Hydrogen water chemistry for BWRs

Progress in Nuclear Energy. Vol. 20. No. I. pp. 43 71), 1987.

0149-1970/87 $(L0(I + .50 Copyright ~ I9{47PergamonJournals lad

Printed in Great Britain. All rights reserved.

H Y D R O G E N W A T E R C H E M I S T R Y FOR BWRS

WARREN

BII.ANIN, DANIEL CUBICCIOTFI, ROBIN

L.

JONES, ALBERT J. MACHIELS,

LARRY NELSON and CHRIS J. WOOD

Nuclear Power Division, Electric Power Research Institute, 3412 Hillview Avenue, Palo Alto, CA 94303, U.S.A.

(Received 16 June 1987)

Abstract - Intergranular stress corrosion cracking (IGSCC) has been responsible for more than 1,000 cases of cracking in austenitic stainless steel piping systems in boiling water reactors (BWRs). This paper presents the status of efforts in the United States to prevent ICSCC in BWRs during power operation by modifying the chemistry of the reactor water. The technical basis for this alternative water chemistry, called hydrogen water chemistry (HWC), is described and the results are presented of an ongoing in-plant program, at Commonwealth Edison's Dresden-2 plant, to verify the HWC concept and systematically assess the consequences of using it in an operating BWR. In addition, progress toward implementation of HWC at other U.S. plants is summarized. I.

INTRODUCTION

Cracking in stainless steel piping systems has been a costly problem in boiling water reactors (BWRs) since 1974. The underlying phenomenon, intergranular stress corrosion cracking (ICSCC) of the sensitized material adjacent to girth welds, has been extensively studied and a number of effective remedies have been developed and are being implemented. These remedies are based on improved materials and reduced local stresses. The results of laboratory investigations have also provided a basis for the development of a potentially attractive water chemistry remedy for ICSCC, called hydrogen water chemistry (HWC). This paper summarizes the progress of the HWC development program in the United States. Intergranular stress corrosion cracking (IGSCC) of sensitized material adjacent to welds in Type 304 and Type 316 stainless steel piping systems has been responsible for more than i000 cases of pipe cracking in boiling water reactors (BWRs) since the problem first became apparent in 1974. Although these cracks are not thought to pose a serious threat to public health and safety (Cheng et al., 1980) inspections and repairs associated with BWR pipe cracking have proved costly to the utillties and substantial R&D programs have been undertaken to understand the ICSCC phenomenon and develop remedial measures (Jones et al., 1985). Mechanistic studies have shown that at a fundamental level, the initiation and growth of IGSCC involves a complex interplay between the rates of a number of processes occurring simultaneously in the metal, in the water environment, and at the metal-environment interface (see, for example, Ford, 1982). These studies have also revealed that from a practical point of view, there are three separate approaches that can be taken to mitigate pipe cracking, namely, to reduce either the tensile stress level, or the susceptibility of the material~ or the aggressiveness of the service environment. Much of the early remedy development work focused on alternative materials or local stress reduction and a number of successful remedies were developed and are being

43

W. BILANIN et al.

44

implemented. More recently, it was recognized that water chemistry control might offer a feasible method of controlling pipe cracking in operating BWRs, and the effects of water chemistry parameters on the ICSCC process have received increasing attention. A complete understanding of the interrelationship between BWR water chemistry variables and ICSCC of sensitized stainless steels has not yet emerged but some important features have been identified, resulting in the development of an alternative, less aggressive water chemistry for BWRs based on injection of hydrogen to control the oxidizing species produced by the radiolysis of water. 2.

TECHNICAL BASIS FOR HYDROGEN WATER CHEMISTRY

Laboratory studies (Ford, 1982; Childs, 1983; Dillman et al., 1985) have shown that ICSCC can occur in simulated BWR startup environments but that most of the damage leading to pipe cracking probably occurs during power operation (Childs, 1983). The reactor water in a BWR during power operation is at about 280°C and contains radiolysis products formed in the core (principally ~200 ppb oxygen and ~15 ppb hydrogen but also peroxides and other short-lived species) and small quantities of ionic and nonionic impurities (Ljungberg and Hallden, 1984; Dillman et al., 1985) which find their way into the reactor water from a variety of sources (Fig. i). The effects on ICSCC of many of these substances have been investigated (Andresen, 1983; Cordon, 1980; Andresen and Indig, 1982; Indig et al., 1983; Davis and Indig, 1983; Kurtz et al., 1983; Cordon et al., 1983; Burley, 1982; Indig and Weber, 1983; Gordon et al., 1985; Ruther et al., 1983); dissolved oxygen and ionic impurities have received particularly detailed attention. Numerous laboratory investigations (Ford, 1982; Cordon et al., 1983; Gordon et al., 1985) have demonstrated that ICSCC of sensitized austenitic stainless steels will not occur in high-purity water at 280°C if the dissolved oxygen content of the water is less than a critical value. Although the quantitative results vary somewhat, the critical value is typically about 20 ppb. Mechanistically, this observation is thought to be associated with the effect of dissolved oxygen on the corrosion potential of stainless steel, which is a measure of the thermodynamic driving force for corrosion. As shown in Fig. 2, the stainless steel corrosion potential decreases with decreasing dissolved oxygen content~ falling most rapidly in the range of oxygen concentrations where ICSCC disappears. Although detailed fundamental understanding is still incomplete, it is believed that the value of the corrosion potential determines whether or not IGSCC of sensitized stainless steel is thermodynamically possible (Ford, 1982). Thus, as the dissolved oxygen content of laboratory water is decreased~ the stainless steel corrosion potential falls and at about 20 ppb oxygen, it reaches a value below which the thermodynamic driving force is insufficient to support ICSCC. This value is often termed the protection potential~ and laboratory measurements for sensitized austenitic stainless steels in high purity water usually give estimates falling in the range -200 to -350 mV on the standard hydrogen electrode (SHE) scale.

