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UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001

August 22, 1994

NRC INFORMATION NOTICE-94-60: POTENTIAL OVERPRESSURIZATION OF

MAIN STEAM SYSTEM

Addressees

All holders of operating licenses or construction permits for pressurizedwater 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

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IN 94-60

August 22, 1994

Page 2 of 2

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

IN 94-60

August 22, 1994

Page 1 of 9

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

IN 94-60

August 22, 1994

Page 2 of 9

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

IN 94-60

August 22, 1994

Page 3 of 9

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

SP = 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

IN 94-60

August 22, 1994

Page 4 of 9

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

transient. 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

IN 94-60

August 22, 1994

Page 5 of 9

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

IN 94-60

August 22, 1994

Page 6 of 9

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

feedwater). Thus, an algorithm for use in defining the revised Technical

Specification table setpoint values would be:
Attachment 1

IN 94-60

August 22, 1994

Page 7 of 9

(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

IN 94-60

August 22, 1994

Page 8 of 9

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

IN 94-60

August 22, 1994

Page 9 of 9

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

IN 94-60

August 22, 1994

Page 1 of 1

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 Microbiologically 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 CopesVulcan 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