Information Notice 2010-11, Potential for Steam Voiding Causing Residual Heat Removal System Inoperability
| ML100640465 | |
| Person / Time | |
|---|---|
| Issue date: | 06/16/2010 |
| From: | Mcginty T, Tracy G Division of Construction Inspection and Operational Programs, Office of Nuclear Reactor Regulation |
| To: | |
| Beaulieu, D P, NRR/DPR, 415-3243 | |
| References | |
| IN-10-011 | |
| Download: ML100640465 (5) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
OFFICE OF NEW REACTORS
WASHINGTON, DC 20555-0001 June 16, 2010
NRC INFORMATION NOTICE 2010-11: POTENTIAL FOR STEAM VOIDING CAUSING
RESIDUAL HEAT REMOVAL SYSTEM
INOPERABILITY
ADDRESSEES
All holders of or applicants for an operating license or construction permit for a nuclear power
reactor issued under Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic
Licensing of Production and Utilization Facilities, except those that have permanently ceased
operations and have certified that fuel has been permanently removed from the reactor vessel.
All holders of or applicants for a standard design certification, standard design approval, manufacturing license, or combined license issued under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.
PURPOSE
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform
addressees of an issue at three pressurized-water reactor (PWR) plants where on multiple
occasions, their residual heat removal (RHR) systems were inoperable because of the potential
for steam voids at the RHR pump suction piping. Recipients should review the information for
applicability to their facilities and consider actions to avoid similar occurrences. The
suggestions contained in this IN are not NRC requirements, and no specific action or written
response is required.
DESCRIPTION OF CIRCUMSTANCES
In 2008 and 2009, the licensees at the Shearon Harris Nuclear Power Plant, Prairie Island
Nuclear Generating Plants, and Wolf Creek Generating Station discovered that their RHR
systems were potentially inoperable during shutdown periods because of elevated system
temperatures at the RHR pump suctions. The elevated system temperatures resulted from the
licensees lack of adequate procedures to ensure RHR system operability during all modes of
operation. At each of these plants, the fluid in the piping between the reactor coolant system
(RCS) hot leg to RHR system connection and the RHR minimum-flow line return connection
remained stagnant and at elevated temperatures following forced cooling as a result of
unrecognized system flow characteristics; namely, forced flow did not occur in that section of
pipe. Consequently, each licensee concluded incorrectly that the RHR system was properly
cooled, prior to shifting the RHR system to Emergency Core Cooling System (ECCS) injection
mode, when the system temperature was actually such that the affected RHR systems could
have incurred steam voiding if they had been used for emergency core cooling purposes.
Additional information is available on the Wolf Creek Generating Station in the NRC Focused
Baseline Inspection Report 05000482/2009006, dated August 12, 2009; the Wolf Creek
Licensee Event Report (LER) 50-482/2008-008-02, dated August 25, 2009; the Shearon Harris
LER 50-400/2009-002, dated December 15, 2009; and the Prairie Island LER 50-282/2009-004, dated June 5, 2009. These documents can be found on the NRCs public Web site under
Agencywide Documents Access and Management System (ADAMS) Accession
Nos. ML092240087, ML092450426, ML093580024, and ML091560611, respectively.
BACKGROUND
Gas accumulation in ECCS is an enduring issue associated with commercial nuclear power
plant operations. To address this problem the NRC issued Generic Letter 2008-01, Managing
Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray
Systems (ADAMS Accession No. ML072910759). During system reviews in response to the
generic letter, several PWR licensees discovered the potential for their RHR systems to become
inoperable during certain shutdown cooling evolutions.
When an RHR train is used for cooling the reactor coolant system, the temperature of the water
in that RHR train can reach 350 degrees Fahrenheit (F) (Mode 4 upper temperature limit). With
the temperature of the water in the RHR train as high as 350 degrees F, if its suction source is
switched from the RCS hot leg to the refueling water storage tank (e.g., during ECCS operation
in response to a loss-of-coolant accident) or to the containment sump (e.g., during extended
response to a loss-of-coolant accident), conditions at the suction of the RHR pump would result
in steam voiding since the temperature at the RHR pump suction would exceed the saturation
temperature. Steam voiding can result in binding of an RHR pump and the refueling water
storage tank discharge check valve, system flow interruptions, and water hammer; potentially
inhibiting the capability of the RHR system to fulfill its ECCS function. Some PWR plants
occasionally use multiple RHR trains to perform plant cooldowns. In such cases, multiple RHR
trains can become simultaneously inoperable for emergency core cooling.
The pressure and temperature at the suction of the RHR pump depends on the RHR system
lineup and the as-built system configuration. For example, when the RHR system is aligned for
shutdown cooling, the RHR pump suction pressure is the same as RCS pressure; during safety
injection and containment sump recirculation operations, the pressure at the suction of an RHR
pump is equal to the static head pressure created by the refueling water storage tank and the
containment sump, respectively. In all system lineups, the as-built configuration also
determines the head loss associated with different system configurations. The range of
possible pump suction pressures makes the RHR system susceptible to steam voiding and
water hammer during system lineup changes with suction temperatures above certain values.