Inorganlcs Resins Condensate demineralizer

Resins Reactor water I cleanup system ~~I

Residualheat removalsystem

Reactor Inorganlcs Organics

Inorganlcs Organlc~

Condensate storagetank ~ makeupwater! Inorganlcs Organ,cs

T

I Radwaste

I

Inorganlcs Resins

T

Inorgan*os Organics

Resins

Figure

1. Potential sources of impurities in BWR reactor water.

Inorganlcs Organics

Hydrogen water chemistry for BWRs

45

0.2 =~

0.1

-

-

Range o f . _ . ~

. . , , , ~ ~ ' ~

o

~. -0.1

g ~

- 0.2

-0.3 -0,4 "E ~o -0,5 I I 10 100 Dissolved Oxygen (ppb)

-0.6

Figure

1000

2, Relationship between dissolved oxygen and corrosion potential for Type-304 stainless steel,

Part of the reason why laboratory measurements of the IGSCC protection potential give different results is because the protection potential depends on the quantities and identities of the ionic impurities present in the water. It has been known for some time that certain ionic impurities aggravate IGSCC, and several investigators have reported the results of laboratory studies of the effects of individual impurity species on IGSCC initiation and growth rates (Andresen, 1983; Gordon, 1980; Andresen and Indig, 1982; Indig et al., 1983; Davis and Indig, 1983; Kurtz et al., 1983; Ruther et al., 1983). In many cases, these studies have been performed as a function of the stainless steel corrosion potential. Fig. 3 is a schematic representation of the type of behavior that has been observed for aggressive anions like Cl- and SO4 . It is apparent in figure 3 that corrosion potential and ionic impurity concentration must both be considered in specifying a BWR water chemistry to protect sensitized austenitic stainless steels from ICSCC. At stainless steel corrosion potentials typical of those measured (Gordon et al., 1983; Burley, 1982) in BWR reactor water during power operation (50 ± i00 mV SHE), sensitized austenitic materials are susceptible to IGSCC even when the ionic impurity concentration in the water is so low that the conductivity approaches the theoretical value for pure water (0.056 ~S/cm at 25=C). In addition, the rate of cracking is observed to increase with increasing impurity content. This means that although BWR pipe cracking cannot be prevented by conductivity control alone, maintenance of the best possible water quality should prolong piping life (BWR Guidelines, 1985). However, a region of immunity to IGSCC in laboratory tests does exist in low conductivity water at lower corrosion potentials (Fig. 3). The precise value of the protection potential in water of a given conductivity depends on the ionic impurities present and also on the nature of the IGSCC test procedure. For the most aggressive impurity species at concentrations leading to a conductivity of 0.3 ~S/cm, the protection potential is about -230 mV SHE (Gordon et al., 1985). By using good operational practices, reactor water conductivities below 0.3 ~S/cm can be maintained in BWRs for essentially 100% of the time at power. Therefore, BWR pipe cracking can be prevented if a way can be found to reduce the stainless steel corrosion potential during power operation to values less than -230 mV SHE while still maintaining good water quality (conductivity <0.3 ~S/cm). This is the basic concept underlying hydrogen water chemistry, which uses hydrogen injection into the feedwater as the method of reducing the corrosion potential of stainless steel in the reactor water.

46

W. BILANIN et al.

Stainlesssteel ] corrosion potential 2 1 Increasing Normal crackingrate BWR rangeL / > /

BWRrHnV~C-~

Y~'~ Y

~

~

Increasing . ~ ~

j

Rangeof protection potentials )

Impurityconcentration or conductvity

Figure 3. Schematicsummaryof the resultsof laboratorystudies of the effectsof impuritieson intergranularstress corrosion cracking(IGSCC)of sensitizedaustenitic stainless steels.

