Information Notice 1994-60, Potential Overpressurization of Main Steam System
| ML033090523 | |
| Person / Time | |
|---|---|
| Issue date: | 08/22/1984 |
| From: | Grimes B Office of Nuclear Reactor Regulation |
| To: | |
| Fields E N 301-415-1173 | |
| References | |
| IN-94-060 | |
| Download: ML033090523 (12) | |
-
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C.
20555-0001
August 22, 1994
POTENTIAL OVERPRESSURIZATION OF
MAIN STEAM SYSTEM
Addressees
All holders of operating licenses or construction permits for pressurized- water reactors (PWRs).
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to a potential for overpressurizing the main steam
system during periods when one or more main steam safety valves are
inoperable. It is expected that recipients will review the information for
applicability to their facilities and consider actions, as appropriate, to
avoid such problems.
However, suggestions contained in this information
notice are not NRC requirements; therefore, no specific action or written
response is required.
Description of Circumstances
Westinghouse Nuclear Safety Advisory Letter NSAL-94-001, "Operation at Reduced
Power Levels With Inoperable MSSVs," January 20, 1994 (Enclosure 1) describes
a deficiency in the basis for STS Table 3.7.1, "Operable Main Steam Safety
Valves Versus Applicable Power in Percent of Rated Power." The basis assumes
that the maximum allowable initial power level is a linear function of main
steam safety valve capacity. Westinghouse has determined that this assumption
is -not correct and notes that, when operating at low power in accordance with
Standard Technical Specification Table 3.7.1, with one or more safety valves
inoperable, a loss-of-load/turbine trip transient concurrent with a loss of
feedwater could result in overpressurization of the main steam system.
Should a plant operating at full power experience such an initiating event, the transient would be terminated by a reactor trip. The reactor would trip
on either high pressurizer pressure, overtemperature delta temperature, or
steam generator low-low level.
Secondary-side overpressure protection is
provided by the actuation of the main steam safety valves. When a plant is
operating at a reduced power level, a reactor trip may not be initiated early
9408170105
. .
..-
August 22, 1994 in the transient. This results in a longer time during which primary heat is
transferred to the secondary side.
The reactor eventually trips on low-low
steam generator water level.
However, if this event occurred while the unit
is operating in accordance with Standard Technical Specification Table 3.7.1, with one or more inoperable safety valves, the trip may not occur before
secondary pressure exceeds 110 percent of the design pressure.
The equation used in the original Westinghouse bases for Table 3.7.1 reduces
the power range neutron flux trip setpoint linearly with the assumed reduction
in safety valve capacity. The same equation is used in Babcock & Wilcox and
Combustion Engineering plant designs. Therefore the potential for
overpressurization of the main steam system also exists for these plants.
NSAL 94-001 recommends the use of a more conservative equation to calculate
the power range high neutron flux trip setpoint.
However, the inclusion of
NSAL 94-001 with this information notice should not be construed as an NRC
endorsement of the revised equation as the only means to address this problem.
Operation at appropriate power levels when one or more main steam safety
valves are inoperable can avoid exceeding 110 percent of design pressure as a
result of assumed initiating events. With regard to inadequacies of existing
plant technical specification bases, the NRC staff is considering the need for
generic resolution of this issue. In the interim, the staff does not expect
licensees that implement administrative controls also to propose changes to
plant technical specifications.
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.
Brian K. Grimes, Director
r
Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical contacts: N. Fields, NRR
(301) 504-1173
C. Liang, NRR
(301) 504-2878
T. Dunning, NRR
(301) 504-1189 Attachments:
1. Westinghouse Nuclear Safety Advisory Letter 94-001
2. List of Recently Issued NRC Information Notices
-~
Attachment 1
August 22, 1994 NUCLEAR SAFETY ADVISORY
i
Westinghouse
Energy
LETTER
Systems
Business
Unit
THIS IS A NOTIFICATION OF A RECENTLY IDENTIFIED POTENTIAL SAFETY ISSUE PERTAINING TO BASIC
COMPONENTS SUPPLIED BY WESTINGHOUSE. THIS INFORMATION IS BEING PROVIDED TO YOU SO THAT A
REVIEW OF THIS ISSUE CAN BE CONDUCTED BY YOU TO DETERMINE IF ANY ACTION IS REQUIRED.
