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NUREG/CR-6001

PNL-8020

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g Assessment of BWR

dby Liquid Control Systems

I~~~~~~~~~~~~~~

Prepared by

G. D. Buckley, R. D. Orton, A. B. Johnson, Jr., L. L Larson

Pacific Northwest Laboratory

Operated by

Battelle Memorial Institute

Prepared for

U.S. Nuclear Regulatory Commission

J. F. MARCHATERRE

DIRECTDR

AUC 1t 1992

I Reactor EngineeringI
AVAILABILITY NOTICE

Availability of Reference Materials Cited In NRC Pubiations

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Although the listing that follows represents the majority of documents cited In NRC publications, It Is not

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Referenced documents avaliable for Inspection and copying for a fee from the NRC Public Document Room

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NUREG/CR-6001

PNL-8020

RM, RV, R9

Aging Assessment of BWR

Standby Liquid Control Systems

Manuscript Completed: July 1992

Date Published: August 1992

Prepared by

G. D. Buckley, R. D. Orton, A. B. Johnson, Jr., L L Larson

Pacific Northwest Laboratory

Richland, WA 99352

Prepared for

Division of Engineering

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

NRC FIN B2911
Abstract

Pacific Northwest Laboratory conducted a Phase I aging

assessment of the standby liquid control (SLC) system

used in boiling-water reactors. The study was based on

detailed reviews of SLC system component and operating experience information obtained from the Nuclear

Plant Reliability Database System, the Nuclear Document System, Licensee Event Reports, and other databases. Sources on sodium pentaborate, borates, and

boric acid, as well as the effects of environment and

corrosion in the SLC system were also reviewed to characterize chemical properties and corrosion characteristics of borated solutions.

setpoint drift in relief valves, which has been discovered

during routine surveillance and is thought to be caused

Uby mechanical wear. A higher setpoint results in loss of

system overpressure protection, and a decrease in setpoint results in a reduction of boron injection rate.

Degradation was also observed in pump seals and internal valves, which could prevent the pumps from operating as required by the technical specifications. In

general, however, the results of the Phase I study indicate that age-related degradation of SLC systems has

not been serious.

Relatively few SLC component failures were attributed

to sodium pentaborate buildup or corrosion. The leading aging degradation concern to date appears to be

iii iii ~~~~~NUREG/CR-60
Contents

Page

Abstract ........................................................................................ iii

Glossary ............................................................................. v

Summary ............................................................................. xi

1 I ntroduction ............................................................................. 1

1.1 Overview of SLC System ................................................................... 1

1.2 Safety Significance of the SLC System ............................ 1.......................

1.2.1 Risk Analysis ........................................................................ 1

1.2.2 Regulatory Guidance ................................................................. 2

1.3 Scope of Evaluation . ...................................................................... 2

1.4 Scope of Report . ......................................................................... 2

2 SLC System Description . ...................................................................... 5

2.1 Description of SLC Components ...................................... .5

2.1 Storage.1ink ........................................................................ 5

2.1.2 Tank Heaters .................................... ................. 5

2.1.3 Pump Suction Valves (SLC-01A, O1B) ............ ...................................... 5

2.1.A Injection Pumps (lA, 1B) ................................... . S

2.1.5 Relief Valves (RV-01A, -O1B) .................................. . 6

2.1.6 Injection Valves (SLC-04A, -04B) ................................... . 6

2.1.7 Containment Isolation Valves (SLC-06,-07, &-08) .................................. . 6

2.1.8 Injection Sparger .. 6

2.1.9 lzst Tank ....................................................................... 6

2.1.10 Neutron Absorber . .................................................................. 6

v NUTREG/CR-6001
Contents

Page

2.2 Verifying System Performance .............................................................. 7

2.3 System Operation . ........................................................................ 7

2.3.1 Standby Mode ....................................................................... 7

2.3.2 Injection Mode . ...................................................................... 7

2.3.3 Recirculation Test Mode ................... ........................................... 8

2.3.4 Simulated Injection Test Mode .............. .......................................... 8

2.4 SLC Emergency Operating Procedures ...................................................... 8

2.4.1 System Initiation ..................................................................... 8

2.4.2 EOP Contingency . .................................................................... 8

2.5 System Differences . ....................................................................... 9

2.6 SLC System Operating History ....................... ....................................... 9

3 Aging Assessment Methodology ................................................................ 11

3.1 Database Review ........................................................................... 11

3.2 Characterization of Sodium Pentaborate Effects .............................................. 11

4 Results of Database Review . .................................................................... 13

4.1 Databases ............................................................................. 13

4.2 Screening Database Results . ................................................................ 13

4.3 Preliminary Results . ...................................................................... 13

4.3.1 Relief Vilves ........................................................................ 13

4.3.2 Accumulators . ....................................................................... 13

4.3.3 Pumps .............................................................................. 13

4.3.4 Instrumentation . .................................................................... 14

4.4 Summary of Database Review ................... ........................................... 14

5 Sodium Pentaborate Characterization ................ ........................................... 15

5.1 Use of Sodium Pentaborate ..................... ........................................... 15

NUREG/CR-6001 vi
Contents

Page

5.2 Preparation .................... . .......................................... 15

5.3 Properties ................... , .. ...................................... 15

5.3.1 NuclearProperties ......................................................... 15

5.3.2 Chemical Properties ... 15

6 Corrosion in SLC System Environments . .. 19

6.1 SLC System Environments and Materials . . .19

6.2 Assessment of Expected Corrosion Characteristics.. 19

6.2.1 Corrosion-Induced Mailures of Stainless Steel in Aerated, Borated Solutions . . . 19

6.2.2 FMilure Study Conclusions and Recommendations . . . 20

6.3 Comparison of FSP and SLC System Conditions .............. . . ......... , 20

7 Conclusions and Recommendations ... 23

8 References ........ I ,................. 25

vii vii ~~~~~~NLREO/CR
Contents

Page

Figures

1.1 Standby liquid control system schematic .............. ........................................... 3

2.1 Boron injection temperature curve ................... ........................................... 8

Tables

4.1 Effects of SLC component degradation ............................ ............... 14

NURIEG/CR-6001 ilgeiii
Summary

Pacific Northwest Laboratory (PNL),' in support of the

U.S. Nuclear Regulatory Commission (NRC) and its

Nuclear Plant Aging Research (NPAR) Program, conducted a Phase I study to determine the effects of agerelated degradation on the standby liquid control (SLC)

system used in boiling-water reactors (BWRs). The goal

of this study was to develop an overview of the aging

characteristics of SLC systems in sufficient depth to

assess whether further research is needed to detect and

manage aging degradation. The scope of this evaluation

encompassed all SLC system designs found in comnercial BWRs.

The SLC system is safety-related because it provides

backup capability for reactivity control in the event that

the normal operating systems fail. Based on calculations

performed to determine the relative importance of reactor protection systems, the safety importance of the SLC

system, as measured by core-melt frequency, is eighth of

approximately thirty reactor protection systems. Additionally, guidance is provided from several NRC Regulatory Guides that help to establish the safety significance

of the SLC system.

The study of the SLC system involved reviews of information on SLC system components and operating experience, which were obtained from the Nuclear Plant

Reliability Database System, the Nuclear Document

System, Licensee Event Reports, NRC Generic Issues,

and other databases. Sources on sodium pentaborate,

borates, and boric acid, and the effects of environment

and corrosion in the SLC system were also reviewed to

characterize chemical properties and corrosion characteristics of borated solutions. Characterizing the sodium pentaborate solution provides a useful basis for

assessing potential aging mechanisms in SLC systems.