In-reactor experiments (Burley, 1982) have shown that the dissolved oxygen content of the reactor water of an operating BWR can be sufficiently reduced to attain corrosion potentials below the protection values by injecting hydrogen into the feedwater. The feasibility of this approach was first demonstrated in two short-term tests in a Swedish BWR and, subsequently, a more detailed one-month investigation was conducted in the United States in 1982 at Commonwealth Edison's Dresden-2 plant (Burley, 1982). Some of the measurements made at Dresden-2 are shown in Fig. 4. These data indicate that adding ~1.5 ppm hydrogen to the feedwater suppresses the oxygen content of the recirculating water to i0 to 20 ppb at full power. Electrochemical measurements showed that this lowers the corrosion potential of stainless steel to below -230 mV SHE. Based on these results, the HWC regime at Dresden-2 was preliminarily defined as 520 ppb 02 combined with ~0.3 ~S/cm conductivity. In-plant slow strain rate tests confirmed that IGSCC of sensitized austenitic stainless steel was suppressed when Dresden-2 was operating in this regime (Indig and Weber, 1983; Gordon et al., 1981) and also supported laboratory data indicating that HWC is a more innocuous service environment than normal BWR water for other plant structural materials (Gordon et al., 1985). Currently, a long-term HWC verification program is in progress at Dresden-2. The work began in 1983 and is planned to last two or three 18-month fuel cycles. The scope includes surveillance of fuel and core materials performance, additional evaluation of plant structural materials behavior, water chemistry monitoring, and assessment of the radiological impact of HWC. The reactor water chemistry goals are ~ 20 ppb oxygen and ~0.3 ~S/cm conductivity for >90% of the time at >25% rated power (conditions shown by testing to establish corrosion potentials <0.230 V, SHE in Dresden-2. In a nuclear plant, species other than oxygen affect the corrosion potential of stainless steel (e.g., peroxides, some metal ions) and a variety of ionic impurities contribute to water conductivity. The amounts of these species present in the reactor water of an operating plant vary with time and this is likely to lead to a time-varying margin against IGSCC.

Hydrogen water chemistry for BWRs

47

250

~" 200

'- 100 o

O

~, 50

=

20

~

10

.o_

83o/o

I

I

I

I

I

0.5 1.0 H 2 Concentration in Feedwater (ppm) Figure

4.

1 1.5

Recirculation water oxygen concentration versus feedwater hydrogen concentration at Dresden-2 for two reactor power levels.

Accordingly, the Dresden-2 program includes periodic corrosion potential measurements and IGSCC tests to verify that the HWC specification (~20 ppb oxygen combined with ~0.3 uS/cm conductivity), established on the basis of the initial one-month feasibility study~ is adequate to provide continuing protection against IGSCC at Dresden-2 over the long term. The later sections of this paper update a previous summary of the status of the HWC verification program at Dresden-2 (Roberts et al., 1985) by detailing the results obtained through 1986. Progress that has been made recently toward implementing HWC at other plants in the United States is also described. 3.

HWC VERIFICATION

PROCRAM AT DRESDEN-2

The hydrogen water chemistry verification program at Dresden-2 (at the beginning of operating cycle 9) with the objectives of

started

in April

1983

demonstrating the feasibility of operating a BWR in the HWC regime continuously during power operation and defining the radiological consequences verifying that ICSCC does not initiate and that existing cracks do not propagate while the plant is operating within the HWC regime and investigating the effects of transient operation outside of the HWC regime determining

the effects of HWC on the performance

of fuel and core components.

The plant operated on HWC throughout cycle 9 and has continued with HWC during cycle I0, which began in April 1985 and ended in December 1986.

48

W. BILANIN et al.

3.1

Operational

experience

during cycles 9 and i0

3.1.1 Injection system performance. The hydrogen addition flowsheet is shown in Fig. 5. High-purity (99.999%) hydrogen is purchased as gas which is delivered in tank trucks carrying 115,000 or 50,000 scf at 2400 psi. These tanks supply hydrogen at about 200 psi through a 1/2" carbon steel llne into a flow control panel in the condensate pump room. This panel varies the hydrogen flow in proportion to feedwater flow to give a constant but adjustable hydrogen concentration in the reactor feedwater. Maintenance of feedwater hydrogen at 1.5 ppm results in dissolved oxygen concentrations of 20 ppb or less throughout the primary coolant system. The hydrogen addition system is not operated continuously. It is shut off for a variety of reasons, inc|udin~ maintenance activities on the hydrogen addition system itself, maintenance in other areas of the plant where there are high N-16 radiation levels due to hydrogen addition, and during the extinguishing of fires in the off-gas treatment system. In addition, during reactor startups and shutdowns, hydrogen addition is not used when reactor power is below 25% maximum rated power. When the hydrogen injection system is operating, it significantly reduces the oxygen concentration in the reactor water. During 1984 and 1985, the injection system was available for about 90% of the time and the oxygen was controlled below 20 ppb over 80% of the time (Fig. 6). It is worthwhile to investigate the nature of the time the reactor water was greater than 20 ppb oxygen because of the existence of a "memory effect" discussed below. As a result of this effect, IGSCC appears to be suppressed for times up to about i0 hours after shutdown of the hydrogen injection system. The statistical database shows that greater

Reactor ~ T u r b i n e preTSsU;e~ ~.j'~-~ondenser ~ ~<',--7 ~J~ ~ Demineralizer

Feedwatersystem

® [o'en"a' llICo°='ex'en=l measurements ratetest ]

Condensate demineralizer effluent

Finalfeedwater 3 Coreexitwater(plannedbut neverinstalled) 4 Recirculationloop 5 Cleanupinlet 6 Cleanupoutlet 7 Mainsteam Figure 5 Hydrogenadditionflowsheet.