P. 0. Box 355, Pittsburgh, PA 15230-0355 Subject: Operation at Reduced Power Levels with Inoperable MSSVs
Number: NSAL-94-001 Basic Component: Loss of Load/Turbine Trip Analysis for Plant Licensing Basis
Date: 01/20/94 Plants: See Enclosed List
Substantial Safety Hazard or Failure to Comply Pursuant to 10 CFR 21.21(a)
Yes
No
Transfer of Information Pursuant to 10 CFR 21.21(b)
Yes 3 Advisory Information Pursuant to 10 CFR 21.21(c)(2)
Yes 0
SUMMARY
Westinghouse has identified a potential safety issue regarding plant operation within Technical Specification Table 3.7-1. This
issue does not represent a substantial safety hazard for your plant pursuant to 10 CFR 21. However, this issue does represent a
condition which may impact your plant's licensing basis.
Table 3.7-1 allows plants to operate with a reduced number of operable MSSVs at a reduced power level, as determined by the
high neutron flux trip setpoint. The FSAR loss of load/turbine trip (LOL/rT) analysis from full power bounds the case where
all MSSVs are operable. The FSAR (LOLLTT event may not be bounding for the allowable operating configurations of Table
3.7-1 since the high neutron flux trip setpoint, which is identified in Table 3.7-1 for a corresponding number of inoperable
MSSVs, may not be low enough to preclude a secondary side overpressurization condition. As a result, the basis for Table 3.7-
1 may not be sufficient to preclude overpressurization of the secondary side of the steam generator.
Therefore, it is recommended that you review the enclosed information to determine the applicability of this issue to your plant.
The enclosed information contains a more detailed description of the issue and identifies solutions that you may wish to pursue
to address this issue. These solutions include, but are not limited to, a re-evaluation of the LOL/TT analysis and/or a change to
Technical Specification Table 3.7-1.
Additional information, if required, may be obtained from the originator. Telephone 412-374-6460.
Originator: *
J. W. Fasnacht
Strategic Licensing Issues
H. A. Sepp, Manager
Strategic Licensing Issues
- Electronic copy of signed original provided to NRC courtesy of Westinghouse.
Attachment 1
August 22, 1994 Plants Affected
D. C. Cook 1 & 2 J. M. Farley 1 & 2 Byron 1 & 2 Braidwood I & 2
V. C. Summer 1
Zion 1 & 2
Shearon Harris 1
W. B. McGuire I & 2 Catawba 1 & 2 Beaver Valley 1 & 2 Turkey Point 3 & 4 Vogtle I & 2 Indian Point 2 & 3 Seabrook 1 Millstone 3 Diablo Canyon 1 & 2 Wolf Creek
Callaway 1 Comanche Peak 1 & 2 South Texas 1 & 2 Sequoyah 1 & 2 North Anna 1 & 2 Watts Bar 1 & 2 Sizewell B
Kori 1, 2, 3 & 4 Yonggwang 1 & 2 Salem 1 & 2
Attachment 1
August 22, 1994 Issue Description
Westinghouse has identified a deficiency in the basis for Technical
Specification 3.7.1.1.
This Technical Specification allows the plant to
operate at a reduced power level with a reduced number of operable Main Steam
Safety Valves (MSSVs). The deficiency is in the assumption that the maximum
allowable initial power level is a linear function of the available MSSV
relief capacity. The linear function is identified in the Bases Section for
Technical Specification 3/4.7.1.1 and is provided as follows:
SP -
(X) - (Y)(V) X (109)
x
=
Reduced reactor trip setpoint in % of RATED THERMAL POWER
V
=
Maximum number of inoperable safety valves per steam line
X
=
Total relieving capacity of all safety valves per steam
line in lbm/hour
Y
=
Maximum relieving capacity of any one safety valve in
1bm/hour
(109)
=
Power range neutron flux-high trip setpoint for all loops
in operation
Under certain conditions and with typical safety analysis assumptions, a Loss
of Load/Turbine Trip transient from part-power conditions may result in
overpressurization of the main steam system when operating in accordance with
this Technical Specification.
