Information Notice 2002-10, Manual Reactor Trip and Steam Generator Water Level Setpoint Uncertainties: Difference between revisions

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{{#Wiki_filter:UNITED STATESNUCLEAR REGULATORY COMMISSIONOFFICE OF NUCLEAR REACTOR REGULATIONWASHINGTON, DC 20555-0001June 28, 2002NRC INFORMATION NOTICE 2002-10, SUPPLEMENT 1:DIABLO CANYON MANUALREACTOR TRIP AND STEAM
{{#Wiki_filter: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


===GENERATOR WATER LEVEL===
SETPOINT UNCERTAINTIES
SETPOINT UNCERTAINTIES


==ADDRESSEES==
==ADDRESSEES==
All holders of operating licenses for nuclear power reactors, except those who havepermanently ceased operations and have certified that fuel has been permanently removed
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.
from the reactor.


==PURPOSE==
==PURPOSE==
The U.S. Nuclear Regulatory Commission (NRC) is issuing this supplement to give addresseesfurther information about the manual reactor trip of Diablo Canyon Unit 2 which resulted from a
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
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
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
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
(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
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.
contain any NRC requirements and does not require any specific action or written response.


==BACKGROUND==
==BACKGROUND==
Diablo Canyon Nuclear Power Plant reported a manual reactor trip of Unit 2 which resulted froma loss of main feedwater to a steam generator (LER 2-2002-002-00) and that the narrow-range
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
steam generator water level instrumentation did not respond as expected to initiate an
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automatic reactor trip and emergency feedwater actuation on low-low water level in the steam
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
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
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
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
addition, Diablo Canyon has completed its Licensee Event Reports (LERs), Westinghouse has
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presence of the void content of the two phase mixture above the mid-deck plate, and other
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 FederalRegulations (10 CFR 50.72).
facilities have generated reports under Title 10, Section 50.72, of the Code of Federal
 
Regulations (10 CFR 50.72).


==DESCRIPTION OF CIRCUMSTANCES==
==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 waterML021820008 IN 2002-10 Sup 1 level of Steam Generator 2-4.  The indicated narrow-range water level decreased to 7.5 percentand leveled out.  Operators tripped the Unit 2 reactor within approximately 1 minute after the
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


main feedwater regulating valve closed. On February 14, 2002, after the Unit 2 restart but while
ML021820008
 
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
still investigating the event, the licensee identified a potentially unanalyzed condition involving


the narrow-range steam generator water level instrumentation. The licensee determined that
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
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
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
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
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
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
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 DescriptionSteam generators designed by Westinghouse incorporate two-stage moisture separation.  Thefirst stage uses centrifugal separators, and the second stage uses chevron-type separators.  A
low-low setpoint was non-conservative during the loss of normal feedwater transient.


mid-deck divider plate separates the two stages. The steam generator water level
===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
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
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
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
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


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 allowmoisture removed from the second-stage separators to flow back into the downcomers, acted
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
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
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
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
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
upper steam generator water level instrument taps were connected above the plate, a pressure
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difference acted on the four instruments and provided a bias that caused the instruments to
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
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
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 AnalysisFollowing this event, the NRC completed an onsite special team inspection at Diablo CanyonNuclear Power Plant.  The inspection examined the events surrounding the Unit 2 reactor trip
of the bias drops as the steam flow decreases.


on February 9, 2002, as they relate to safety and compliance with the Commission's rules and
===Post Trip Analysis===
Following this event, the NRC completed an onsite special team inspection at Diablo Canyon


regulations and the conditions of the Diablo Canyon license. The inspection consisted of
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
examining procedures and records, and interviewing station personnel and staff members, as


well as the reactor plant contractor. The NRC's Special Inspection Team also developed a
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
detailed sequence of events and organizational response time line which is summarized in the