Hydrogen water chemistry for BWRs

~

49 4

99.99 99.90

3

~,. 9 9 . 0 0

90.00 o

0 o

A- 50.00

_~ lO.OO E

d

__

1.00 0.10 0.01 0.1

Figure 6

Cycle

S

-

9

-1

-2 -3

III

I IIIIIIII t11111111 I I}lllJ-4 1.0

10.0 Oxygen (ppb)

100.0

1000.0

Probability plot of oxygen data for Dresden-2 during cycles 9 and 10.

than 70% of the cumulative time the Dresden-2 reactor water was above 20 ppb was for periods of less than I0 hours in any given 24-hour period. As a result, it is thought that over 90% of the total cumulative operating time during cycles 9 and i0 has been in the "IGSCC-immune" regime. 3.1.2 Off-gas system performance. Under normal BWR operation, oxygen and hydrogen are produced in stoichiometric amounts by core radiolysis. These noncondensible gases are treated by off-gas systems to recombine the 02 and Hp into water. For various reasons~ fires have been observed in these systems. Off-gas "fires" mean that O2/H 2 recombination is taking place within the piping upstream of the recombiner. Although this is not a desirable operating condition, off-gas systems are designed to accommodate this occurrence. The composition of radiolytic gas going to the off-gas system changes under HWC. At Dresden-2~ the hydrogen concentration remains about the same and the oxygen c+oncentration decreases three-fold. Because Dresden-2 uses a catalytic recombiner to treat this gas, oxygen must be added upstream of the recombiner under HWC operation to assure complete H 2 removal. Oxygen from a liquid oxygen storage trailer provides 60 psi oxygen gas to add into the off-gas system at the first stage steam jet air ejector (SJAE). Oxygen flow is controlled manually based on the reading of oxygen meters sampling outlet gas. A steady-state oxygen flow is set to maintain the oxygen concentration at the recombiner outlet between 8 and 12 volume percent. When hydrogen injection rate increases are to be made, oxygen flow is increased first to be sure there is always excess oxygen. The Dresden-2 unit had some difficulties with fires in its off-gas system during cycle 9 and the early part of cycle i0. Similar fires occurred during the same period at another plant of identical off-gas system design which operated on normal water chemistry. An in-depth study showed that the cause in both cases was most likely due to catalyst fines that had migrated upstream. During flow perturbations, the catalyst fines are apparently physically disturbed in a wa? that changes their heat transfer characteristics, and a fire is initiated. The correlation between abrupt changes in off-gas flow and initiation of fires is evident at both units, but a higher number of fires occurred at Dresden-2 because each cessation of hydrogen addition causes a flow perturbation. Recommendations on how to modify the off-gas system at Dresden-2 to preclude catalyst migration have been made. After switching from the "B" off-gas train to the "A" off-gas train in the early part of cycle 10, no more fires have occurred. Thus, the phenomenon was equipment related and not associated with HWC off-gas chemistry. 3.1.3 Plant water chemistry. Control of ionic impurities was very good in Dresden-2 throughout cycle 9. Conductivity was maintained < 0.2 uS/cm for 98% of the time, and pH was maintained in the 7 to 7.5 range. Performance in the first half of cycle i0 has been even better with conductivity below 0.I0 uS/cm for over 95% of the time. Fig. 7 shows a comparison of the two cycles.

50

W. BILANINet

al.

Ion chromatography studies of impurities in the coolant at Dresden-2 showed that under continuous operation conditions, no more than 30 ppb carbonate, 5 ppb chloride, and 5 ppb sulfate were present. The balance of the conductivity is made up of hydronium and hydroxyl ions. Thus, the species making up the conductivity at Dresden-2 are less aggressive than the sodium sulfate often used to assess conductivity effects in laboratory studies (Ruther et al., 1983). In addition to monitoring conductivity, pH, and oxygen, the HWC verification program also includes the measurement of various metals and nuclides in the feedwater and reactor water. Soluble and insoluble corrosion products have been collected continuously, with samples changed at roughly 3-day intervals. Each sample fraction has been analyzed for Fe, Cu~ Ni~ Co, Zn, Cr~ and Mn. The dominant impurity in the Dresden-2 feedwater is insoluble iron. Fig. 8 shows the concentration of insoluble iron as a function of time for a period following July 15, 1983. The 20-ppb spikes are the result of several condensate demineralizer changeouts over a short interval; the 40-ppb spike encompasses an orderly shutdown for pipe crack inspection and the restartup of the reactor. No long-term adverse consequences of hydrogen addition are evident. The time-base plots for other elements that were analyzed all show similar patterns, with no element showing a strong upward trend with time. The concentrations of the other insoluble metals are generally less that 0.i ppb.