The following discussion describes the issue in
more detail and provides recommended alternatives for addressing the issue.
Technical Evaluation
The Loss of Load/Turbine Trip (LOL/TT) event is analyzed in the FSAR to show
that core protection margins are maintained (DNBR), the RCS will not
overpressurize, and the main steam system will not overpressurize.
The
Attachment 1
August 22, 1994 analysis assumes an immediate loss of steam relieving capability through the
turbine and coincident loss of all main feedwater.
No credit is taken for the
direct reactor trip on turbine trip, since this trip would not be actuated for
the case of a loss of steam load. Rather, the transient is terminated by a
reactor trip on high pressurizer pressure, overtemperature AT, or low steam
generator water level.
Secondary side overpressure protection is provided by
actuation of the Main Steam Safety Valves (MSSVs), which are designed to
relieve at least full power nominal steam flow.
The analysis verifies that
the MSSV capacity is sufficient to prevent secondary side pressure from
exceeding 110 percent of the design pressure.
The FSAR only analyzes the LOL/TT transient from the full power initial
condition, with cases examining the effects of assuming primary side pressure
control and different reactivity feedback conditions. With fully operational
MSSVs, it can be demonstrated that overpressure protection is provided for all
initial power levels. However, for most plants, Technical Specification 3.7.1.1 allows operation with a reduced number of operable MSSVs at a reduced
power level as determined by resetting the power range high neutron flux
setpoint. This Technical Specification is not based on a detailed analysis, but rather on the assumption that the maximum allowable initial power level is
a linear function of the available MSSV relief capacity. Recently, it has
been determined that this assumption is not valid.
The problem is that if main feedwater is lost, a reactor trip is necessary to
prevent secondary side overpressurization for all postulated core conditions.
At high initial power levels a reactor trip is actuated early in the transient
as a result of either high pressurizer pressure or overtemperature T. The
reactor trip terminates the transient and the MSSVs maintain steam pressure
below 110% of the design value.
At lower initial power levels a reactor trip may not be actuated early in the
An overtemperature AT trip isn't generated since the core thermal
margins are increased at lower power levels. A high pressurizer pressure trip
isn't generated if the primary pressure control systems function normally.
This results in a longer time during which primary heat is transferred to the
secondary side.
The reactor eventually trips on low steam generator water
level, but this may not occur before steam pressure exceeds 110% of the design
Attachment 1
August 22, 1994 value if one or more MSSVs are inoperable in accordance with the Technical
Specification.
Due to the wide variety of plant design features that are important to the
LOL/TT analysis, it is difficult to perform a generic evaluation to show that
the issue does not apply to certain plants. The following key parameters have
a significant effect on the secondary side pressure transient:
'
MSSV relief capacity
>
Moderator Temperature Coefficient (MTC)
Margin between the MSSV set pressures (including tolerance) and the
overpressure limit
>
Low-low steam generator water level reactor trip setpoint
Safety Significance
The Technical Specifications for most plants allow operation at a reduced
power level with inoperable MSSVs.
From a licensing basis perspective, this
condition may result in secondary side overpressurization in the event of a
LOL/TT transient. The licensing basis for anticipated operational occurrences
(ANS Condition II events) typically requires that the secondary side
pressure remain below 110% of the design value.
Westinghouse has determined that this issue does not represent a substantial
safety hazard. There are several mitigating factors which provide assurance
that there is no loss of safety function to the extent that there is a major
reduction in the degree of protection provided to the public health and
safety. These include, but are not limited to, the following:
1. Adequate overpressure protection is provided at all power levels if
all of the MSSVs are operable.