"Overview and Sequence of Events" Attachment 2 to this IN.
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 unrecognizedphenomenon.  Many plant events involve equipment behaving in an unexpected manner, but
IN 2002-10 Sup 1 This event provided an unusual challenge to the licensee in that it involved an unrecognized


the failure mechanisms are usually well-understood. However, a well-structured corrective
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
action process should still be effective under these circumstances by being sufficiently rigorous
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to recognize conditions that are adverse to quality and then treating them according to their
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 NRC
safety significance. From a review of the post trip review, the NRCs Special Inspection Team


's Special Inspection Teamconcluded that the licensee
concluded that the licensees process was narrowly focused on finding, understanding, and


's process was narrowly focused on finding, understanding, andcorrecting the cause of the trip. While the station
correcting the cause of the trip. While the stations post-trip analysis procedure contained steps


's post-trip analysis procedure contained stepsto review plant behavior before, during, and after the event, this was effectively not performed.
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.
The cause of the event was readily apparent without the need to analyze plant parameters.


However, by not performing a methodical review of the plant
However, by not performing a methodical review of the plants behavior and comparing it to the


's behavior and comparing it to thebehavior expected under those conditions, the licensee failed to recognize that an automatic
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
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
recognized, the licensee would probably have delayed restarting the plant until after the cause


and implications were understood. The licensee
and implications were understood.


's review of the anomalous steam generator water level attempted to explain whywide-range indication did not track with narrow-range indication, which was thought to have
The licensees review of the anomalous steam generator water level attempted to explain why


indicated accurately.  The NRC
wide-range indication did not track with narrow-range indication, which was thought to have


's Special Inspection Team concluded that the licensee
indicated accurately. The NRCs Special Inspection Team concluded that the licensees theory


's theoryand supporting data were not compared with other available but conflicting indications. The
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
licensee calculated that the event would have resulted in loss of approximately 75 percent of
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the initial water mass in the affected steam generator, and should have caused the wide-range
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
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
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
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
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 fullysatisfied that the issue was well-understood. However, rather than clarify the issue
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
immediately, the licensee used a station administrative process that required resolution of the
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issue within 30 days, and declared the problem to be an issue needing validation to determine
issue within 30 days, and declared the problem to be an issue needing validation to determine


impact on operability. The NRC
impact on operability. The NRCs Special Inspection Team concluded that this process was not


's Special Inspection Team concluded that this process was notintegrated with the station
integrated with the stations operability determination process, and could permit an issue that


's operability determination process, and could permit an issue thatwas thought to relate to an operability question to be studied for 30 days before addressing the
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
operability question. Although this issue was resolved in 4 days, this approach was considered
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to be contrary to Generic Letter 91-18, which provided guidance on the need to perform prompt
to be contrary to Generic Letter 91-18, which provided guidance on the need to perform prompt


operability assessments. The following paragraphs present examples of corrective actions from other licensees:CallawayCallaway 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
operability assessments.
 
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
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
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
reduction was commenced at 4:58 p.m. on February 28, 2002, to decrease reactor power to
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plate differential pressure condition will no longer result in a non-conservative setpoint.
plate differential pressure condition will no longer result in a non-conservative setpoint.


IN 2002-10 Sup 1 SalemSimilar to Callaway, the low-low level trip at 9 percent did not account for the uncertainties andSalem reduced power to 38 percent.SequoyahSequoyah personnel also performed an assessment and determined that the existing 10.7percent low-low level trip setpoint did not account for the uncertainties associated with the
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
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
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
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
(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
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
uncertainty was 8.3 percent with a narrow-range span uncertainty of 5.3 percent, Sequoyah
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determined that operating with the EAMs continuously actuated would allow continued
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 forthe steam generator with the additional margin to account for this newly identified error. The
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
licensees completed this instrument recalibration before increasing the plants power level to full


' power level to fullreactor power. ConclusionThe event described in this IN highlights the potential impact of steam generator water levelsetpoint errors.  These errors could delay the expected automatic reactor trip and emergency
reactor power.