Theoretical limit / I for pure water ~ 7 '

99.99 99.90

+.oo

sOOOlo.oo

~•>

'~

--

..,=7

Cycle~ - - ~

>" 99.00

4 3

~c~9

o~

1.00

-1 -2~

0.10

-3

OOl

I l I l'l IIII

0.01 Figure

7

I I t I I I~

-+ 1.00

0.10 Conductivity (~Slcm)

Probability plot of conductivity data for reactor water at Dresden-2.

50 40 30

~ o0

g~

~o lO

0 0

Figure 8

40

80 120 160 200 Days Since July 15, 1983

Insoluble

Iron

in

240

Feedwater

280

H y d r o g e n water chemistry for B W R s

51

In the reactor water, the dominant activation product is cobalt-60. No clear trend, either upward or downward, was observed in the first cycle on hydrogen water chemistry (cycle 9), but slightly lower-than-average concentrations were measured at the start of the second cycle, as shown in Fig. 9. The measured values of 0.1-0o2 ~Ci/L are consistent with data from other BWRs. The recently measured lower reactor water cobalt-60 may be reflecting the fuel deposits formed on the "HWC conditions only" fuel. This crud should contain a lower elemental cobalt concentration than earlier cycles because of the lower soluble elemental cobalt in the reactor water during HWC cycles (by a factor of 2-3). This effect is confirmed by fuel crud scraping analysis which shows significant lower concentrations of elemental cobalt on fuel exposed to cycle 9 only (see later). Of the soluble species that were monitored in the feedwater, cobalt is the element of greatest concern because of its activation to cobalt-60 in the reactor core and accompanying potential for radiation buildup. Fig. i0 shows the concentration of soluble cobalt in the feedwater as a function of time. The downward and stable trend of cobalt is also observed for the other elements that were analyzed. In contrast, there was no definitive upward or downward trend in the cobalt data in the reactor water; average values of 40 ppt for soluble cobalt agree with measurements from other BWRs. Because of the operational practice changes to the condensate treatment system that were implemented during cycle 9, it cannot be explicitly concluded that these trends are a result of hydrogen water chemistry. Nonetheless 9 there appear to be no adverse consequences of hydrogen water chemistry for soluble species in either the reactor water or feedwater. 1.0 [] Cycle 9 0.9

--

0.8

--

L3 0.7

--

._8 0.6

--

0.5

--

0.4

--

,9, 0.3 o

--

0.2

--

0.1

--

g 8

• Cycle 10

o

o

0.0 0

100

200 300 Days Since Beginning of Fuel Cycle

Figure

9

400

Trend in reactor water soluble cobalt.

0.06 o. 0.05

g

0.04

8

0.03

O

o 0.02 o 0.01

o

I 0

Figure 10

40

80 120 160 200 Days Since July 15, 1983

I 240

280

Trend in feedwater soluble cobalt during cycle 9.

52

W. BILANIN et al.

Summarizing the corrosion-product data, there is no indication that hydrogen water chemistry has caused any detrimental change in the soluble and insoluble corrosion-product transport in the feedwater or in the reactor water. In the feedwater, spikes of insoluble substances are common to all BWRs, and they are expected to propagate to the reactor water. The generally downward trend in feedwater solubles probably is the result of a gradual implementation of good operational practices centered around the management of the condensate treatment system. The concentrations of the impurities in the feedwater and reactor water are consistent with data from other plants having deep bed demineralizers. 3.1.4 Radiation buildup. Both shutdown dose rate surveys and gamma scanning measurements were made at various locations in the drywell at selected times during cycles 9 and i0. Fig. Ii shows that drywell dose rates increased moderately during cycle 9, although comparisons with other plants show that the Dresden-2 dose rate is well within the experience band of other operating plants. Part of the overall increase is possibly due to the increase in soluble cobalt-60 in the reactor water. The gamma scanning results show that the cobalt-60 incorporated into the corrosion film on reactor piping increased by 14% during cycle 9 but the rate of increase appears to be tapering off. The contribution to area dose rates from crud-related local hot spots has increased significantly. Analysis of the dose rate and gamma scan data indicates that this behavior is probably due to sloughing of a portion of the corrosion film from the piping surfaces. Since a thinner film thickness is known to be stable under HWC conditions, this behavior is not surprising. The Dresden-2 unit recirculation system was decontaminated at the end of fuel cycle 9, causing a discontinuity in the radiation buildup behavior. Further sets of dose rate and gamma scan measurements were taken in cycle i0 (Fig. 12). The results at Dresden-2 show similar recontamination rates to those observed at other plants after decontamination, as shown in Fig. 13. The cobalt-60 buildup behavior is consistent with laboratory loop data. For instance, the modest increase on "old film" exposed to HWC conditions would be predicted from loop experiments (Fig. 14), as would the relatively low rate of recontamination of clean surfaces (Fig. 15) (Lin, 1985).