Relatively few SLC system failures were connected to

sodium pentaborate buildup or corrosion of valve internals. Current industry experience on corrosion in SLC

systems indicates that there is no unpredicted corrosion

of components in contact with the sodium pentaborate

solution because the solution provides a mild environment. Additionally, intergranular stress corrosion

cracking (IGSCC) of the type that occurred in

pressurized-water reactor (PWR) spent fuel storage

pools is not expected to occur in SLC components.

Higher pH levels in the sodium pentaborate solutions is

one of the mitigating factors compared to the lower pH

levels in the boric acid chemistries of spent fuel pools.

The results of the Phase I study indicate that the aging

mechanisms active in the SLC system result in the failure of only a few specific components. A review of

reported incidents indicates that the system relief valves

(SRVs) are the major component affecting the operation of the system. The SRVs are susceptible to setpoint

drift, which has been discovered during routine surveillance. >Though the root cause of relief valve failures was

not always determined, the majority of the reported failures were attributed to mechanical wear. Instances of

degradation of pump seals, internal valves, and packing

,were also identified. This degradation could prevent the

SLC pumps from operating as required by the technical

specifications.

Although most SLC system components are being

evaluated in other studies, insights obtained from this

investigation suggest opportunities for improved aging

management to minimize the risk of component corrosion in the presence of sodium pentaborate, specifically

the risk of IGSCC.

The following are practices which, if implemented,

would further minimize the risk of failures caused by

IGSCC:

(1) The contaminant level of chloride and sulfate

should be independently evaluated, to verify the

certification values received from the Borax and

boric acid vendor, before addition of the chemicals

to the SLC system storage tank.

(2) The concentration of contaminant chlorides and

sulfates in the SLC storage tank should be monitored periodically to determine whether concentrations are at levels that could contribute to IGSCC.

ix ix ~~~~~~NUREG/CR-6
Summary

(3) Selected, specific nondestructive evaluations

should be conducted to detect whether IGSCC is

occurring, particularly in weld heat-affected zones.

In general, the results of the Phase I study indicate that

age-related degradation of SLC systems has not been

serious. Deterioration of relief valve performance has

been the leading SLC system aging degradation problem; therefore, further study of relief valves is recommended. If the study were conducted, the principal

aging vulnerabilities of the SLC system would be addressed, and a separate Phase U aging study of SLC systems would not be necessary.

NUREGICR 6001 x
Glossary

BIT

BWR

DES

EOP

FSP

HAZ

IGSCC

LER

NPAR

NPRDS

NRC

NUDOCS/AD

PNL

PWR

RES

RPV

SLC

SRV

boron injection temperature

boiling-water reactor

Division of Engineering Safety

emergency operating procedure

fuel storage pool

heat-affected zones

intergranular stress corrosion cracking

License Event Reports

Nuclear Plant Aging Research

Nuclear Plant Reliability Database System

Nuclear Regulatory Commission

Nuclear Documentation System

Pacific Northwest Laboratory

pressurized-water reactor

Office of Nuclear Regulatory Research

reactor pressure vessel

standby liquid control

system relief valve

la xi ~~~~~~NUREG/CR-6
1 Introduction

The U.S. Nuclear Regulatory Commission (NRC),

Office of Nuclear Regulatory Research (RES), Division

of Engineering Safety (DES), is implementing the ''

Nuclear Plant Aging Research (NPAR) program plan

(US NRC 1991) to resolve technical safety iisues related

to the aging of commercial nuclear power plants.

NPAR aging assessments involve identifying and characterizing material and component degradation during

service.

The NPAR aging assessments are being conducted to:

* identify and characterize aging effects that could'

cause degradation and thereby impair plant safety

* identify methods of inspection, surveillance, and

monitoring that Will ensure timely detection of significant aging degradation before loss of safety

function

* evaluate the effectiveness of storage, maintenance,

and replacement practices in mitigating the rate and

extent of aging degradation.

In support of the NPAR ProgramnPacific Northwest

Laboratory PNL1 conducted a Phase I aging assessment

of the standbyliquid control (SLC) system used in com-,

mercial boiling-water reactors (BWRs). 'The objective:

of this irzestigation was to evaluate the SLC system to

deternine if further aging research Is needed to detect

and mitigate degradation due to aging mechanisnms such

as corrosion and-mechanical wear. This preliminary

investigation comprised reviews of information on SLC

system events and component aging phenomena.'

1.1 Overview of SLC System

The SLC system is unique to General Electric BWR

designs. The SLC system is a backup reactivity control

system and is maintained in an operable status whenever

1

PNL is operated for the Department of Energy by Battelle Memorial

Institute under Contract DE-ACO6-76RL0 1830.

a control rod is removed from a loaded reactor core. 'In

this mode, the SLC system performs t safety-related

function. 1lRb of the design-basis requirements'for the

SLC system are as follows:

* Backup capability for reactivity control is provided,

independent of normal reactivity control, to facili-"

tate shut down of the reactor if notmal reactivity

control becomes inoperative.

* The system has the capacity for controlling the reactivity difference between the steady-state rated operating condition and the cold shutdown condition,

including shutdown margin, to ensure complete

shutdown from the most reactive condition at any

time in core life.'

1.2 Safety Significance of the SLC

System

The SLC system is important to safety because it provides backup capability for reactivity control in the event

of failure of the normal operating systems. When initiated, the SLC system injects neutron absorbing sodium

pentaborate solution into the reactor vessel to shut

down the reactor, independent of any control rod '

motion, and maintains the reactor in a subcritical condition. The following two subsections address the safety

ranking and classification of the SLC system.

1.2.1 Risk Analysis

In a study by Vo et aL (1990), risk analysis calculations

were performed to determine the relative importance of

each reactor protection system. The initial order of

ranking was determined by the Birnbaum Importance

Measure, ie., the change in risk that is associated with a

total system failure leading to core melt. The SLC system importance to safety as measured by core-melt frequency was eighth out of approximately thirty reactor

protection systems. This initial ranking can also be used

as an indicator of the SLC system relative importance

within the NPAR program aging assessments.

1 1 ~~~~~~NLUREG/C
Introduction

1.2.2 Regulatory Guidance

The NRC has developed regulatory guides that designate the relative Importance of systems and components

to ensure their adequacy, availability, and capability to

perform their intended functions. The following regulatory guides support the safety significance of the SLC

system:

NRC Regulatory Guide 1.70, -Standard Format and

Content of Safety Analysis Reports for Nuclear

Power Plants*" (US NRC 1978a) This regulatory

guide classifies the SLC system as a 'System

Required for Safe Shutdown.' The explosive

(squib) valves, valves and piping between the isolation valves, and the reactor pressure vessel are classified as Safety Class 1.

The following items were excluded from the final SLC

system evaluation. They were, however, examined for

significance to the components included in the

evaluation.

* electrical breakers

* motor control centers

* cables

* pUMp motors

* motor operators for valves

* pipe supports

* snubbers.

* NRC Regulatory Guide 1.26, 'Quality Group Classifications and Standards for Water-, Steam-, and

Radioactive-Waste-Containing Components of

Nuclear Power Plants.' (US NRC 1976) This regulatory guide classifies the SLC system explosive

valves, valves, and piping between the isolation

valves and the reactor pressure vessel as Quality

Group 'A'.