feedwater actuation. The IN identifies additional accident analyses and systems associated
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
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
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.
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 haveany questions about the information in this notice, please contact any of the technical contacts
IN 2002-10 Sup 1 This information notice does not require any specific action or written response. If you have


listed below or the appropriate project manager from the NRC
any questions about the information in this notice, please contact any of the technical contacts


's Office of Nuclear ReactorRegulation (NRR). /RA/William D. Beckner, Program Director
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


===Operating Reactor Improvements Program===
Division of Regulatory Improvement Programs
Division of Regulatory Improvement Programs


Office of Nuclear Reactor RegulationTechnical contacts:Jerry Dozier, NRRNeil O
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


'Keefe, Region IV(301) 415-1014(361) 972-2507 Email:  jxd@nrc.govEmail:  nfo@nrc.govHukam Garg, NRR(301) 415-2929 Email:  hcg@nrc.gov
3. List of Recently Issued NRC Information Notices


Attachments: 1. List of References
IN 2002-10 Sup 1 This information notice does not require any specific action or written response. If you have


2.  Overview and Sequence of Events
any questions about the information in this notice, please contact any of the technical contacts


3.  List of Recently Issued NRC Information Notices
listed below or the appropriate project manager from the NRCs Office of Nuclear Reactor


IN 2002-10 Sup 1 This information notice does not require any specific action or written response. If you haveany questions about the information in this notice, please contact any of the technical contacts
Regulation (NRR).


listed below or the appropriate project manager from the NRC
/RA/
                                      William D. Beckner, Program Director


's Office of Nuclear ReactorRegulation (NRR). /RA/William D. Beckner, Program Director
Operating Reactor Improvements Program


===Operating Reactor Improvements Program===
Division of Regulatory Improvement Programs
Division of Regulatory Improvement Programs


Office of Nuclear Reactor RegulationTechnical contacts:Jerry Dozier, NRRNeil O
Office of Nuclear Reactor Regulation
 
Technical contacts:   Jerry Dozier, NRR            Neil OKeefe, Region IV


'Keefe, Region IV(301) 415-1014(361) 972-2507 Email: jxd@nrc.govEmail: nfo@nrc.govHukam Garg, NRR(301) 415-2929 Email:  hcg@nrc.gov
(301) 415-1014               (361) 972-2507 Email: jxd@nrc.gov            Email: nfo@nrc.gov


Attachments: 1.  List of References
Hukam Garg, NRR


2. Overview and Sequence of Events
(301) 415-2929 Email: hcg@nrc.gov


3. List of Recently Issued NRC Information NoticesDISTRIBUTION
Attachments:
1. List of References


:ADAMS
2. Overview and Sequence of Events


IN FileADAMS ACCESSION #:*See previous concurrenceDOCUMENT NAME: G:RORP\OES\Dozier\in2002-10s1shortversion.wpdOFFICERSE:RORP:DRIPTECH EDITORRSE:RIIIRSE:DE:EEIBNAMEIJDozierNO
3. List of Recently Issued NRC Information Notices


'Keefe*HGarg*DATE06 /03/2002 06/03/2002 06/03/200206/25/2002OFFICEBC:DE:EEIBSC:OES:RORP:DRIPPD:RORP:DRIPNAMEJCalvo*TReisWDBecknerDATE06/25/200206/27/200206/28/2002OFFICIAL RECORD COPY
DISTRIBUTION:
ADAMS


Attachment 1IN 2002-10 Sup 1 REFERENCESLER 1-2002-001-00, "Technical Specification Violation Due to Non-Conservative SteamGenerator 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 SteamGenerator," Diablo Canyon Nuclear Power Plant, April 10, 2002.NRC Special Team Inspection Reports 50-275/02-07 and 50-323/02-07 for Diablo CanyonNuclear Power Plant, April 8, 2002.NRC Information Notice 2002-10, "Non-Conservative Water Level Setpoints on SteamGenerators," March 7, 2002.Westinghouse Nuclear Safety Advisory Letter NSAL-02-3, "Steam Generator Mid-Deck Plate
IN File