• Nine Mile Point 1 ® Oskarshamn-1 41, Oskarshamn-2 O Browns Ferry-1 • Browns Ferry-2 ¢ 8arseback-1 ® Quad Cities-1 0 Quad Cities-2 [] Brunswick-2 .,1 Tsuruga ,1 Hatch-2 O VermontYankee

• A V • 0

Dresden-2 Dresden-3 Millstone-1 Monticello Funushima-1 Funushima-3 b, Ringhals-1 • Hamaoka-1 • Shimane • NNM r~ Pilgrim • Hatch-1

~" 1000 n- 900 E 800

% % ° °®

700 o a

600 500 f

0

•

®

¢

o

Start of HWC

~_ 4OO ._~ 300

,•

_

x7

v

v

-~ 200 •~ rr

100 0 0

1

Figure ]1

2

3 4 5 6 7 Operating Time (effective full power years) BWR radiation levels showing the small impact of implementing HWC at Dresden-2.

8

9

Hydrogen water chemistry for BWRs

53

26 A--suction A--discharge B--discharge

24 22 d " 20 E 18

m

16 >, 14

b

>

= 12-< 10-O t.O

& (5

Cycle ~ ~9

8--

~ Cycle 10

6-4-2--

I

0

7.0

I

I

7.4

I

I

7.8

8.2 8.6 9.0 Effective Full Power Years

Figure 12.

9.4

9.8

Dresden-2 recirculation system pipe gamma surveys.

500

t-

~" 400 E rr

_

• o • • •

Dresden-2 Nine Mile Pt 1 Vermont Yankee Pilgrim Monticello

300 --

h- 200

_g

"~ 100 n-

I

1

2.0

2.4

O

0.0

0.4

0.8 1.2 1.6 Effective Full Power Years

Figure 13.

Recontamination of BWR piping.

W . BILANIN et al.

54

1.5

Water chemistry: Test sample: Test No.: (sample ID):

E

o

g

NWC-Ref.~HWC-Ref. 304 prefilmed; 500-h, NWC A/1~C/2 (A8) A ~ a B/1~C/2 (B16) l l ~ ©

1.0

c o

o

0.5

Normal water ~ chemistry (NWC)

_

----> HWC

/k

Go o o

o

1 0

1000

2000

Figure 14.

3000 4000 Exposure Time (h)

6000

5000

Laboratory loop tests of Cobalt-60 deposition on Type-304 stainless steel.

3.1.5 Radiation fields during plant operation. An important concern associated with HWC is the impact on radiation levels in the turbine building, plant environs, and off site. A plot of the recorded values of the main steam line monitor activity is shown in Fig. 16 for various power levels. A fivefold increase between normal water and HWC conditions is seen at full power. The increase in N-16 steam activity is due to chemical changes that occur in the core with hydrogen addition. The N-16 isotope (7-sec half life) is formed by an (n,p) reaction with 0-16 in the reactor water. Under normal oxidizing chemistry conditions, the majority of the N-16 probably exists as nitrate (NO3-), which is nonvolatile. With the more reducing core chemistry conditions of HWC, a greater fraction of the N-16 is in volatile compounds (an~nonia and nitrogen oxides) which partition into the steam phase. (The chemical concentrations of N-16 compounds are at less than the part per trillion level, so N-16 chemistry is only of radiological interest.) The impact of this increase is expected to be plant specific, as discussed in section 4.

2.0 1.8 1.6 1.4 o 1.2

7O

1.0

n

0.8

0

0.6 0.4 0.2

1

0.0

0

1 0.4

I

1 0.8

1

1 1.2

1

1 1.6

1

1 2.0

Exposure Time (103 h) Figure 15

Cobalt deposition on Type-316 SS in normal water chemistry and hydrogen water chemistry.

I

I 24

H y d r o g e n w a t e r chemistry for B W R s

3000

f

55

g~

~2000 U rr = 1000

?,

-

~

~j~ ~/ ~//

--

o

I

[

0.5 Figure

16

Experi~l data z~ Steam line radiation field at 95o/0 power O Steam line radiation field at 830/o power [] Steam line 16N activity at 83o/0 power (scale x ® Recirculation loop radiation field at 83°/0 power

1.0 1.5 2.0 H2 Concentration in Feedwater (ppm)

25

Variation of radiation fields at Dresden-2 with hydrogen addition rate.

Detailed field measurements have been made in and around Dresden-2. The results indicate that although the impact of HWC is significant and measurable in the vicinity of the turbine building, it diminishes rapidly away from the plant and is negligible offsite. Fig. 17 indicates the contribution of HWC to the total dose rate on a series of contours drawn from the turbine building. At Dresden-2, operation on HWC adds about i0 man-rem to the 2,000 man-rem currently incurred each year. However, in terms of exposure for IGSCC-related repairs, HWC is ALARA effective. For example, pipe replacement programs typically lead to an exposure of i000 to 2000 man-rem.

Figure 17

Net HWC contribution to Dresden environs dose rates (~RIh).

