Water Level Instrumentation, Diablo Canyon Nuclear Power Plant, April 15, 2002.
LER 2-2002-002-00, Unit 2 Manual Reactor Trip Due to Loss of Main Feedwater to a Steam
Generator, Diablo Canyon Nuclear Power Plant, April 10, 2002.
NRC Special Team Inspection Reports 50-275/02-07 and 50-323/02-07 for Diablo Canyon
Nuclear Power Plant, April 8, 2002.
NRC Information Notice 2002-10, Non-Conservative Water Level Setpoints on Steam
Generators, March 7, 2002.
Westinghouse Nuclear Safety Advisory Letter NSAL-02-3, Steam Generator Mid-Deck Plate
Pressure Loss Issue, February 15, 2002.
Westinghouse Nuclear Safety Advisory Letter NSAL-02-3, Rev. 1, Steam Generator Mid-Deck
Plate Pressure Loss Issue, April 8, 2002.
Westinghouse Nuclear Safety Advisory Letter NSAL-02-4, Maximum Reliable Indicated Steam
Generator Water Level, February 19, 2002.
Westinghouse Nuclear Safety Advisory Letter NSAL-02-5, Steam Generator Water Level
Control Uncertainty Issue, February 19, 2002.
Westinghouse Steam Generator Water Level Uncertainty Issues Handout from NRC Public
Meeting in Rockville, Maryland, March 20, 2002.
Event Report 38697, Technical Specification Required Shutdown of Both Units Because
Steam Generator Water Level Instrument Channels are Inoperable Due to a Non-Conservative
Water Level Low-Low Setpoint, Diablo Canyon Nuclear Power Plant, February 14, 2002.
Event Report 38713, Safe Shutdown Capability Impacted by Non-Conservative SG NR
Setpoint, Sequoyah Nuclear Power Plant, February 20, 2002.
Event Report 38740, Safe Shutdown Capability Impacted by Non-Conservative SG NR
Setpoint, Callaway Nuclear Power Plant, February 28, 2002.
Event Report 38702, Discovery of a Non-Conservative Low-Low Level Trip Setpoint Due to a
Differential Pressure Phenomenon that Causes Steam Generator Narrow-Range Level
Channels to Read Higher Than Actual Water Level at High Steam Flows, Salem, February 15,
2002.
Attachment 2 IN 2002-10 Sup 1 Overview and Sequence of Events
This section discusses applicable events and actions before, during, and following the failure of
steam generator feedwater regulating valve number 4.
In 1989, Diablo Canyon revised the steam generator narrow-range low-low level trip setpoint in
accordance with License Amendment 34 for Unit 1 and Amendment 33 for Unit 2. The nuclear
steam supply system (NSSS) vendor (Westinghouse) provided the analysis to reduce the low- low trip setpoint in topical report, WCAP-11784, Calculation of Steam Generator Level Low and
Low-Low Trip Setpoints With Use of a Rosemount 1154 Transmitter. The licensee reduced
the setpoint from 15 to 7.2 percent narrow-range on May 10, 1989, for Unit 1, and on April 12,
1989, for Unit 2. However, the licensee did not factor the mid-deck plate differential pressure
into the narrow-range low-low level trip setpoint at this time because the mid-deck phenomenon
was unknown.
In 1998, Westinghouse recognized the mid-deck phenomenon while evaluating the design of
new (replacement) steam generators using computer modeling tools that were not available
during the design review for the original steam generators. Westinghouse began accounting
for this bias in the setpoint calculation during design work for replacement steam generators.
Westinghouse began assessing the potential impact of the mid-deck plate for original model
steam generators in late 1999. Westinghouse was planning to issue Nuclear Safety Advisory
Letters about the mid-deck phenomenon at about the time of the Diablo Canyon manual reactor
trip.
On February 9, 2002, the Main Feed Regulating Valve (MFRV) FW-2-FCV-540 failed closed, stopping the feedwater flow to steam generator 2-4. After a failed attempt to reopen the MFRV,
operators initiated a manual reactor trip. The cause of the MFRV failure was excess current in
the coil of an Asco model L206-381-6F solenoid valve (SV), which in turn caused the failure of a
power fuse and forced the MFRV to close.
During discussions with the resident inspectors after the event, the operations manager and
shift supervision expressed skepticism that the steam generator level dropped as low as
observed by the steam generator wide-range instrument during the trip. The shift technical
advisor confirmed that the wide-range level indication reached 10 percent. Diablo Canyons
Engineering Services reported that steam generator structural integrity was not affected by low
wide-range level. Engineering Services preliminarily concluded that dynamic processes
contributed to inaccurate wide-range level indication. Later that night, during a conference call
with the NRC staff to discuss the licensees plans for the restart of Unit 2, the licensee reviewed
its corrective actions for the feedwater regulating valve and other failed components. The NRC
staff expressed concern that wide-range indicated level was abnormally low for this transient.
The licensee explained its theory that the actual level was higher because of the difference
between the transient conditions (hot, dynamic) and the calibration conditions (cold, static).
The licensee believed that the steam generator narrow-range level response was normal, and
the wide-range level indication was overly conservative but did not impact operator response to
such an indication. The NRC decided to conduct follow up activities on level anomalies. The
Plant Staff Review Committee reviewed the results of the trip event response team investigation
and readiness for restart. The steam generator wide-range water level anomaly issue was
Attachment 2 IN 2002-10 Sup 1 discussed and determined not to be a restart issue. The issue was classified as an issue
needing validation to determine impact on operability (INVDIO). The Station Director granted
permission to restart the plant.
