View source for Information Notice 2001-15, Non-Conservative Error Minimum Critical Power Ratio Limits

{{Adams
| number = ML012710347
| issue date = 10/29/2001
| title = Non-Conservative Error Minimum Critical Power Ratio Limits
| author name = Imbro E
| author affiliation = NRC/NRR/DRIP/REXB
| addressee name = 
| addressee affiliation = 
| docket = 
| license number = 
| contact person = Hodge,   CV, NRR/DRIP/REXB, (415-1861)
| case reference number = TAC MB02454
| document report number = IN-01-015
| document type = NRC Information Notice
| page count = 5
}}
{{#Wiki_filter:UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001 October 29, 2001 NRC INFORMATION NOTICE 2001-15:                NON-CONSERVATIVE ERRORS IN MINIMUM

CRITICAL POWER RATIO LIMITS

==Addressees==
All holders of operating licenses or construction permits for boiling water reactors (BWRs).

==Purpose==
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to inform the

addressees that NRC recently received reports of nonconservative errors affecting the

minimum critical power ratio (MCPR) operating and safety limits. It is expected that recipients

will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice are not

NRC requirements; therefore, no specific action or written response is required.

Background

Thermal limits on nuclear power plant operation are established to ensure that fuel cladding

integrity is not lost because of overheating. Fuel cladding integrity provides one of the

protections of public health and safety against uncontrolled release of radioactivity into the

environment. One of these limits, the MCPR safety limit, is defined as the smallest allowed

ratio of the critical bundle power (defined as that required to produce a critical stage of boiling

called transition boiling somewhere in the bundle) to the actual bundle power. A higher value of

this ratio may be used as an operating limit based on the single failure (operator error or

equipment malfunction causing a transient) design basis requirement.

==Description of Circumstances==
In June 2001, the fuel vendor General Electric reported that licensees that implemented stability

detect and suppress trip systems at their plants may be making nonconservative errors in their

licensing calculations for reloads, resulting in inadequate MCPR safety limit protection (Part 21 Report 2001-23-0).

Optional stability solutions requiring these calculations are defined as Options I-D, II, and III in

the vendors document NEDO-32465-A, Reactor Stability Detect and Suppress Solutions

Licensing Basis Methodology for Reload Applications, August 1996. This document gives two

generic so-called Delta CPR/Initial CPR Vs. Oscillation Magnitude (DIVOM) curves, one for

core-wide mode oscillations and one for regional mode oscillations. The curves relate

normalized critical power ratio to hot bundle oscillation magnitude.

In Option I-D, the generic core-wide curve is used to confirm that the flow-biased average

power range monitor (APRM) flux trip provides adequate MCPR safety limit protection for a

core-wide mode oscillation initiating on the rated flow control line.

Option II is not specifically addressed in the vendors document, but the vendor states that the

generic regional mode curve has been used at Nine Mile Point Unit 1 to confirm that the APRM

trip gives adequate MCPR protection for a regional mode oscillation initiating on the rated flow

control line.

In Option III, the generic regional mode oscillation curve is used to determine the setpoints for

the implemented stability detect and suppress trip system used to provide adequate MCPR

protection. Plants using this option sometimes call these systems oscillation power range

monitors.

In recent evaluations, the vendor identified a nonconservative error in the high peak bundle

power-to-flow ratios in the generic regional DIVOM curve and in the high core-average power- to-flow ratios in the generic core-wide DIVOM curve. The generic regional DIVOM curve

overpredicts the Option III system trip setpoint, and the generic core-wide DIVOM curve

overpredicts the flow-biased APRM flux trip setpoint.

The vendor stated that it had informed all affected nuclear power plants, including General

Electric boiling water reactors (BWRs), and the industry BWR owners group. The vendor

described compensatory actions taken by the Hatch licensee for both Units 1 and 2. The Perry, Nine Mile Point Unit 2, and Fermi Unit 2 licensees notified the NRC of their compensatory

actions (Event Notifications 38099, 38106, and 38119 respectively).

The vendor provided updated generic DIVOM curves and a corresponding figure of merit to

determine curve applicability in August 2001. The vendor expected to review applicable studies

and, if necessary, develop new methodologies and obtain licensing approval for them by the

end of 2002.

In March 2001, the fuel vendor Framatome ANP Richland reported having used an

inappropriate reference temperature in establishing MCPR operating limits for several BWRs.

The result was to overpredict the thermal conductivity of the fuel. The affected plants were

Dresden Units 2 and 3, Quad Cities Units 1 and 2, LaSalle Units 1 and 2, and Susquehanna

Units 1 and 2. The largest correction was to increase the previous MCPR limit by 0.02 for

power less than 60-percent rated power (Part 21 Report 2001-14-0).

The vendor stated that it had demonstrated continued compliance with the overpressurization

criteria of the American Society of Mechanical Engineers (ASME) code and had installed the

appropriate reference temperature in the computer code used for the MCPR analysis. In October 2000, Siemens Power, the predecessor of Framatome ANP Richland, reported a

deviation in the pump model of the COTRANSA2 computer code, resulting in improper

calculation of the flow resistance during a pump seizure event. The affected nuclear power

plants were Dresden Units 2 and 3 and Susquehanna Units 1 and 2 (Part 21 Report 2000-23-
0).

In October 2000, the Hope Creek licensee was doing surveillance testing of a recirculation

pump motor-generator set. The licensee reported finding that the scoop tube mechanical and

electrical stop overspeed setpoints were nonconservatively high (less than or equal to 109 percent for the mechanical stop and less than or equal to 107 percent for the electrical stop). If

the analyzed reactor recirculation pump runaway transient occurs, the flow-dependent MCPR

may be exceeded (Event Notification 37418).