W. BII_ANIN et al.

50

3.2

In-plant materials test program

3.2.1 Slow Strain Rate Tests. In order to verify the beneficial effects of hydrogen water chemistry on stress corrosion cracking, slow strain rate (SSR) tests were carried out at the Dresden-2 site on various structural materials (Gordon et al., 1985; Davis et al., 1983; Cowan et al., 1985). The tests were performed in autoclaves at 274°C with the influent supplied from the reactor recirculation loop water system. Fig. 18 shows the stress versus time behavior of sensitized Type 304 stainless steel during SSR testing at several oxygen levels. It is clear from these graphs that as the oxygen level decreases, the ductility and maximum stress level increase, thus indicating the beneficial effect of low oxygen levels for suppression of IGSCC in Type 304. Table i lists the results of all the SSR tests performed at Dresden-2. For Type 304, an oxygen level of <20 ppb is needed in order to eliminate ICSCC--the Type 304 sample tested at 40 ppb (test #i-82) exhibited 35% IGSCC on the fracture surface. The results also indicate that short interruptions in the injection of hydrogen, leading to temporary increases of the oxygen concentration in the water, are not necessarily damaging with regard to IGSCC. For example, test #4-84 experienced a total time of 30.3 hours during which the oxygen level was >30 ppb and no IGSCC was observed. During that period a single oxygen excursion to >200 ppb for 10.4 hours was noted. However, some ICSCC did occur in test #2-83 which was out of the HWC regime continuously for 15 hours. This indicates that between 10.4 and 15 hours is the maximum time that a plant could be off HWC and still not initiate IGSCC in the piping. This so-called "memory effect," although applicable to IGSCC initiation, does not seem to be related to the corrosion potential, i.e., the corrosion potential follows the oxygen concentration in the water. This point will be discussed in more detail in a later section, entitled "Corrosion Potential Measurements." In addition to testing smooth Type 304 SSR specimens, one precracked specimen (test #8-84) was also tested under HWC conditions. The precracking was carried out in the laboratory by slowly straining the specimen in 200-ppb oxygen water for 67 hours (conditions which had generated IGSCC in earlier specimens). The specimen was then shipped to the Dresden-2 site for further SSR testing in HWC with results shown in Table I. Subsequent examination of the fracture surface revealed that the cracks that had initiated during the 200-ppb oxygen water exposure did not propagate in the HWC environment and that the final fracture was completely ductile. It should also be noted that during the 301 hours of in-plant testing, 10.4 hours were in water with an oxygen level of 30-200 ppb. These results suggest that existing cracks in reactor piping will not propagate under HWC conditions, and that short excursions from the HWC regime will not exacerbate existing SCC damage.

100

80

60

I

< 20 ppb oxygen

e

40

20

0

100

200 Test Time (h)

Figure

18

Stress versus test time curves for several sensitized Type-304 SSR specimens tested at various oxygen levels at Dresden-2.

300

304

304

304

304

304

304

304

A533-B

A533-B

A508-8

AI06-B

1-82

2-82

1-83

4-84

3-83

2-83

8-845

5-82

3-82

7-84

5-84

270

8-15

12-18

20

150-280

2-20

5-23

3-30

3-19

5-20

10-20

35-45

0.12

0.09

0.29

0.29

0.09

0.17

0.13

0.09

0.19

0.29

0.37

0.29

in 200 ppb 02 water.

7Test t e r m i n a t e d due to system breakdown.

6This Eime i n c l u d e s 67 hours f o r p r e c r a c k i n g .

5Specimen was precracked

4Test terminated due to thermal overload,

31GSCC = Intergranular Stress Corrosion Cracking TGSCC = Transgranular Stress Corrosion Cracking

2Type 304 was furnace sensitized at 621°C (1150°F) for 24 hours.

and AI06.

.

.

24

12

44

--

45

46

.

38

20

12

.

.

.

Elongation (%)

.

.

.

.

.

.

.

.

.

0

.

.

.

0.5/1.0

0

6.4/24.5

3.6/9.3

0

129

.

Max/Tot

.

.

.

.

. 0

.

.

. .

.

37

7.9/9.4

15/15

7.6/24.5

10.4/21.0

3.0/4.8

.

Max/Tot

3 0 - 2 0 0 ppb >200 ppb

2

0

37

10.4

15

49

30.3

5

4.8

129

Total

Ductile Fracture

Ductile Fracture

Ductile Fracture

40% TGSCC

No ICSCC Propagation

Minor IGSCC

Ductile Fracture

Ductile Fracture

No ICSCC

Ductile Fracture

35% IGSCC

70% ICSCC

Results 3

(0.001 in/h) for Type 304 and 2.1 x 10 -5 mm/s (0.003 in/h) for alloys A533, A508,

94

577

63

37

3686

1814

396

400

2084

297

143

108

Time to F a i l u r e (h)

02 E x c u r s i o n s (h)

Results of Dresden-2 reactor slow strain rate tests .±

Conductivity a t 25°C ()S/cm)

iExtension rates were 7.1 x 10 -6 mm/s

304

4-82

Test #

Nominal O Level M a t e r i a l 2 2(ppb)

Table i.

,,<

r~

G.

58

W. BILANINet

al.