* NRC Regulatory Guide 1.29, 'Seismic Design Classification.' (US NRC 1978b) This regulatory guide

classifies the SLC system as Seismic Category L

1.3 Scope of Evaluation

The scope of this evaluation encompasses all SLC systems found in commercial BWRs. Because of differences in nuclear plant design and product lines, SLC

system nomenclature and component makeup differ

from plant to plant Regardless of the nomenclature

and physical design variations, the system that provides

the functional purpose indicated for the SLC system is

included in the scope of this evaluation. A simplified

one-line diagram of the SLC system, that illustrates the

scope of this evaluation, is shown in Figure 1.1.

These components are being evaluated under separate

NPAR program aging assessments.

1.4 Scope of Report

Section 2.0 of this report describes the important components of the SLC system, the various operation

modes, and system differences reflecting the design of a

particular BWR fatclity. Sections 3.0 and 4.0 present

the SLC aging assessment methodology and results.

Section 5.0 characterizes technical information on

sodium pentaborate solutions and describes the use of

sodium pentaborate in the SLC system and the chemical

properties of the solution. Section 6.0 discusses corrosion and corrosion-related failures in the SLC system

environments. Section 7.0 presents the conclusions and

recommendations based on the results of the SLC system aging assessment

NUREG/CR-60012 2
Introduction

"11Zt2

Figure 1.1 Standby liquid control system schematic

3 NUREG/CR.6001
2 SLC System Description

The SLC system described in this section represents systems at typical BWR facilities. The system differences

that may be encountered at different facilities are discussed in Section 2.5.

The SLC system shown in Figure 1.1 consists of a heated

storage tank, two 100% capacity pumps, two motor

operated suction valves, two explosive actuated discharge valves, and the necessary valves and associated

piping to inject the sodium pentaborate into the reactor

vessel In addition, the system includes a test tank with

the necessary valves and piping to adequately test system

performance by injecting demineralized water instead of

sodium pentaborate into the reactor pressure vessel

(RPV).

The SLC system is manually initiated from the control

room according to the facility emergency operating procedures (EOP) when conditions exist that prompt the

operators to take action to assure control of core reactivity. Several BWR facilities also incorporate a logic

circuit to automatically start the SLC system at predetermined conditions.

The SLC system is relatively slow compared to other

shutdown systems; It takes approximately two hours to

inject sufficient sodium pentaborate to shut down the

reactor. The system cannot be considered as a backup

to the control rods for accidents requiring a rapid shutdown or reactor scram. The SLC does have the capacity

to shut down the reactor with adequate margin to control the reactivity difference between the steady-state

rated operating condition and the cold shutdown condition. This margin of control is ensured during the

period of the core life when reactivity is greatest, and

takes into account dilution by the reactor recirculation

system and non-homogenous mixing.

21 De of SLC Components

The SLC components shown in Figure 1.1 are described

in functional sequence, starting from the sodium pentaborate storage tank to the point at which the-sodium

pentaborate is injected into the RPV

2.1.1 Storage aink

The storage tank is stainless steel with about a 5200 gallon capacity. It is equipped with an overflow line at

about the 5150 gallon level An internal air sparger is

used to mix the solution. The pump suction piping

enters the side of the tank to prevent plugging of the

suction line.

2.1.2 Tank Heaters

Heaters in the tank maintain the solution temperature

in the 21 to 300C (70 to 86WF) range to keep the sodium

pentaborate in solution. Additional heaters are also

installed in the tank to permit heating the contents to

about 650C (1500F) for preparation of the sodium

pentaborate solution.

2.1.3 Pump Suction Valves (SLC-OIA, -GlB)

There are two motor-operated suction valves. These

valves are normally closed to prevent inadvertent injection of the sodium pentaborate into the RPV during system testing. Both valves open on system initiation and

each valve is interlocked with the pump so that the

pump will not start until one of the valves is fully open.

These valves are also interlocked with the test tank supply valve so that the pump suction valves do not open if

the test tank supply valve is open. Heat tracing on the

pump suction line betwieen the storage tank and the

SLC-01 pump suction valves keeps the solution in the

lines from crystallizing during stagnant operation. The

suction line to each pump is cross-connected to allow

both suction valves to supply both injection pumps.

2.1A [ujection Pumps (IA, 1B)

There are two 100% capacity pumps. These pumps are

positive displacement type with motors rated at about

40 hp. Each pump is rated at 43 gpm at 1250 psig. The

pumps are interlocked so that they will not start unless

one of the two pump suction valves (SLC-O1A or 01B) is

5s 5NUREG/CR-6001
SLC System Description

fully open or the test tank supply valve is open. The

local pump start switch bypasses the pump start

interlocks.

2.1.5 ReliefValves (RV-OIA,-GiB)

Each pump discharge is protected with a relief valve that

discharges back to the pump suction line. The setpoint

for the relief valve is 1400 psig. A check valve downstream of the relief valve connection on each pump discharge line prevents the sodium pentaborate solution

from bypassing through the idle pump or through the

relief valve in the event that it opens below setpoint.

2.1.6 Injection Valves (SLC-04A, -04B)

The discharge line from each pump has an associated

explosive actuated valve. During system initiation the

explosive valve receives an open signal that causes the

charges to explode. The explosion forces a ram to

extend and shear off the top of the sealing end cap

within the valve. The ram and cap form in a shape that

prevents flow blockage. The explosive charge circuitry is

monitored continuously for continuity by a small current. The discharge lines from each pump are cross connected at a point just upstream of the injection valves to

allow both injection valves to discharge for either injection pump.

2.1.7 Containment Isolation Valves

(SLC-06, -07, -08)

Passive isolation of the SLC system is provided by two

in-line check valves in series, one inside containment

(SLC47) and one outside containment (SLC-06). Vent

valves between the two check valves allow detection of a

leaking check valve. A manual isolation valve (SLC-08)

located downstream of the inside containment check

valve is a maintenance valve. It is normally locked open.

2.1.8 Injection Sparger

The injection sparger penetrates the RPV at the bottom

head near the edge of the core below the core plate. It is

a tube-in-tube design with the inner tube being used for

sodium pentaborate injection. The end of the inner

tube is perforated to allow better distribution of the

sodium pentaborate during injection. The injection

sparger is also used for other RPV functions and instrumentation such as:

* detecting breaks in high pressure core spray lines

* jet pump differential pressure tap

* measuring the core differential pressure

* measuring the bottom head drain flow to the reactor

water cleanup system

* control rod drive water and cooling water differential pressure tap.

2.1.9 Test Unk

The SLC system is provided with a test tank that is normally filled with demineralized water. The tank is stainless steel and holds about 210 gallons. The test tank is

used to flush and to performance test the SLC system.

The tank is piped to the suction of the SLC pumps and

is provided with a manual isolation valve. The SLC

pump discharge is also piped to the test tank and is controlled by two manual isolation valves in series.

2.1.10 Neutron Absorber

Sodium pentaborate solution is used as the reactivity

control agent because boron-lO has a high absorption

cross section for thermal neutrons (Wlker et al. 1983).