Unit 2 was restarted the next day (February 10, 2002). The licensee continued its investigation
of the Unit 2 steam generator response during the transient. Diablo Canyon personnel sent
plant trip data to Westinghouse for review. The licensee began to focus on steam generator
narrow-range indication as a potential concern. During a conference call between the licensee
and Westinghouse on February 14, 2002, Westinghouse informed the licensee about a new
process measurement error term related to mid-deck plate differential pressure that had not
been included in the existing setpoint analysis. Operators in both units declared all channels of
narrow-range level instrumentation inoperable and entered Technical Specification 3.0.3.
Operations issued a shift order to manually trip the reactor on loss of feedwater flow. Operators
in both units began reducing power to less than 60-percent thermal power to restore the
narrow-range instruments to an operable condition and exit Technical Specification 3.0.3. On
the basis of information received from Westinghouse, the licensee promptly completed an
operability assessment which concluded that the existing trip setpoint at a 7.2 percent narrow- range level remained operable at or below 60-percent reactor power. Operators stopped power
reductions at 60-percent power. This action had to be taken because the failure to correct this
condition prior to restart resulted in Unit 2 changing Modes (Mode 3 to 2 to 1) with the reactor
trip system and engineered safety system steam generator water level low-low instrumentation
inoperable and subsequent operation of Units 1 and 2 (Mode 1) in a condition prohibited by
Technical Specification 3.3.1. This Technical Specification requires that the reactor trip system
instrumentation for steam generator level low-low be operable for Modes 1 and 2 and the
engineered safety feature actuation instrumentation steam generator water level-low-low be
operable in Modes 1, 2, and 3.
On February 15, 2002, the licensee implemented setpoint changes on both units to raise the
steam generator low-low setpoint to 15 percent. After implementation, operators increased
power to 100 percent in both units. Westinghouse issued NSAL 02-3, Steam Generator Mid- Deck Plate Pressure Loss Issue on February 19, 2002. That NSAL warned plants with
Westinghouse-designed steam generators that the error source has not been accounted for
and has potentially adverse effects on steam generator level low-low uncertainty calculations as
a bias in the indicated high direction. Westinghouse further warned that for plants for which
Westinghouse maintains the calculation of record, this pressure drop effect may require a
maximum decrease of approximately 9 percent (in percent narrow-range span) in the safety
analysis limit (SAL) for establishing the low-low steam generator water level reactor trip for the
loss of normal feedwater (LONF) transient or the loss of offsite power (LOOP) transient to
compensate for this bias. NSAL 02-3 added additional transients to consider, such as the
steamline break mass and energy release, and for plants with feed line check valves inside
containment, the feedline break transient, to compensate for this described bias. Revision 1 to
the NSAL 02-3 also provided updated information regarding the steam generator water level
mid-deck plate pressure loss issue. Specifically, Westinghouse revised the NSAL to address
the impact of this issue on the feedwater line break analysis (when feedwater check valves
were located inside containment), the ATWS mitigation system actuation circuitry system, and
steamline break mass and energy release calculations. Westinghouse subsequently issued
NSAL 02-4, Maximum Reliable Indicated Steam Generator Water Level, and NSAL 02-5, Steam Generator Water Level Control System Uncertainty Issue on February 19, 2002.
Revision 1 to NSAL 02-5 was written to clarify the need to calculate the steam generator water
Attachment 2 IN 2002-10 Sup 1 level uncertainties at normal operating level, including the impact, if any, of the mid-deck plate
pressure differential and to compare the uncertainties used in the initial condition of the safety
analyses to determine if they remain bounding. NSAL 02-5, Rev. 1, also discusses the
potential impact on safety analyses performed at reactor power levels other than 100 percent
and the impact of steam generator water level uncertainty on LOCA mass and energy release.
These letters covered other effects of the same physical phenomenon as Nuclear Safety
Advisory Letter 02-3 and the void content of the two phase mixture above the mid-deck plate
(NSAL 02-4). Westinghouse also held a workshop with industry representatives on
February 28, 2002 and a public meeting with the NRC staff on March 20, 2002.
In its LER, Diablo Canyon indicated that it will submit a license amendment request to revise
the Technical Specifications to account for the mid-deck plate differential pressure in the steam
generator narrow-range low-low level protection setpoints.
Attachment 3 IN 2002-10 Sup 1 LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
_____________________________________________________________________________________
Information Date of
Notice No. Subject Issuance Issued to
_____________________________________________________________________________________
2002-22 Degraded Bearing Surfaces in 06/28/2002 All holders of operating licenses
GM/EMD Emergency Diesel for pressurized- or boiling-water
Generators nuclear power reactors, including
those that have ceased
operations but have fuel on site.
2002-21 Axial Outside-Diameter 06/25/2002 All holders of operating licenses
Cracking Affecting Thermally for pressurized-water reactors
Treated Alloy 600 Steam (PWRs), except those who have
Generator Tubing permanently ceased operations
and have certified that fuel has
been permanently removed from
the reactor.
2002-19 Medical Misadministrations 06/14/2002 All nuclear pharmacies and
Caused By Failure to Properly medical licensees.
Perform Tests on Dose
Calibrators for Beta-and Low- Energy Photon-Emitting
Radionuclides
2002-18 Effect of Adding Gas Into 06/06/2002 All holders of operating licenses
Water Storage Tanks on the for nuclear power reactors, Net Positive Suction Head For except those who have
Pumps permanently ceased operations
and have certified that fuel has
been permanently removed from
the reactor.
2002-17 Medical Use of Strontium-90 05/30/2002 All U.S. Nuclear Regulatory
Eye Applicators: New Commission medical licensees
Requirements for Calibration that use strontium-90 (Sr-90) eye
and Decay Correction applicators.
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issued by subscribing to the NRC listserver as follows:
To subscribe send an e-mail to <listproc@nrc.gov >, no subject, and the following
command in the message portion:
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______________________________________________________________________________________
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