2. If a reactor trip does not occur but main feedwater flow is
maintained, operation in accordance with the Technical Specification
Table 3.7-1 will not result in an overpressure condition.
3.
In any LOL/TT transient, the atmospheric steam dump valves and/or
condenser steam dump valves actuate to relieve energy from the steam
Attachment 1
August 22, 1994 generators prior to the opening of the MSSVs, and continue to
relieve steam if the MSSVs do open. Since it is not a safety-grade
function, steam dump is not assumed to operate in the safety
analysis; however, in reality it is the first line of defense in
protecting the secondary system against overpressurization. It is
very improbable that all these components would be inoperable
coincident with inoperable MSSVs.
4. Even near the beginning of core life with a positive or zero MTC,
the primary coolant heatup resulting from the transient would tend
to drive the MTC negative, which would reduce the core power and
heat input to the coolant. This would result in a lower required
MSSV capacity to prevent secondary overpressurization.
The safety
analysis does not credit the reduction of MTC during the transient.
NRC Awareness / Reportability
Westinghouse has not notified the NRC of this issue, based upon the
determination that it does not represent a substantial safety hazard pursuant
to 10 CFR 21.
However, Westinghouse will send a copy of this letter to the
NRC since this issue impacts information contained in NUREG-1431, "Standard
Technical Specifications, Westinghouse Plants".
Recommendations
To address this issue, the following actions may be considered:
(1) Modify Technical Specification 3.7.1.1 (or equivalent) and the associated
basis such that the maximum power level allowed for operation with
inoperable MSSVs is below the heat removing capability of the operable
MSSVs. A conservative way to do this is to set the power range high
neutron flux setpoint to this power level, thus ensuring that the actual
power level cannot exceed this value. To calculate this setpoint, the
governing equation is the relationship q = m h, where q is the heat
input from the primary side, m is the steam flow rate and Ah is the heat
of vaporization at the steam relief pressure (assuming no subcooled
Thus, an algorithm for use in defining the revised Technical
Specification table setpoint values would be:
Attachment 1
August 22, 1994 (wshfgN)
Hi
= (100/Q)
K
where:
Hi
=
Safety Analysis power range high neutron flux setpoint, percent
Q
Nominal NSSS power rating of the plant (including reactor
coolant pump heat), Mwt
K
=
Conversion factor, 947.82 (Btu/sec)
Mwt
Ws
=
Minimum total steam flow rate capability of the operable
MSSVs on any one steam generator at the highest MSSV
opening pressure including tolerance and accumulation, as
appropriate, in lb/sec.
For example, if the maximum
number of inoperable MSSVs on any one steam generator is
one, then w should be a summation of the capacity of the
operable MSSVs at the highest operable MSSV operating
pressure, excluding the highest capacity MSSV. If the
maximum number of inoperable MSSVs per steam generator is
three then w should be a summation of the capacity of the
operable MSSVs at the highest operable MSSV operating
pressure, excluding the three highest capacity MSSVs.
hfg
=
heat of vaporization for steam at the highest MSSV
opening pressure including tolerance and accumulation, as
appropriate, Btu/lbm
N
=
Number of loops in plant
Attachment I
August 22, 1994 The values calculated from this algorithm must then be adjusted lower for
use in Technical Specification 3.7.1.1 to account for instrument and
channel uncertainties (typically 9% power).
The maximum plant operating
power level would then be lower than the reactor protection system
setpoint by an appropriate operating margin.
It should be noted that the use of this equation will resolve the issue
identified in this letter by enabling you to re-calculate your Technical
Specification 3.7.1.1 setpoints without further modifications to the
structure of the Technical Specification. The re-calculated setpoints
are likely to be lower than those currently allowed by the Technical
Specification.
However, you should be aware of at least two
conservatisms with the equation. You may wish to review these
conservatisms to evaluate the use of the equation relative to your plant
specific operating objectives. It is possible to relax some of these
conservatisms for use in the Technical Specification. However, relaxation of the conservatisms are likely to result in more significant
changes to the structure of the Technical Specification.