The limits for temperature, tank level, and concentration ensure that the minimum shutdown requirement of

660 ppm boron concentration within the 50- to

125-minute period is met. The boron reactivity control

limits are summarized as follows:

* 660 ppm boron concentration is used to shut down

the reactor, 185 ppm is added for non-homogenous

mixing; 250 ppm is added for residual heat removal

shutdown cooling dilution, which yield a total boron

concentration of 1095 ppm

* the temperature should be maintained above 10rC

(50F) to ensure that the solute remains in solution

NUREG/CR-6001 6
SLC System Description

* 125 minutes maximum to ensure that the sodium

pentaborate is injected fast enough to overcome

cooldown and xenon depletion reactivity effects

* 50 minutes minimum to ensure that there are no

power oscillations caused by high boron concentrations cycling in and out of the core due to poor

mixing

* the concentration and volume limits are generally

applied within an operating envelope. Operating

within the envelope ensures that limits a, c, and d

above are met.

2.2 Verifying System Performance

Adequate performance of the SLC system is verified by

several methods. The following are examples of verifications that are required by technical specifications:

* periodic verification of sodium pentaborate solution

volumes concentration, and temperature

* periodic verification of heat trace temperature

* periodic system performance check by recirculating

demineralized water from/to the test tank

* continuity check circuitry for the explosive valves

* initiation of SLC injection, including the explosive

valve and performance check, while injecting demineralized water from the test tank into the RPV

* verification of ASME Boiler Code Section XI

requirements for minimum flow and pressure

* verification that the heat-traced piping between the

storage tank and the RPV is unblocked

* verification that the tank heaters are functioning

properly

2.3 System Operation

The SLC system operates in a variety of modes. These

modes are described in the following subsections.

23.1 Standby Mode

The SLC system must be maintained in an operational

standby status when there is fuel in the reactor vessel

and the control rods are removed. The time that the

SLC system can be out of service during operation is

limited by technical specifications. The system is in

standby when the valves are in the following lineup:

* pump suction valves (SLC-01A, -O1B) CLOSED

* test tank supply valve

* injection valves (SLC-04A, -04B)

* injection valve continuity lights

CLOSED

CLOSED

ON

The SLC solution temperature, concentration, and tank

level also have limits that determine if the SLC is

operable in standby.

23.2 InJection Mode

The SLC system is initiated with keylocked switches.

Each switch controls one 100% capacity pump subsystem. When the switch for Pump IA is placed in the

RUN position, the following actions occur:

(1) the pump suction valve (SLC-O1A) opens

(2) both explosive pump discharge valves (SLC-04A,

-04B) actuate and open

(3) the reactor water cleanup system is isolated

(4) the SLC Pump 1A starts when the suction valve is

full open.

* verification of the relief valve set point.

7 7 ~~~~~~NUREG/CR-60
SLC System Description

The SLC storage tank level is monitored during injection by the control room operators to ensure that the

solution is being injected into the RPV Reactor power

and other complementary plant parameters are monitored closely to track decreasing power. When the reactor is confirmed to be shutdown or the SLC tank level is

at zero and further injection is not required, the switch

is placed in the STOP position.

2.3.3 Recirculation ¶st Mode

The recirculation test mode allows operators to verify

proper operation of the SLC system pumps without

injecting solution into the reactor. This test can be performed during reactor operation or shutdown periods.

The recirculation test mode allows the system to be

lined up to the test tank so that demineralized water can

be recirculated from the tank, to the pumps, and back to

the tank.

23A Simulated Injection fst Mode

This mode is performed periodically during reactor

shutdown conditions. The procedure is to align the suction How path of the SLC pumps to the test tank, then

to activate the system with the control switch. The following automatic actions are verified: the proper system

valves open, the SLC pumps start, and demineralized

water is injected into the RPV In this way pump discharge capacity and explosive valve performance can be

verified. 7ypically, one explosive valve is fired each year,

alternating every year. This test and replacement cycle

appears adequate to address explosive valve aging

concerns.

2.4 SLC Emergency Operating

Procedures

2.4.1 System Initiation

The SLC system is manually initiated in accordance with

EOPs when plant conditions require injection. With the

reactor at full power and with heat being added to the

suppression pool, it is possible that the heat capacity

limit could be exceeded. The heat capacity limit is based

upon not exceeding the design containment pressure or

the design suppression pool temperature if an emergency RPV blowdown occurs.

The boron injection temperature (BMI) limit is defined

by a curve (Figure 2.1) with reactor power on the x-axis

and suppression pool temperature as the y-axis. The

BIT curve is used by the operators to determine the conditions at which SLC injection must be initiated to

ensure that injection of the hot shutdown weight of

boron can be completed before the heat capacity limit is

exceeded.

If the shaded area of the BIT curve is entered, the SLC

system must be activated. It is important to point out

that if an anticipated transient without a scram event is

in progress and even though the suppression pool temperature is less than the BIT, the operator may initiate

the SLC in anticipation of exceeding the BIT limit. The

operator is guided to start SLC injection as soon as possible if insertion of the control rods, by any means, is

unsuccessfuL

2.4.2 EOP Contingency

The BWR emergency operating procedures utilize the

SLC system as a low priority source of makeup water to

the reactor pressure vesseL During emergency conditions in which adequate high pressure makeup water is

200

ISO~ ~~ ~ ~ ~~~~ f

I * 1I I I san_ _ _

0 ~~2 4 6 S 10 12 .14 R.aclm P0wmf)*flI.

Figure 2.1 Boron Injection temperature curve

NUREG/CR-60018 8
SLC System Description

not available, a contingency action may be required to

use the SLC system as a source of makeup water to the

RPV The procedure to Implement this contingency

action is to manually valve in the test tank to the SLC

pump suction, as would be performed for the simulated

injection test mode. The system is then activated so that

demineralized water from the test tank is injected into

the RPV The demineralized water fill valve is throttled

open to keep the tank full.

2.5 System Differences

The SLC system components, operating procedures, and

boron concentrations can differ depending on the design

of a BWR facility. These differences are as follows:

* Boron-iO content can vary from natural (19.9 atom

percent) to as high as 54.5 atom percent.

* Some facilities require that both SLC pumps be

started and inject solution to meet the minimum

shutdown criteria.

* Some facilities utilize automatic initiation logic circuits to activate the SLC system.

* Some facilities have manual suction valves that

remain open instead of motor-operated suction

valves that are maintained closed and are automatically opened when the system is initiated.

lines between the pump suction valves (SLC-O1A, -O1B)

and the pump suction remain full of water and are free

of sodium pentaborate solution. This helps prevent line

plugging caused by crystallization of the sodium pentaborate if the pump suction valves leak.

* Some facilities have accumulators on the pump discharge lines to dampen the pressure pulsation from

operation of the positive displacement pumps.

* Some facilities Inject the sodium pentaborate solution into the high pressure core spray injection line

*at a point just downstream of the inboard isolation

valve.

2.6 System Operating History

Most BWR safety systems are multipurpose, that is, they

have more than one mode of operation. However, the

SLC system and its components have one major purpose: to inject neutron absorbing solution (poison) into

the reactor. Some facility emergency operating procedures list the SLC system as an alternative method of

high pressure injection. However, the SLC system is a

low priority high pressure source, because It has such a

low flow rate. Since this system has never been called

into emergency service to serve its intended purpose, the

SLC system's entire operating history is recorded in its

testing and maintenance records.

* Some facilities have a head tank that maintains a

volume of demineralized water at an elevation

higher than the storage tank level to ensure that the

', .~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

71_ ,. \

NUREG/CR-6001
3 Aging Assessment Methodology

3.1 Database Review

Part of the preliminary aging assessment of the SLC system involved a review of Industry and regulatory databases for SLC system operating experience. The

databases consulted for this study were as follows:

* Nuclear Plant Reliability Database System

(NPRDS)

* Nuclear Documentation System

(NUDOCS/AD)/licensee Event Reports (LERs)

* NRC Generic Issues.

