Information Notice 2001-13, Inadequate Standby Liquid Control System Relief Valve Margin
| ML012210146 | |
| Person / Time | |
|---|---|
| Issue date: | 08/10/2001 |
| From: | Marsh L Operational Experience and Non-Power Reactors Branch |
| To: | |
| References | |
| TAC MB1505 IN-01-013 | |
| Download: ML012210146 (5) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001 August 10, 2001 NRC INFORMATION NOTICE 2001-13: INADEQUATE STANDBY LIQUID CONTROL
SYSTEM RELIEF VALVE MARGIN
Addressees
All holders of operating licenses for boiling water reactors.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert
addresses to a recent staff finding regarding inadequate standby liquid control system relief
valve margin. 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 actions or
written response is required.
Background
The control rod drive (CRD) system provides the primary means to control reactivity, as
required by 10 CFR Part 50, Appendix A. In the original plant design of Susquehanna Units 1 and 2, the standby liquid control system (SLC) was provided as an independent and diverse
(from the CRD system) method for shutting down the reactor under conditions of normal
operation. Its specific function was to provide the capability to inject into the reactor a neutron- absorbing solution that was capable of achieving and maintaining subcriticality. At
Susquehanna Units 1 and 2, the system included two redundant pumps, each capable of
performing the design function.
In 1984 the NRC issued 10 CFR Section 50.62, "Requirements for reduction of risk from
anticipated transients without scram (ATWS) events for light water-cooled nuclear power
plants" (the ATWS rule). The ATWS rule added more stringent injection rate requirements for
the SLC system and required the ATWS functions to be performed under conditions of
anticipated operational occurrences. Specifically, the rule required that each boiling water
reactor have a SLC system with the capability of injecting into the reactor pressure vessel
(during anticipated operational occurrences) a borated water solution at a flow rate such that
the resulting reactivity control was at least equivalent to that resulting from the injection of 86 gallons per minute (gpm) of 13 weight percent sodium pentaborate decahydrate (boron)
solution.
Description of Circumstances
To comply with the ATWS rule, the Susquehanna licensee implemented a modification which
revised the SLC pump start logic to a simultaneous initiation of both pumps. This resulted in a
flow rate of at least 82.4 gpm and a corresponding required concentration of boron solution of
13.6 weight percent. This boron concentration became the licensing basis and was
subsequently included in the Susquehanna improved technical specifications.
The change to the pump start logic caused a significant increase in system pressure losses in
the pump discharge lines. These losses were the result of the increased fluid velocity in the
common injection line as the flow rate doubled from 41.2 gpm to 82.4 gpm. As a result of the
ATWS modification, the licensee determined that the maximum discharge pressure at the SLC
pumps was 1276 psig. This value was based on the lowest setpoint (1076 psig) of the main
steam safety relief valves (SRVs) in the pressure relief mode, the system friction losses for two- pump operation, and the elevation losses. Subsequently, in 1993, the licensee determined a
new maximum SLC pump discharge pressure of 1319 psig, based on a power uprate
modification. The change was due to a 30 psig increase to the SRV setpoint, and an increase
in calculated core flow. The licensee determined that the calculated value of 1319 psig was
acceptable because it maintained a 75 psig design margin requirement between the maximum
SLC pump discharge pressure and the minimum setting of the SLC pump discharge relief
valves (1400 psig).
During a recent design inspection at Susquehanna, the NRC found that the licensee's
assumption for reactor vessel pressure used in the maximum pump discharge pressure
calculation was non-conservative and disagreed with a vendor ATWS analysis for two of the
transients analyzed. Specifically, the inspection team found that for the main steam isolation
valve (MSIV) closure transient, the analysis indicated that, at the time of SLC system manual
initiation, the reactor vessel pressure would be as high as 1133 psig. Similarly, for the loss of
offsite power (LOOP) transient, the reactor pressure at various times in the event was a
nominal 1200 psig. The much higher pressure calculated for the LOOP transient event was
due to the loss of power to the containment instrument gas compressors and the resulting loss
of gas required to open the SRVs. Although each SRV was equipped with a gas accumulator, the amount of gas available in each accumulator was sufficient for only a few SRV actuations.
Therefore, the SRV would eventually lift on its higher spring setting (safety mode) and not in its
normal pressure relief mode.
Based on the above, the inspection team concluded that the maximum reactor vessel pressure
of 1106 psig assumed by the licensee in the design calculations of record was non- conservative. The increases in main steam SRV lift pressure setpoints through the years due
to valve simmering concerns and power uprate considerations contributed to the loss of
adequate margin between maximum expected pump discharge pressures and the system relief
valve settings. This resulted in the likelihood that the SLC pump discharge relief valves would
lift during at least one of the ATWS transient scenarios, the loss of offsite power. The lifting of
the SLC pump discharge relief valves would cause the sodium pentaborate solution to be
recycled to the pump suction and, therefore, prevent the system from meeting the equivalent
flow capacity required by the ATWS rule. Discussion
The licensee modified the Susquehanna Unit 2 SLC system, during a recent refueling outage.
The modification increased the flange pressure rating of both pumps (from 1400 psig to 1500
psig) and raised the lift pressure of the pump discharge relief valves to 1500 psig. The licensee
intends to perform the same modification on the Unit 1 system. Additional details regarding the
issue identified during the inspection can be found in Inspection Report 05000387/01-
004;05000388/01-004, Accession # ML011420068.
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/ Patrick M. Madden FOR
Ledyard B. Marsh, Chief
Operational Experience
and Non-Power Reactors Branch
Division of Regulatory Improvement Program
Office of Nuclear Reactor Regulation
Technical contacts: Frank Arner, DRS John Richmond, DRP
(610) 337-5194 (570) 542-2134 E-mail: fja@drs.gov E-mail: jer4@drp.gov
Neil Della Greca, DRS
(610) 337-5046 E-mail: ald1@drs.gov
Attachment: List of Recently Issued NRC Information Notices
- Publicly Available G Non-Publicly Available G Sensitive : Non-Sensitive
OFFICE RGN I Tech Editor RGN I RGN I RGN I REXB
NAME FArner PKleene/NF for JRichmond* NDellaGreca LDoerflein* NFields*
DATE 08/01/2001 05/14/2001 07/31/2001 7/31/2001 08/01/2001 08/01/2001 C:SRXB REXB C:REXB
JWermiel* JTappert LMarsh
08/03/2001 8/10/2001 8/10/2001
Attachment LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
_____________________________________________________________________________________
Information Date of
Notice No. Subject Issuance Issued to
______________________________________________________________________________________
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
2001-08 Update on the Investigation of 06/06/01 All Medical Licensees
Supplement 1 Patient Deaths in Panama, Following Radiation Therapy
2001-08 Treatment Planning System 06/01/01 All medical licensees
Errors Result in Deaths of
Overseas Radiation Therapy
Patients
______________________________________________________________________________________
OL = Operating License
CP = Construction Permit