First, the above equation (and the existing Technical Specification 3.7.1.1) is conservative since it is based on the maximum number of
inoperable MSSVs per loop.
For example, a representative four loop
plant, in accordance with the current Technical Specification, should
reduce the neutron flux setpoint to 87% if it has up to one inoperable
MSSV on each loop. This means that the plant should use this setpoint
whether there are one, two, three or four inoperable MSSVs, as long as
there is only one inoperable MSSV per loop. Thus, the existing Technical
Specification and the above equation are conservative and bounding.
However, any relaxation of this conservatism must be interpreted with
care. The reason is that the steam generators must be protected from an
overpressurization condition during a loss of load transient. There are
several events that could lead to a loss of load, including the
inadvertent closure of one or all MSIVs.
The affected steam generator
must have a sufficient number of operable MSSVs to protect it from an
overpressurization condition, if the MSIV (or MSIVs) was inadvertently
closed.
Attachment 1
August 22, 1994 Another conservatism in the above equation (and the existing Technical
Specification 3.7.1.1) is in w which is the minimum total steam flow
rate capability of the operable MSSVs on any one steam generator. This
value is conservative since it assumes that if one or more MSSVs are
inoperable per loop, the inoperable MSSVs are the largest capacity MSSVs, regardless of whether the largest capacity MSSVs or the smaller capacity
MSSVs are inoperable. The assumption has been made so that the above
equation is consistent with the current structure of Technical
Specification 3.7.1.1.
(2) As an alternative, plant-specific LOL/TT analyses could be performed to
maximize the allowable power level for a given number of inoperable
MSSVs. Depending on key specific plant parameters, these analyses may be
able to justify the continued validity of the current Technical
Specification.
(3) Consider modifying, as required, the Bases Section for Technical
Specification 3/4.7.1.1 so that it is consistent with the plant safety
analysis. The safety analysis criterion for preventing
overpressurization of the secondary side is that the pressure does not
exceed 110% of the design pressure for anticipated transients. However, in reviewing several plant technical specifications, it was noted that
the bases for some plants state that the safety valves insure that the
secondary system pressure will be limited to within 105 or even 100% of
design pressure. This is not consistent with the safety analysis basis
and should be revised to indicate 110%.
Attachment 2
August 22, 1994 LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
Information
Date of
Notice No.
Subject
Issuance
Issued to
94-30,
Supp. 1
94-59
94-58
94-57
94-56
Leaking Shutdown Cooling
Isolation Valves at
Cooper Nuclear Station
Accelerated Dealloying of
Cast Aluminum-Bronze
Valves Caused by Micro- biologically Induced
Corrosion
Reactor Coolant Pump
Lube Oil Fire
Debris in Containment
and the Residual Heat
Removal System
Inaccuracy of Safety Valve
Set Pressure Determinations
Using Assist Devices
Problems with Copes-
Vulcan Pressurizer
Power-Operated Relief
Valves
Deficiencies Found in
Thermo-Lag Fire Barrier
Installation
Failures of General
Electric Magne-Blast
Circuit Breakers to
Latch Closed
Possible Malfunction of
Westinghouse ARD,
BFD,
and NBFD Relays, and
A200 DC and DPC 250
Magnetic Contactors
08/19/94
08/17/94
08/16/94
08/12/94
08/11/94
08/04/94
08/04/94
08/01/94
07/29/94
All holders of OLs or CPs
for nuclear power reactors.
All holders of OLs or CPs
for nuclear power reactors.
All holders of OLs or CPs
for pressurized water
reactors.
All holders of OLs or CPs
for nuclear power reactors.
All holders of OLs or CPs
for nuclear power reactors.
All holders of OLs or CPs
for nuclear power reactors.
All holders of OLs or CPs
for nuclear power reactors.
All holders of OLs or CPs
for nuclear power reactors.
All holders of OLs or CPs
for nuclear power reactors.
94-55
91-79, Supp. 1
94-54
91-45, Supp. 1 OL = Operating License
CP = Construction Permit