These databases were evaluated to determine if system

components evidenced age-related degradation. Those

instances where aging degradation was evident were

evaluated to determine whether or not the aging effects

would cause the SLC system to faLil when required to

operate during an emergency. The results are presented

in Section 4.0.

3.2 Characterization of Sodium

Pentaborate Effects

Characterizing sodium pentaborate effects involved literature searches, including a review of publications on

the subject of sodium pentaborate, borates and boric

acid. The emphasis was to describe the chemical properties and corrosion characteristics of borated solutions.

Findings from the literature searches are presented in

Sections 5.0 and 6.0.

11 11 ~~~~~~NLJREG/CR
4 Results of Database Review

4.1 Databases

The NPRDS contained 510 operating records related to

the SLC system. Each record provided a description of

the event, the cause, the resolution, or what was done to

repair the problem.

The NUDOCS/AD database had 122 LERs that applied

to the SLC system.

4.2 Screening Database Results

The database information was screened to segregate

instances of aging degradation from immediate damage

due to operator error (procedural error), maintenance

error (e g., failure to tighten connections), abuse

(e.g., valves broken by unknown causes) and design

error (upgrading required to meet system demands).

Normal aging degradation, produced by error-free preservice and operating conditions (Grant and Miller

1992), was typically detected during testing and operator

rounds.

Once the age-related failures were identified, it was

determined whether the failure, if undetected, would

hinder the operation of the SLC system during critical

need. The original NPRDS and LER files were then

condensed to 110 records.

4.3 Preliminary Results

The reliability of the SLC system components was

assessed based on the screened database information.

The four components that evidenced aging degradation

were the relief valves, accumulators, pumps, and the

SLC system instrumentation. The nature of the degradation and the potential effects on operation of the SLC

system are summarized in Tbble 4.1 and discussed in the

following sections.

4.3.1 Relief Valves

The system relief valves were the most troublesome

component identified in the final 110 records. Relief

valve failure was noted 84 times. Relief valve failure is

considered to affect SLC system operation In two ways.

First, if the relief valve does not open during operation

the system could lose its over-pressure protection capability. The system requires over-pressure protection

because positive displacement pumps have the capability to easily damage the system if a vent path does not

exist. Second, if the valves leak by or lift early the system could bypass its injection flow back to the pump

suction or the storage tank and thereby reduce the

boron injection rate in times of critical need

The root cause of the failure of the relief valves was difficult to assess from the descriptions given in the databases. Fifty-five of the failures were attributed to

setpoint drift due to mechanical wear. Eight of the failures could be connected to sodium pentaborate build

up. Only two failures were related to corrosion of the

valve internals.

43.2 Accumulators

The system accumulators were found to maintain operational integrity during service. Although the accumulators were subject to age-related degradation, the impact

of the degradation was not considered detrimental to

SLC system operation during critical need. In most

cases, the degradation was identified through a4partial

or complete loss of the nitrogen blanket pressure. Loss

of nitrogen pressure was caused by valve wear And failure of the gas bladder. Even though the pressure was

lost, the accumulators stayed intact and would still pass

flow when needed.

4.3.3 Pumps

Twelve records showed that pump failures would have

hindered SLC system operation. Most of the failures

resulted from the aging degradation of seals and internal

check valves. These failures typically prevented the

pump from operating within technical specifications for

a given flow at a given pressure. These failures were

found during surveillance testing. Packing leaks are

13 13 ~~~~~NUREG/CR
Results

Table 4.1 Effects of SLC component degradation

SLC component Identified form(s) of degradation Potential effect(s) on operation

Relief valves 55 valve failures attributed to set- Relief valve failure could hinder

point drift, probably caused by SLC operation; higher valve setmechanical wear, 8 failures attrib- point results in a loss of system

uted to sodium pentaborate over-pressure protection; a

buildup; 2 failures related to cor- decrease in setpoint could reduce

rosion of valve internals. the rate of boron injection.

Accumulators Instances of partial or complete Degradation effects minimal;

loss of nitrogen blanket noted. accumulator would function

properly during critical need.

Pumps Degradation of seals and internal Degradation could prevent

valves identified; packing leaks pumps from operating within

also noted. technical specifications.

Instrumentation Some degradation noted. Minimal effects; Instruments are

not necessary for system opera-

-tion during critical need.

considered an aging effect, but failure of packing has not

been considered a critical issue that would hinder the

operation of the SLC system during critical need.

43.4 Instrumentation

The instrumentation required to monitor the system,

such as component status lights, tank level and temperature gages, and system pressure, flow, and explosive

valve continuity gages, is subject to aging but is not necessary for system operation during critical need.

4.4 Summary of Database Review

Information from the maintenance data base, NPRDS,

has shown that the SLC system relief valves are the most

troublesome system component that could affect system

operation during critical need. However, the validity of

the data base depends on the participation from utilities

and the accuracy of the information provided by the

utilities. Discussions with component repair vendors

could be valuable in determining the root cause of the

relief valve failures.

NUREG/CR-6001 14
5 Sodium Pentaborate Characterization

This section summarizes the information obtained during a literature review of the characteristics of the

sodium pentaborate used in the SLC system Characterizing the sodium pentaborate solution provides a useful

basis for assessing potential aging mechanisms in the

SLCsystem.

5.1 Use of Sodium Pentaborate

The amount of sodium pentaborate available for injection into the RPV depends on the volume of stored

solution, the usable volume of the storage tank, and the

inventory of boron-10 isotope in the solution. The

minimum required concentration of sodium pentaborate In solution depends on two factors: 1) the pumping

capacity, and 2) the volume of solution for injection.

With two pumps available and a capacity of 5000 gallons

of solution, the sodium pentaborate concentration in

weight percent can range from about 10% to about 24%.

The normal concentration is about 12%, which corresponds to about 4300 gallons of solution to provide sufficient boron for reactivity controL

The maximum concentration of sodium pentaborate in

solution is limited by the saturation point in water. The

saturation point depends on the solution temperature

and the ratio of Borax to boric acid. For example, at a

solution temperature of 300C (860F) the saturation concentration can be as high as 14% by weight An increase

of dissolved boron, compared to boric acid dissolved in

water alone or Borax dissolved in water alone, is attrbuted to the formation of the polyborate ion.

5.2 Preparation

The chemical formula for sodium pentaborate decahydrate is Na2O -5B203410H20. It is formed when

Borax (Na2B4 07) and boric acid (H3BO3) are mixed

with water. This reaction occurs when the molar ratio of

Na2O/B2 03 is equal to about 0.2 (Kirk-Othmer 1964,

p. 85). Heat is released when boric acid and borate solutions are mixed, indicating that the formation of the

polyborates is exothermic (Kirk-Othmer 1964, p. 79).

The formation of sodium pentaborate depends on the

relative amounts of the reactants and the temperature.

For example, at 30WC (860F) with a mole ratio of 0.27, a

new polyborate, 2Na20-5.1B20 3-7H20, begins to

form. The same compound begins to form at 50C (410F)

with a mole ratio of 022.

53 Properties

53.1 Nuclear Properties

There are two naturally occurring Isotopes of boron.

