United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 97-27: Effect of Incorrect Strainer Pressure Drop on Available Net Positive Suction Head

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR REACTOR REGULATION
                         WASHINGTON, D.C.  20555-0001

                                 May 16, 1997


NRC INFORMATION NOTICE 97-27:  EFFECT OF INCORRECT STRAINER PRESSURE DROP ON   
                               AVAILABLE NET POSITIVE SUCTION HEAD

Addressees

All holders of operating licenses or construction permits for light-water
power reactors, except those licensees who have permanently ceased operations
and have certified that fuel has been permanently removed from the reactor
vessel.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees that two licensees of boiling-water reactors (BWR)
have recently identified inaccurate assumptions in licensing-basis calcula-
tions for net positive suction head (NPSH).  One of the licensees has decided
to immediately shut down the reactor to replace suction strainers in the
emergency core cooling system (ECCS) before the next refueling outage.  It is
expected that recipients will review the information for applicability to
their facilities and consider actions, as appropriate, to detect or avoid
similar problems.  However, suggestions contained in this information notice
are not NRC requirements; therefore, no specific action or written response is
required.

Description of Circumstances

Monticello

On April 15, 1997, Northern States Power, the licensee for the Monticello
Nuclear Power Plant, notified the NRC staff that the NPSH available to the
core spray pumps may not meet the required NPSH under all accident conditions. 
During a review of the ECCS pump NPSH requirements, the licensee calculated a
new higher head loss, approximately 3.6 meters (11.7 feet) per 630 L/s (10,000
gpm) rather than 0.3 meter (1 foot) per 630 L/s (10,000 gpm), for clean ECCS
suction strainers.  The specific scenario of concern involved a failure of the
low-pressure coolant-injection loop select logic to select the intact reactor
recirculation loop.  As a result, the licensee determined that the core spray
pumps may not have adequate NPSH available during the first 10 minutes
following a design-basis loss-of-coolant accident (LOCA).


9705150337.                                                            IN 97-27
                                                            May 16, 1997
                                                            Page 2 of 3


Dresden Units 2 and 3

On December 20, 1996, Commonwealth Edison, the licensee for the Dresden
Nuclear Power Station Units 2 and 3, notified the NRC staff that the ECCS may
be susceptible to NPSH problems since the suction strainer design was based on
an incorrect head loss value.  While conducting a plant-specific analysis in
support of its response to NRC Bulletin 96-03, the licensee discovered that
the new value for the head loss across clean suction strainers was 1.8 meters
(5.8 feet) per 630 L/s (10,000 gpm) versus 0.3 meter (1 foot) per 630 L/s
(10,000 gpm) as described in the Updated Final Safety Analysis Report.  With
the calculated head loss, no credit taken for containment overpressure, and
the accident conditions described above, the licensee determined that the ECCS
may fail to operate as intended.

Discussion

During its review of the Monticello operability evaluation, the NRC staff
questioned whether the licensee would be able to reflood the reactor core
following a LOCA and prevent exceeding the 1204 �C (2200 �F) peak cladding
temperature limit as required by 10 CFR 50.46.  Monticello has large
quantities of fibrous insulation in its drywell.  The Monticello ECCS suction
strainers have a very small surface area and very high approach velocities. 
Even a small quantity of insulation reaching the suction strainers, which are
located in the torus, can cause sufficient head losses across the strainer to
cause a loss of NPSH.  This scenario is outside the original licensing basis
of the plant, but it could occur, given the current knowledge available.

On May 9, 1997, the licensee for Monticello decided to shut down the reactor
and replace the ECCS suction strainers.  The existing strainer on each of the
four suction lines is a truncated cone design about 43 cm (17 inches) in
diameter and 25 cm (10 inches) long.  Each strainer will be replaced with two
larger strainers, each of which is about 102 cm (40 inches) in diameter and
213 cm (84 inches) long.  The plant will remain shut down until the strainers
can be procured and installed, currently estimated to be 3 to 4 months.

The Dresden units have large quantities of reflective metallic insulation
(RMI) in their drywell.  The Dresden Units 2 and 3 strainers are also very
small with very high approach velocities.  However, the head loss associated
with debris beds composed mostly of RMI is typically less severe than head
loss associated with debris beds composed of fibrous materials.  The existing
strainers are being replaced with larger strainers at Dresden Unit 3 during
the current refueling outage and the licensee plans to replace them at Unit 2
during the next outage.  As an interim corrective action the licensee
submitted on January 13, 1997, an emergency technical specification amendment
requesting that the staff evaluate an Unreviewed Safety Question (USQ)
associated with the operation of Dresden Units 2 and 3 with the existing
strainers.  The licensee's submittal sought staff approval of operation of
both units with the increased head loss across the clean ECCS strainers;
revised Technical Specifications (TS) values for a lower allowable water
temperature in the suppression chamber and the ultimate heat sink; and a
revised TS bases that states credit was taken for 13.8 kilopascals (2 psig) of
containment pressure (this compensates for a slight increase in the amount of
NPSH deficiency during the first 10 minutes following a LOCA).  The NRC staff
issued the amendments on January 28, 1997..                                                            IN 97-27
                                                            May 16, 1997
                                                            Page 3 of 3 


The staff notes that continued operation by BWR licensees during the
development of the resolution of the BWR ECCS strainer clogging issue is based
on the premise that licensees will be able to reflood the reactor core
immediately following a design-basis LOCA.  In response to NRC Bulletin 93-02,
Supplement 1, BWR licensees implemented interim measures to ensure that they
could mitigate a design-basis LOCA should the ECCS strainers clog.  The
acceptability of the licensees' interim measures depended upon (1) adequate
time for operators to respond to clogged strainers to align alternate water
sources (both safety-related and nonsafety-related sources), (2) emergency
operating procedures (EOPs) which provide adequate guidance on mitigating a
strainer clogging event, (3) operator training to mitigate a strainer clogging
event, (4) recent cleaning of suppression pools, and (5) the removal of loose
and temporary fibrous materials stored in the containment.

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.


                                          signed by S.H. Weiss for

                                       Marylee M. Slosson, Acting Director
                                       Division of Reactor Program Management
                                       Office of Nuclear Reactor Regulation

Technical contacts:  Robert Elliott, NRR        
                     301-415-1397         
                     E-mail:  rbe@nrc.gov 

                     Kerri Kavanagh, NRR
                     301-415-3743
                     E-mail:  kak@nrc.gov
Page Last Reviewed/Updated Tuesday, December 03, 2013