Pressure Loss Issue," February 15, 2002.Westinghouse Nuclear Safety Advisory Letter NSAL-02-3, Rev. 1, "Steam Generator Mid-DeckPlate Pressure Loss Issue," April 8, 2002.Westinghouse Nuclear Safety Advisory Letter NSAL-02-4, "Maximum Reliable Indicated SteamGenerator Water Level," February 19, 2002.Westinghouse Nuclear Safety Advisory Letter NSAL-02-5, "Steam Generator Water LevelControl Uncertainty Issue," February 19, 2002.Westinghouse Steam Generator Water Level Uncertainty Issues Handout from NRC PublicMeeting in Rockville, Maryland, March 20, 2002.Event Report 38697, "Technical Specification Required Shutdown of Both Units BecauseSteam Generator Water Level Instrument Channels are Inoperable Due to a Non-Conservative
ADAMS ACCESSION #:                                                    *See previous concurrence


Water Level Low-Low Setpoint," Diablo Canyon Nuclear Power Plant, February 14, 2002.Event Report 38713, "Safe Shutdown Capability Impacted by Non-Conservative SG NRSetpoint," Sequoyah Nuclear Power Plant, February 20, 2002.Event Report 38740, "Safe Shutdown Capability Impacted by Non-Conservative SG NRSetpoint," Callaway Nuclear Power Plant, February 28, 2002.Event Report 38702, "Discovery of a Non-Conservative Low-Low Level Trip Setpoint Due to aDifferential Pressure Phenomenon that Causes Steam Generator Narrow-Range Level
DOCUMENT NAME: G:RORP\OES\Dozier\in2002-10s1shortversion.wpd


Channels to Read Higher Than Actual Water Level at High Steam Flows," Salem, February 15, 2002.
OFFICE RSE:RORP:DRIP TECH EDITOR                      RSE:RIII          RSE:DE:EEIB


Attachment 2IN 2002-10 Sup 1 Overview and Sequence of EventsThis section discusses applicable events and actions before, during, and following the failure ofsteam generator feedwater regulating valve number 4. In 1989, Diablo Canyon revised the steam generator narrow-range low-low level trip setpoint inaccordance with License Amendment 34 for Unit 1 and Amendment 33 for Unit 2.  The nuclear
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


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 andLow-Low Trip Setpoints With Use of a Rosemount 1154 Transmitter.
NAME JCalvo*                TReis                    WDBeckner


"  The licensee reducedthe setpoint from 15 to 7.2 percent narrow-range on May 10, 1989, for Unit 1, and on April 12,
DATE      06/25/2002        06/27/2002              06/28/2002 OFFICIAL RECORD COPY
1989, for Unit 2. However, the licensee did not factor the mid-deck plate differential pressure
 
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
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 ofnew (replacement) steam generators using computer modeling tools that were not available
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
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.
for this bias in the setpoint calculation during design work for replacement steam generators.
Line 301: Line 472:
Westinghouse began assessing the potential impact of the mid-deck plate for original model
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
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
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,
trip.
operators initiated a manual reactor trip. The cause of the MFRV failure was excess current in
 
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
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 andshift supervision expressed skepticism that the steam generator level dropped as low as
power fuse and forced the MFRV to close.


observed by the steam generator wide-range instrument during the trip.  The shift technical
During discussions with the resident inspectors after the event, the operations manager and


advisor confirmed that the wide-range level indication reached 10 percent.  Diablo Canyon
shift supervision expressed skepticism that the steam generator level dropped as low as