In September 2000, the Grand Gulf licensee reported that a main generator partial load

rejection can actuate a control circuit that may not always activate a reactor scram or

recirculation pump downshift as assumed in the analysis. This condition could adversely affect

MCPR limits (Event Notification 37342).

In July 2000, the Fitzpatrick licensee reported finding the mechanical stops to limit speed of

both recirculation pumps were set at values exceeding those assumed in the calculated

flow-dependent MCPR in the core operating limit report. Assuming initial low flow, this error

could result in exceeding the MCPR safety limit under postulated runaway recirculation flow

(Event Notification 37196).

In January 1999, the fuel vendor Asea Brown Boveri-Combustion Engineering (ABB-CE)
reported that BWR MCPR analyses using the BISON fast-transient analysis code could be

nonconservative because the code incorrectly models the reactor vessel lower plenum volume

(Part 21 Report 1999-07-0).

Also in January 1999, ABB-CE reported a potential nonconservative critical power ratio

correlation (XL-S96) for SVEA-96 BWR fuel. This correlation is based on a cosine axial power

shape, which could lead to nonconservative MCPR operating limits and nonconservatively

monitored critical power ratios for top-peaked axial power shapes. This error affected WNP-2 for Cycle 14 after 5,600 MWD/MTU fuel exposure (Part 21 Report 1999-08-0).

In November 1998, ABB-CE reported that a defect in modeling radial power in the BWR fast- transient analysis code BISON-SLAVE could lead to nonconservative MCPR operating limits.

The defect involves an input option that was incorrectly described in the code user manual and

not the program coding itself. The affected nuclear power plants were WNP-2 and

Susquehanna Unit 1 (Part 21 Report 1998-69-0).

In September 1998, the fuel vendor Siemens Power reported that BWR MCPR operating limits

were nonconservative because in determining gap heat transfer coefficients used by the

RODEX2 code for BWR transient analyses, fuel cladding was assumed to be cold-worked

instead of annealed. The affected nuclear power plants were LaSalle Unit 2 for Cycle 8 and

Quad Cities Unit 2 for Cycle 15 (Part 21 Report 1998-55-0). In April 1998, the fuel vendor Siemens Power reported that additive constants used by the core

monitoring system to determine MCPR for Atrium-9B fuel were nonconservative and more

uncertain than previously estimated. The affected nuclear power plants were Dresden Unit 3 and Quad Cities Unit 2 (Part 21 Report 1998-33-0).

Discussion

All of these 11 events involved BWRs. They may be categorized as follows:
*        6  thermal modeling errors

*        2  other modeling errors

*        2  incorrect settings

*        1  control circuit modeling error

Each category represents a shortcoming in quality control measures. Individually, the events

had little safety significance. Collectively, a significant number of similar events in a relatively

short time may indicate a trend of deteriorating quality control of the MCPR safety parameter.

This information notice requires no specific action or written response. If you have any

questions about the information in this notice, please contact one of the technical contacts

listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.

/RA/
                                                 Eugene V. Imbro, Acting Chief

Operational Experience

and Non-Power Reactors Branch

Division of Regulatory Improvement Programs

Office of Nuclear Reactor Regulation

Contacts:     V. Hodge, NRR                            Tai L. Huang, NRR

301-415-1861                            301-415-2867 E-mail: cvh@nrc.gov                     E-mail: tlh1@nrc.gov

Attachment: List of Recently Issued Information Notices

ML012710347                                  Template #:NRR-052
      : Publicly Available          G Non-Publicly Available         G Sensitive     G Non-Sensitive

OFFICE REXB                TECH EDITOR SRXB                   SRXB             REXB                  C:REXB

NAME      CVHodge*         cvh for     *    THuang*           RCaruso*         JTappert              EVImbro

DATE      09/19/2001 09/19/2001             09/24/2001        09/24/2001          / /2001               / /2001

Attachment LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

_____________________________________________________________________________________
Information                                         Date of

Notice No.              Subject                     Issuance   Issued to

______________________________________________________________________________________
2001-14           Problems with incorrectly-       10/03/01     All holders of operating licenses

Installed Swing-Check Valves                  for nuclear power reactors, except those who have ceased

operations and have certified that

fuel has been permanently

removed from the reactor vessel

2001-13           Inadequate Standby Liquid        08/10/01     All holders of operating licenses

Control System Relief Valve                   for boiling water reactors

Margin

2001-12           Hydrogen Fire at Nuclear         8/08/01      All holders of operating licenses

(ERRATA)          Power Stations                                or construction permits for

nuclear power reactors except

those who have ceased

operations and have certified that

fuel has been permanently

removed from the reactor vessel

2001-12           Hydrogen Fire at Nuclear         7/13/01      All holders of operating licenses

Power Stations                                or construction permits for

nuclear power reactors except

those who have ceased

operations and have certified that

fuel has been permanently

removed from the reactor vessel

2001-11           Thefts of Portable Gauges        07/13/01     All portable gauge licensees

2001-10           Failure of Central Sprinkler     06/28/01     All holders of licenses for nuclear

Company Model GB Series                       power, research, and test

Fire Sprinkler Heads                          reactors and fuel cycle facilities

2001-09           Main Feedwater System            06/12/01    All holders of operating licenses

Degradation in Safety-Related                for pressurized water nuclear

ASME Code Class 2 Piping                     power reactors, except those who

Inside the Containment of a                  have permanently ceased

Pressurized Water Reactor                    operations and have certified that

fuel has been permanently

removed from the reactor vessel

______________________________________________________________________________________
OL = Operating License

CP = Construction Permit}}

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