The boron-10 proportion is 19%6 to 20%, and that of

boron-il is 80% to 81%. The boron-10 isotope is of

particular importance in the operation of nuclear reactors because it has a large cross section for thermal neutron capture. Most BWR facilities use boron with naturally occurring isotope concentrations. Some BWR

facilities are using boron with a boron-10 enrichment up

to 54 atom percent.

5.3.2 Chemical Properties

53.2.1 Solubllity

Sodium pentaborate is much more soluble than the individual Borax or boric acid compounds. When Borax is

added to a saturated boric acid solution or boric acid is

added to a saturated Borax solution, the solubility,

measured as percent B203 (borate) in solution,

increases markedly. For example, at 300C (860F), the

saturation limit for dissolved borate, Borax, or boric

acid in water alone is <4%; a mixture of Borax and

boric acid can be as high as 14% (Nies and Cambell

1964). The increase is evidence of the formation of the

polyborate ions in solution. Polymerization removes

from solution some of the boric acid molecules and

borate ions that are in equilibrium with the solids.

More Borax and boric acid can therefore dissolve (Nies

and Cambell 1964).

The solubility of sodium pentaborate also increases with

increasing temperature. Solubility, expressed in units of

15 NUREGICR-6001
Characterization

percent anhydrous salt by weight, increases from 6.28%

at 0WC (320F) to 50.30% at 100TC (212F). The temperature of the SLC storage tank is normally maintained at

about 30NC (86WF), which results in a 13.75 wt% solution

(weight of sodium pentaborate per 100 wt of saturated

solution) (Nies and Cambell p. 100, 'Tble VI). The

solubility of the B203 at 30°C ranges from about 11.5%

to 14.5%, corresponding to molar ratios of 0.19 to 0.25

(Nies and Cambell 1964, p. 82, Figure 5). The sodium

pentaborate concentration is normally kept at about

12%. This is less than the saturation limit for 30WC

(86WF)-

Note that the sodium pentaborate concentration examples given in the previous paragraph use the anhydrous

salt form (molecular weight is 410 wt/mole) for determining the weight percent Industry experts state that

the standard calculation of weight percent of sodium

pentaborate in solution uses the decahydrate salt form

(molecular weight is 590 wt/mole). Therefore, 13.75%

sodium pentaborate by weight per 100 wt of solution

would yield 3.63% boron if anhydrous sodium pentaborate is used and 2.52% boron if sodium pentaborate

decahydrate is used. This difference is significant

because 1095 ppm boron is required to be converted

into weight percent of sodium pentaborate (decahydrate) in solution.

5.3.2.2 pH Characteristics of Sodium

Pentaborate Solutions

The pH values of aqueous solutions of sodium pentaborate at 20°C (680F) as a function of weight percent are

(Kirk-Othmer 1964, p. 85):

Based upon the 13.75 wt% given above, the pH should

be in the range from 7.1 to 6.8, adjusted for temperature. In general the pH is near neutral.

53.23 Stability

At 10(rC (2120F) crystalline sodium pentaborate, the

decahydrate, was found to decompose to form the dihydrate and a solution of 58% NaB508 (US NRC 1978a).

This is not expected to be a problem because sodium

pentaborate is batch mixed from Borax and boric acid as

needed; the mixture is not stored in crystalline form.

If hot sodium pentaborate solutions are agitated for

some time, the compound 2Na209B2O 3411H 2 0 will

crystallize if seed crystals are present (US NRC 1978a).

This process only occurs at temperatures near boiling

and thus is not expected to be a problem because the

storage tank solution is maintained at about 300C

(86PF).

Sodium pentaborate solution when left undisturbed for

long periods (more than one week) will stratify into concentration gradients (US NRC 1986). The SLC storage

tank is routinely agitated. This practice maintains the

tank solution well mixed to prevent concentration

stratification.

5.3.2.4 Environmental Conditions: Borated Section of

SLC Systems

General Electric specifications for boric acid indicate

that the purity limit of the granular boric acid must be

99.90% by weight. The chloride impurity limit is

0.0001% by weight.

The US BORAX Service Bulletin No. 1179 identifies

Borax as a corrosion inhibitor. The bulletin states that

"The buffering action of Borax (5 molar or 10 molar)

maintains pH above 7, where acidic compounds cannot

form." The bulletin further states that "Borax inhibits

corrosion by minimizing the rate of oxidation at the sure

face of the metal and is classified as an anodic inhibitor."

Wt%

0.5%

1.0%

2.0%

3.0%

5.0%

10.0%

14.0%

pH

8.5

8.4

8.2

8.0

7.7

7.1

6.8

The US BORAX Service Bulletin No. BX-21 states that

anodic inhibition is "...accomplished by growing an oxide

film on the surface of the metaL This phenomenon is

NUREG/CR-6001 16
Characterization

called passivation and is actually a form of controlled

general corrosion. The bulletin further states that

Borax is a "nonoxidizing anodic inhibitor because [it]

has insufficient oxidizing power of [its] own to effect

passivation (i.e. oxygen must be present)."

The portion of the SLC system in contact with the

sodium pentaborate solution is the storage tank to the

pump suction valves (SLC-01). The concentration of

chlorides and sulfates is the maximum stated in the

material specification for the boric acid. The environmental conditions in the borated sections of SLC systems are as follows:

Temperature

Normal

Mixing

Pressure

pH

Chlorides

Sulfates

Oxygen

21-30mC (70-860F)

650C (150(F) (storage tank only)

<10 psig

>6.8

<0.0001% by weight (<1 ppm)

<0.0002% by weight (<2 ppm)

saturated

These conditions appear to be less aggressive than the

conditions that cause degradation of stainless steels (see

following section).

17 17 ~~~~~~NUREG/CR
6 Corrosion in SLC System Environments

To assess the potential for corrosion of SLC system

components to develop, it is necessary to define the

material and environmental factors and to assess their

relationship to known corrosion regimes. The failure

databases do not reveal specific failures in SLC systems

that are attributed to corrosion: However, it is important to consider whether corrosion can be expected to

emerge as service life is extended.

6.1 SLC System Environments

and Materials

The principal environments in typical SLC systems are

as follows:

* aerated sodium pentaborate solution at 650

C

(150F during mixing and 21-300C (70-860F) during

static operation in the tank and heat-traced lines

* deionized water (aerated) in the test circuits and

tank at ambient temperatures

* exterior surfaces of stainless steel pipes, valve

bodies, and tanks exposed to air in the temperature

range 10 to 3WC (50 to 860F).

The principal material in the SLC system tanks and

pipes is stainless steel. The specific types have not been

surveyed in detail, but both normal grades

(e.g., alloy 304) and low-carbon grades (e.g., 316L) are

represented.

The SLC system resides at low pressures and temperatures in normal operation (see Section 5.3.2). It operates static except during runs in the test circuit.

The factors that contribute to corrosion of stainless

steels include: a contaminated environment, corrosionprone (e.g., sensitized) microstructures, and, in the case

of stress-assisted phenomena, high stress levels.

Because the SLC system is safety-related, high-integrity

weld procedures would have been required, suggesting

that high levels of sensitization would not be expected in

weld heat-affected zones. Stress levels in SLC system

components have not been characterized specifically,

but their effects, if present, seem to be mitigated by

other favorable factors considered below.

6.2 Assessment of Expected

Corrosion Characteristics

Stainless steels are generally not prone to serious corrosion in deionized water under conditions that exist in

the SLC system. Also, the SLC-air-side environments

appear to be benign for stainless steels unless contaminants foreign to normal conditions are introduced.