'sEngineering Services reported that steam generator structural integrity was not affected by low
observed by the steam generator wide-range instrument during the trip. The shift technical


wide-range level. Engineering Services preliminarily concluded that dynamic processes
advisor confirmed that the wide-range level indication reached 10 percent. Diablo Canyons


contributed to inaccurate wide-range level indication.  Later that night, during a conference call
Engineering Services reported that steam generator structural integrity was not affected by low


with the NRC staff to discuss the licensee
wide-range level. Engineering Services preliminarily concluded that dynamic processes


's plans for the restart of Unit 2, the licensee reviewedits corrective actions for the feedwater regulating valve and other failed components. The NRC
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.
staff expressed concern that wide-range indicated level was abnormally low for this transient.
Line 336: Line 513:
the wide-range level indication was overly conservative but did not impact operator response to
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
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
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
and readiness for restart. The steam generator wide-range water level anomaly issue was


Attachment 2IN 2002-10 Sup 1 discussed and determined not to be a restart issue. The issue was classified as an issueneeding validation to determine impact on operability (INVDIO).  The Station Director granted
Attachment 2 IN 2002-10 Sup 1 discussed and determined not to be a restart issue. The issue was classified as an issue


permission to restart the plant.  Unit 2 was restarted the next day (February 10, 2002). The licensee continued its investigationof the Unit 2 steam generator response during the transient.  Diablo Canyon personnel sent
needing validation to determine impact on operability (INVDIO). The Station Director granted


plant trip data to Westinghouse for review. The licensee began to focus on steam generator
permission to restart the plant.


narrow-range indication as a potential concern. During a conference call between the licensee
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
and Westinghouse on February 14, 2002, Westinghouse informed the licensee about a new
Line 354: Line 537:
process measurement error term related to mid-deck plate differential pressure that had not
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
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.
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
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
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
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
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
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
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
condition prior to restart resulted in Unit 2 changing Modes (Mode 3 to 2 to 1) with the reactor
Line 376: Line 559:
inoperable and subsequent operation of Units 1 and 2 (Mode 1) in a condition prohibited by
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
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
instrumentation for steam generator level low-low be operable for Modes 1 and 2 and the
Line 382: Line 565:
engineered safety feature actuation instrumentation steam generator water level-low-low be
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 thesteam generator low-low setpoint to 15 percent. After implementation, operators increased
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
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


" on February 19, 2002.  That NSAL warned plants withWestinghouse-designed steam generators that the error source has not been accounted for
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
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
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
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
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
analysis limit (SAL) for establishing the low-low steam generator water level reactor trip for the
Line 400: Line 587:
loss of normal feedwater (LONF) transient or the loss of offsite power (LOOP) transient to
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
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
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
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
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
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
the impact of this issue on the feedwater line break analysis (when feedwater check valves
Line 414: Line 601:
were located inside containment), the ATWS mitigation system actuation circuitry system, and
were located inside containment), the ATWS mitigation system actuation circuitry system, and


steamline break mass and energy release calculations. Westinghouse subsequently issued
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
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.


" on February 19, 2002.Revision 1 to NSAL 02-5 was written to clarify the need to calculate the steam generator water
Revision 1 to NSAL 02-5 was written to clarify the need to calculate the steam generator water


Attachment 2IN 2002-10 Sup 1 level uncertainties at normal operating level, including the impact, if any, of the mid-deck platepressure differential and to compare the uncertainties used in the initial condition of the safety
Attachment 2 IN 2002-10 Sup 1 level uncertainties at normal operating level, including the impact, if any, of the mid-deck plate


analyses to determine if they remain bounding. NSAL 02-5, Rev. 1, also discusses the
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
potential impact on safety analyses performed at reactor power levels other than 100 percent
Line 432: Line 621:
Advisory Letter 02-3 and the void content of the two phase mixture above the mid-deck plate
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
(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


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 revisethe Technical Specifications to account for the mid-deck plate differential pressure in the steam
the Technical Specifications to account for the mid-deck plate differential pressure in the steam


generator narrow-range low-low level protection setpoints.
generator narrow-range low-low level protection setpoints.