Since the pressure and corresponding saturation temperature at the suction of an RHR pump
depends on RHR system design and as-built configuration, it is important that each licensee
ensure that its particular RHR operating procedures are tailored to their specific systems and
include parameters validated as plant-specific to ensure RHR systems required by Technical
Specifications remain operable.
DISCUSSION
Industry operating experience and guidance has shown that effective methods of maintaining
RHR system temperature within appropriate limits exist. For example, licensees can isolate the
RHR system, or a single train of the RHR system, from the RCS at a low enough temperature
so that the fluid at the RHR pump suction remains below the saturation temperature
corresponding to the pressure at the suction of a running RHR pump.
Another method of maintaining RHR system temperature is forced cooling through the
minimum-flow line. The main function of the RHR minimum-flow line is to allow RHR pump
operation during a safety injection signal when RCS pressure is still above the pumps shutoff
head. A typical minimum-flow line flow path takes suction from the discharge of the RHR pump, after flow has passed through the RHR heat exchanger, and returns flow to the suction of the
RHR pump. Neglecting minor conduction heat transfer, the amount of RHR piping that can be
cooled through the minimum-flow recirculation method is dependent on the location of the
minimum-flow return connection. Specifically, the RHR pipe upstream of this connection would
not be cooled due to there being little to no flow through this section of pipe.
Some licensees incorrectly assumed that using the minimum-flow recirculation method of
cooling their RHR system was sufficient when, in fact, the as-built configuration of the plant did
not allow for complete RHR system cooldown using this method. For example, approximately
140 feet of Wolf Creeks RHR system piping is not subject to minimum-flow recirculation
because it is located upstream of the minimum-flow return connection. The stagnant water
inside this 140 foot section of RHR pipe can only cool through ambient losses; therefore, it
remains at elevated temperatures for extended periods of time, possibly exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Nevertheless, station procedures allowed the RHR system to be realigned to the ECCS injection
mode of operation under these conditions and resulted in both trains of the RHR system being
inoperable during periods of operation in Modes 3 and 4. Sharon Harris and Prairie Island
plants discovered a similar condition occurred in their plants with the length of affected pipe
being the primary variable.
Other licensees may be susceptible to a similar issue in that their plants may contain piping runs
that are not able to be effectively cooled using the forced cooling method; yet their station
procedures may allow them to shift the RHR system to ECCS injection mode during elevated
system temperatures.
Licensees should consider this operating experience to ensure that their RHR system
procedures address their plant specific configurations and that they provide adequate methods
for satisfying the plants RHR system operability requirements as stated in the plants Technical
Specifications.
CONTACT
This IN requires no specific action or written response. Please direct any questions about this
matter to the technical contacts listed below or the appropriate Office of Nuclear Reactor
Regulation (NRR) project manager.
/RA/ /RA by MShuaibi for/
Timothy McGinty, Director Glenn Tracy, Director
Division of Policy and Rulemaking Division of Construction Inspection and
Office of Nuclear Reactor Regulation Operational Programs
Office of New Reactors
Technical Contacts: David Garmon, NRR
301-415-3512 E-mail: david.garmon-candelaria@nrc.gov
Warren Lyon, NRR
301-415-2897 E-mail: warren.lyon@nrc.gov
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under Electronic Reading Room/Document Collections.
CONTACT
This IN requires no specific action or written response. Please direct any questions about this
matter to the technical contacts listed below or the appropriate Office of Nuclear Reactor
Regulation (NRR) project manager.
/RA/ /RA by MShuaibi for/
Timothy McGinty, Director Glenn Tracy, Director
Division of Policy and Rulemaking Division of Construction Inspection and
Office of Nuclear Reactor Regulation Operational Programs
Office of New Reactors
Technical Contacts: David Garmon, NRR
301-415-3512 E-mail: David.Garmon-Candelaria@nrc.gov
Warren Lyon, NRR
301-415-2897 E-mail: Warren.Lyon@nrc.gov
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under Electronic Reading Room/Document Collections.
ADAMS Accession Number: ML100640465 TAC ME3411 OFFICE DIRS/IOEB Tech Editor BC/DIRS/IOEB DSS/SRXB BC/DSS/SRXB
NAME DGarmon KAzariah-Kribbs JThorp WLyon AUlses
DATE 5/13/10 04/05/10 e-mail 05/14/10 e-mail 05/13/10 5/14/10 e-mail
OFFICE D/DSS BC/NRO/SRSB DPR/PGCB DPR/PGCB BC/DPR/PGCB
NAME WRuland JDonoghue CHawes DBeaulieu SStuchell (Acting)
DATE 05/26/10 05/18/10 e-mail 6/10/10 05/27/10 6/11/2010
OFFICE D/NRO/DCIP D/DPR/PGCB
NAME GTracy (MShuaibi TMcGinty
for)
OFFICE 6/15/2010 6/16/2010
OFFICIAL RECORD COPY