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
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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 nonsafety-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 Commission’s 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 NRC’s 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

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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 NRC’s Special Inspection Team

concluded that the licensee’s process was narrowly focused on finding, understanding, and

correcting the cause of the trip. While the station’s 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 plant’s 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 licensee’s 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 NRC’s Special Inspection Team concluded that the licensee’s 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 NRC’s Special Inspection Team concluded that this process was not

integrated with the station’s 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 narrowrange 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

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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

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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 NRC’s 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 O’Keefe, 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
Attachment 1

IN 2002-10 Sup 1

Page 1 of 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

Page 1 of 3

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 lowlow 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 Canyon’s

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 licensee’s 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

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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 narrowrange 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 MidDeck 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

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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.