The principal corrosion consideration in the SLC system

focuses on the portion of the system that is in contact

with borated solutions.

61.1 Corrosion-Induced Failures of Stainless

Steel in Aerated, Borated Solutions

6.2.1.1 Documented Failures

There have been numerous failures of stainless steel in

borated solutions at low temperatures [20-400C (68-

1040 F)]. The failures are illustrated by two cases cited

below. The first involved failure of the piping in the fuel

storage pool cooling system, in a standby loop; the second involved separation of the upper end fitting from an

irradiated fuel assembly in a spent fuel storage pool

(FSP).

6.2.1.2 Three Mile Island 1 Fuel Pool Piping

At the Three Mile Island Unit 1 fuel storage pool, cooling loop A, nine sections of pipe from the fuel pool were

observed to have small through-wall cracks in weld heataffected zones (HAZ). The failures occurred in the

redundant section of the pool coolant system (loop A),

which remained stagnant for most of the time from commissioning in 1974 to observation of the cracks in 1979.

A study (Bruemmer and Johnson 1984) indicated that

sulfur and chlorine were present in the fracture area in a

weld heat-affected zone, and the HAZ material was

sensitized.

19 19 ~~~~~~NUREG/CR-60
Corrosion in SLC System

6.2.13 Prairie Island 1 Fuel Assembly End

Fitting Separation

6.3 Comparison of FSP and SLC

System Conditions

The upper end fitting (nozzle) separated from an irradiated Prairie Island Unit 1 fuel assembly during handling

in the spent fuel pool in December 1981. The suspected

failure mode of the Type 304 stainless steel sleeves in

the Prairie Island-l fuel assembly was IGSCC, which

occurred while the fuel assembly was stored in the FSP

(Bailey and Johnson 1983). The stainless steel guide

tubes were sensitized when the upper tie plate assembly

was heat treated. Bulges that joined these guide tubes

to the Zircaloy tubes were a potential source of stress.

A contaminant which might have contributed to the

IGSCC was not identified but was speculated to be this

sulfate. The NRC is treating this failure as an isolated

case.

The following table summarizes information comparing

FSP and SLC system conditions.

Condition

Temperature, 'C

Pressure

Oxygen

pH

Chlorides, ppm

Sulfates, ppm

FSP

20-50°C

(68-1220F)

static head

of pool

saturated

4.5

<0.15

SLC

21-30°C

(20-680F)

static head of

storage tank

saturated

>6.8

<1.0

6.2.2 Failure Study Conclusions and <2.0

Recommendations

A study by Jones et aL (1981) indicated that Type 304

stainless steel is susceptible to intergranular stress corrosion cracking (IGSCC) under fuel pool conditions.

There have been, however, '. . . no IGSCC failures

observed at 320C (900F) in borated water without [chloride ion) additions,' addressing laboratory tests (Jones

et al. 1981).

Johnson et al. (1981) state that crack initiation in

Type 304 stainless steel in boric acid at 400C (1040F) is a

significant step in the fracture process. The study suggests that several factors be avoided when using

Type 304 stainless steel in FSP: 1) carbon concentrations should be kept below 0.07 wt%; 2) weld preparation grinding should be done with care; 3) weld heat

should be minimized; and 4) system impurities such as

chlorides and thiosulfates should be minimized.

The study by Jones et al. (1981) concludes that 'high

carbon (0.07%) 304SS can be sensitized sufficiently by

welding or heat treating (24 hrs at 6200C) to cause

IGSCC at 320C (900F) in borated water containing

15 ppm [chloride ion]."

The similarities between the FSP and SLC system conditions are stated below

* both involve borated chemistries

* both involve oxygenated conditions

* both operate in similar temperature and pressure

regimes

* both operate under stagnant conditions

* both have stainless steel as the principal material

of construction.

The literature suggests that stress corrosion cracking

generally occurs more aggressively at low pH in lowtemperature aqueous environments, although there are

variations, particularly as temperatures increase (Theus

and Staehle 1973).

NUREG/CR-6001 20
Corrosion in SLC System

Wield (1982) indicates that oxide film breakdown has

an induction time that appears to be extended by

increasing pH and borate content and decreasing chloride and temperature.

The testing that is required by technical specifications

for the SLC borated solution addresses only the concentration of the sodium pentaborate and in some cases the

atom percent of boron-10. It does not address the concentration of chlorides or of sulfates. Therefore, it is

recommended that the concentrations of the contaminant chlorides and sulfates in the SLC storage tank be

monitored as a basis to control susceptibility to stress

corrosion cracking.

The main potential sources of contaminants in the SLC

solution are the bulk chemicals, boric acid and Borax.

The contaminant level of chloride and sulfate ions

should be evaluated periodically, to verify original certifications, before addition to the storage tank.

Key differences between the FSP and SLC are as

follows:

* The pH of the FSP solution is typically -4.5; in the

SLC solution the pH is near neutral (-7.0).

The higher pH in the SLC system would be expected to

mitigate IGSCC of stainless steeL Higher weld integrity

is another potentially favorable factor that has not been

investigated systematically.

In summary, the current industry experience on corrosion in SLC systems indicates that there is no unpredicted corrosion of components in contact with the

sodium pentaborate solution. This result is expected

because the components are subject to relatively mild

conditions: low levels of chloride, higher pH range, low

temperature, and low pressure. The types of failures

experienced in PWR fuel storage pools are not expected

to occur in SLC system materials, based on the less

aggressive conditions. However, selected monitoring of

SLC system components for IGSCC is recommended,

particularly at locations that may have been subject to

sensitization (weld heat-affected zones).

* The source of boron in the SLC system is sodium

pentaborate; in the FSP system, the boron source

is boric acid.

21 21 ~~~~~~NUREG/CR
7 Conclusions and Recommendations

The SLC system is unique to BWRs and is designed to

provide alternative reactivity control in the event that

control rods can not be inserted into the reactor core.

The SLC system Is a stand-alone system and is rarely

operated except to perform required surveillances and

inspections for technical specification compliance.

A survey of reported incidents related to the aging of the

SLC system indicates that degradation of system relief

valves (SRVs) could prevent the valves from opening

properly and thus reduce the boron injection rate. The

relief valves have experienced setpoint drift, which has

been discovered during routine surveillances and is

likely attributed to mechanical wear. Therefore, additional study of the SLC relief valves is recommended. A

minimum relief valve testing interval of every refueling

outage is also recommended.

Degradation of SLC pump seals and pump internals as

well as packing leaks were also noted; this could prevent

the pumps from operating within technical specifications. Additionally, degradation of accumulators has

occurred resulting in partial or complete loss of the

nitrogen blanket.

The SLC sodium pentaborate solution provides a mild

environment with respect to age-related degradation.

The intergranular stress corrosion cracking (IGSCC)

that occurs in PWR fuel storage pools is not expected to

occur in SLC systems. Industry experience and expert

opinion suggest that the SLC system is less vulnerable to

IGSCC than the boric acid conditions that have resulted

in IGSCC in other systems.

The following are practices which, if implemented,

would further minimize the risk of failures due to

IGS-CC:

(1) The contaminant level of chloride and sulfate

should be independently evaluated, to verify the

certification values received from the Borax and

boric acid vendor, before addition of the chemicals

to the SLC system storage tank.