______________________________________________________________________________________OL = Operating License
Attachment 3 IN 2002-10 Sup 1 LIST OF RECENTLY ISSUED
 
NRC INFORMATION NOTICES


CP = Construction PermitAttachment 3IN 2002-10 Sup 1 LIST OF RECENTLY ISSUEDNRC INFORMATION NOTICES
_____________________________________________________________________________________
Information                                              Date of


_____________________________________________________________________________________InformationDate of
Notice No.              Subject                          Issuance        Issued to


===Notice No.        SubjectIssuanceIssued to===
_____________________________________________________________________________________
_____________________________________________________________________________________2002-22Degraded Bearing Surfaces inGM/EMD Emergency Diesel
2002-22          Degraded Bearing Surfaces in          06/28/2002      All holders of operating licenses


Generators06/28/2002All holders of operating licensesfor pressurized- or boiling-water
GM/EMD Emergency Diesel                                for pressurized- or boiling-water


nuclear power reactors, including
Generators                                            nuclear power reactors, including


those that have ceased
those that have ceased


operations but have fuel on site.2002-21Axial Outside-DiameterCracking Affecting Thermally
operations but have fuel on site.
 
2002-21          Axial Outside-Diameter                06/25/2002      All holders of operating licenses


===Treated Alloy 600 Steam===
Cracking Affecting Thermally                          for pressurized-water reactors
Generator Tubing06/25/2002All holders of operating licensesfor pressurized-water reactors


(PWRs), except those who have
Treated Alloy 600 Steam                                (PWRs), except those who have


permanently ceased operations
Generator Tubing                                      permanently ceased operations


and have certified that fuel has
and have certified that fuel has
Line 466: Line 663:
been permanently removed from
been permanently removed from


the reactor.2002-19Medical MisadministrationsCaused By Failure to Properly
the reactor.
 
2002-19          Medical Misadministrations            06/14/2002      All nuclear pharmacies and
 
Caused By Failure to Properly                         medical licensees.
 
Perform Tests on Dose


===Perform Tests on Dose===
Calibrators for Beta-and Low- Energy Photon-Emitting
Calibrators for Beta-and Low- Energy Photon-Emitting


Radionuclides06/14/2002All nuclear pharmacies andmedical licensees.2002-18Effect of Adding Gas IntoWater Storage Tanks on the
Radionuclides


===Net Positive Suction Head For===
2002-18          Effect of Adding Gas Into            06/06/2002      All holders of operating licenses
Pumps06/06/2002All holders of operating licensesfor nuclear power reactors, except those who have


permanently ceased operations
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
and have certified that fuel has
Line 482: Line 685:
been permanently removed from
been permanently removed from


the reactor.2002-17Medical Use of Strontium-90Eye Applicators: New
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.
 
Note:            NRC generic communications may be received in electronic format shortly after they are
 
issued by subscribing to the NRC listserver as follows:
                To subscribe send an e-mail to <listproc@nrc.gov >, no subject, and the following


===Requirements for Calibration===
command in the message portion:
and Decay Correction05/30/2002All U.S. Nuclear RegulatoryCommission medical licensees
                                    subscribe gc-nrr firstname lastname


that use strontium-90 (Sr-90) eye
______________________________________________________________________________________
OL = Operating License


applicators.Note:NRC generic communications may be received in electronic format shortly after they areissued by subscribing to the NRC listserver as follows:To subscribe send an e-mail to <listproc@nrc.gov >, no subject, and the followingcommand in the message portion:subscribe gc-nrr firstname lastname}}
CP = Construction Permit}}


{{Information notice-Nav}}
{{Information notice-Nav}}

Revision as of 05:16, 24 November 2019

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)
v • d • e


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

ML021820008

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

operability assessments.

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:

ADAMS

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.

Note: NRC generic communications may be received in electronic format shortly after they are

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:

subscribe gc-nrr firstname lastname

______________________________________________________________________________________

OL = Operating License

CP = Construction Permit


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