(2) The concentration of contaminant chlorides and

sulfates in the SLC storage tank should be monitored periodically to determine whether concentrations are at levels that could contribute to IGSCC

(3) Selected, specific nondestructive evaluations

should be conducted to detect whether IOSCC is

occurring, particularly in weld heat-affected zones.

In summary, the effects of age-related degradation on

the SLC system have been minimal. Deterioration of

system relief valve performance has been the most significant aging problem. Additional study of the SLC

relief valves and testing of the valves during every refueling outage is recommended. If the NPAR safety relief

valve study were conducted, then the principal vulnerabilities of the SLC system would be addressed, and a

separate NPAR Phase II aging study of the SLC system

would not be necessary.

23 23 ~~~~~~NUJREG/C
8 References

Bailey, W J., and A. B. Johnson Jr. 1983. Wet Storage

Update. PNL-4726, Pacific Northwest Laboratory,

Richland, Washington.

Bruemmer, S. M., and A. B. Johnson, Jr. 1984. "Effect

of Chloride, Thiosulphate Additions on the IGSCC

Resistance of 1ype 304 Stainless Steel in LowIbmperature Water." In Proceedings of International

Symposium on Degradation of Materials in Nuclear

Power System Water Reactors Myrtle Beach, South

Carolina, August 1983.

Grant, W S. and E. J. Miller. 1992. Nuclear Power

Plant Common Aging 7brminolo v. Research Project

2927-7 (Draft), Electric Power Research Institute, Palo

Alto, California.

Johnson, Jr., A. B., W J. Bailey, E. R. Bradley,

S. M. Bruemmer, and D. C. Langstaff. 1981. Annual

Report FY 1980 Spent Fuel Pool Component Integritv.

PNL-3868, Pacific Northwest Laboratory, Richland,

Washington.

Jones, R. H., A. B. Johnson Jr., and S. M. Bruemmer.

1981. An Evaluation of Stress Corrosion Cracking of

Sensitized 304SS in Low Temperature Borated Water.

PNL-SA-9387, Pacific Northwest Laboratory, Richland,

Washington.

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New York.

Nies, N. P., and G. W Cambell. 1964. Boron. MetalloBoron Compounds and Boranes ed. R. M. Adams.

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Theus, G. J., and R. W Staehle. 1973. 'Review of

Stress Corrosion Cracking and Hydrogen Embrittlement in the Austenitic Fe-Cr-Ni Alloys." In Stress Corrosion Cracking and Hydroeen Embrittlement of Iron

Base Alloys. NACE-5 Ed. R. W Staehle, J. Hochmann,

R. D. McCright, and J. E. Slater. National Association

of Corrosion Engineers, Houston, lbxas.

Tlyfield, S. P. 1982. The Corrosion of Sensitized

304 Stainless Steel in Aqueous Borate and Chloride

Solutions - An Electrochemical Study. TPD/B0124/N82,

Central Electricity Generating Board, Berkeley Nuclear

Laboratories.

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California.

U.S. Borax and Chemical Corporation. 1983. 'Borax as

a Corrosion Inhibitor in Automotive Engine Coolants."

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Corporation, Los Angeles, California.

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and Radioactive-Waste-Containing Components of

Nuclear Power Plants.' NRC Regulatory Guide 1.26,

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U.S. Nuclear Regulatory Commission. 1978a. "Standard Format and Content of Safety Analysis Reports for

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U.S. Nuclear Regulatory Commission,

Washington, D.C.

U.S. Nuclear Regulatory Commission. 1978b. "Seismic

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U.S. Nuclear Regulatory Commission,

Washington, D.C.

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the Standby Liquid Control System." NRC Information

Notice 86-48, U.S. Nuclear Regulatory System,

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NUREG-1144, Rev. 2., U.S. Nuclear Regulatory Commission, Washington, D.C.

25- NUREG/CR-6001-
References

Vo, T. V et aL 1990. "Development of Generic Inservice Inspection Priorities for Pressure Boundary Systems." Nuclear 7Tchnologv. 93(3), American Nuclear

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of the Nuclides. Thirteenth Edition, General Electric

Company, San Jose, California.

NUREG/CR-6001 26
NUREG/CR-6001

PNL-8020

Distribution

No. of

Copies

OFFSIrE ONSITE

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System

3000 George Washington Way

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Oak Ridge, TN 37831

37 Pacific Northwest Laboratory

R. P. Allen, P8-10 (10)

D. E. Blahnik, PS-10

G. D. Buckley, K6-46 (5)

M. E. Cunningham, P8-10

L R. Dodd, K648

A. B. Johnson, Jr., P8-10

L L Larson, K7-82 (5)

R L Moffitt, K6-46

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Publishing Coordination

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Attn: J. J. Burns (20)

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Distr.a
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

12-89) lAssigned by NRC. Add VOL, Sum., aev.. NRCM 1102. and Addendum Numbers. If any.)

3201.3202 BIBLIOGRAPHIC DATA SHEET

(See infltuctions on the bevrfse NUREG/CR-6001

2. TITLE AND SUBTITLE PNL-8020

Aging Assessment of BWR Standby-Liquid Control Systems 3. DATEREPORTPUBLISHED

MONTH YEAR

'August 1992

4. FIN OR GRANT NUMBER

B2911

5. AUTHORIS) 6. TYPE OF REPORT

G. D. Buckley, R. D. Orton, A. B. Johnson, Jr., L. L. Larson Technical

7. PERIOD COVERED (inclusive Oares)

B. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC. prem O4Division, Office Region, gr U.S Nuclea ReguAvtt Conn auiond mailing address. it contracror, posve

-ame aed Dealh addss.)

Pacific Northwest Laboratory

Richland, WA 99352

9. SPONSORING ORGANIZATION -NAME AND ADDRESS (If NRC. type "3, as above-; ifeonrtrprovde NRC Division, Office or Region. U.S NscleaRegulaorY Commiiobn,

and Awang eddressJ

Division of Engineering

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 ewor son)

Pacific Northwest Laboratory conducted a Phase I aging assessment of the standby liquid

control (SLC) system used in boiling-water reactors. The study was based on detailed

reviews of SLC system component and operating experience information obtained from the

Nuclear Plant Reliability Database System, the Nuclear Document System, Licensee Event

Reports, and other databases. Sources on sodium pentaborate, borates, and boric acid,

as well as the effects of environment and corrosion in the SLC system were also reviewed

to characterize chemical properties and corrosion characteristics of borated solutions.

Relatively few SLC component failures were attributed to sodium pentaborate buildup or

corrosion. The leading aging degradation concern to date appears to be setpoint drift

in relief valves, which has been discovered during routine surveillance and is thought

to be caused by mechanical wear. A higher setpoint results in loss of system overpressure protection, and a decrease in setpoint results in a reduction of boron injection rate. Degradation was also observed in pump seals and internal valves, which

could prevent the pumps from operating as required by the technical specifications. In

general, however, the results of the Phase I study indicate that age-related degradation

of SLC systems has not been serious.

12. KEY WORDS/DESCR!PTORS (Lit ewods aphreftat will amist eerehers in locaing rheo .poor.I 13. AVAILABILITY STATEMENT

Unlimited

boiling-water reactor 14. SECURITY CLASSIFICATION

standby liquid control system IThs Page)

sodium pentaborate Unclassified

safety-related system Tmhs Re"n

aging degradation Unclassified

Nuclear Plant Aging Research 15. NUMBER OF PAGES

16. PRICE

NRC FORM 33512491
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