SSR tests were also carried out on several low alloy and carbon steel samples (Table i). All the specimens tested in HWC failed in a ductile manner whereas one A533 specimen (test #5-82) that was tested in normal chemistry water exhibited transgranular cracking on the fracture surface. 3.2.2 Crack growth testing. A reversing dc potential drop technique developed by Ceneral Electric has been used to monitor crack growth in a sensitized Type 304 specimen under HWC conditions at the Dresden-2 plant (Cowan et al.). Initial precracking of the specimen was carried out in the laboratory in 200 ppb oxygen water. The specimen was then shipped to the Dresden-2 site where it was exposed to the reactor recirculation loop water at 274°C. During testing, the specimen was loaded to K I = 27.5 MPa/m (25 ksi£in) which is approximately equal to the stress intensity associated with a crack detected in a safe end at Dresden-2. The cumulative crack growth as a function of exposure time is shown in figure 19. No significant crack growth occurred even though two reactor scrams and several hydrogen injection interruptions took place during the test. This result is consistent with the precracked SSR test results discussed in the previous section. It is also consistent with mid-cycle and end-of-cycle in-service inspections conducted by Commonwealth Edison on several recirculation system riser and safe end welds. No changes in the ultrasonic signatures of a number of indications present prior to cycle 9 were observed, indicating that no crack enlargement had taken place. The crack growth results are compared in Fig. 19 to the crack growth of a similar test specimen exposed to 200-ppb oxygen water under laboratory conditions. The conductivity of this environment ranged from 0.3-0.7 ~S/cm as compared to 0.1-0.3 pS/cm at Dresden-2. The 30-50 fold higher growth rate in the laboratory test probably is due to the combined effects of the differences of conductivity and oxygen concentration and not just oxygen concentration alone, but figure 19 provides a graphic illustration of the large influence of water chemistry on the rate of IGSCC. 3.2.3 Corrosion potential measurements. T e corrosion potential attained by a metal in a water environment results from the electrochemical corrosion reactions occurring at its surface, and these corrosion potential values are often correlated with the types of corrosion that can occur. For the case of stainless steel in high-purity water at about 300°C~ it has been found in laboratory tests that IGSCC of sensitized stainless steel occurs only when the corrosion potential of the stainless steel is more positive than about -0.23 volts (on the standard hydrogen scale or SHE). This value is frequently referred to as the protection potential against IGSCC.

0.780

0.760

• 0.740

~

--

f

f

--

~ 0.720

-

5 m/h (149-245 x 10"3 in/yr) -

~ . Dresden-2 HWC da/dt = 5.4 x 10-7 in/h*

0.700 *Excluding interruptions

I

0.680

0.4

0

Figure

19

I

f

I

0.8 1.2 1.6 On-Line Test Time (h x 103) Crack length versus time curves for a sensitized Type-304 compact tension specimen tested at the Dresden-2 reactor during HWC operation.

1 20

Hydrogen water chemistry for B W R s

59

Fig. 20 shows the correlation between corrosion potentials and percent IGSCC observed on fracture surfaces of sensitized Type 304 specimens fractured in SSR tests in the laboratory (open points) and at Dresden-2 (filled points). The results show that ICSCC occurred only at corrosion potentials above -0.23 volts not only in the laboratory tests but also in tests performed at Dresden-2. Thus~ these tests performed at Dresden corroborate the laboratory results and show that the protection potential in the reactor recirculation loop water has the same value (-0.23 volts). These results provide the fundamental basis for the HWC countermeasure for pipe cracking, namely, sufficient hydrogen addition to decrease the corrosion potential of stainless steel below the protection potential~ together with very low conductivity water. During hydrogen injection periods at Dresden-2, it is occasionally necessary to discontinue hydrogen addition (to decrease steam line activities). When hydrogen addition is stopped the corrosion potential of the stainless steel electrode increases and rises above the protection potential. This is illustrated in Fig. 21 which shows the corrosion potential and some excursions when hydrogen was stopped during SSR test #4-84. In spite of the excursions of corrosion potential above the protection potential, the test specimen showed no IGSCC (see earlier section on SSR tests). 3.3

Fuel surveillance

program

3.3.1 Introduction. Means for mitigating [CSCC in stainless steel reeirculation piping can be considered successful only if the proposed remedy does not represent a threat to the integrity or satisfactory performance of the other key components of the system. Hydrogen addition to the water of BWRs creates a water chemistry in the reactor core that is outside the experience base of either BWRs or PWRs. Because the corrosion behavior of the zirconium-based alloys (Zircaloys), which are used almost exclusively in the high-flux region of water-cooled reactors, depends strongly on the reactor water chemistry, it is necessary to verify that the integrity of the fuel components is not adversely affected by the presence of added hydrogen. In particular, the corrosion of the cladding tubes that separate the coolant water from the nuclear fuel and the embrittlement of the structural components that keep the fuel rods evenly spaced from one another are key performance considerations (Franklin et al., 1985). In most projected regimes of corrosion behavior under HWC operation conditions, the effects of the added hydrogen are expected to be small. However, for some postulated regimes, such as for Zircaloy-2 fuel cladding experiencing BWR water chemistry oxidation

100 90

I

/

Protection potential

--

D/ /

80 7O - co

/

/

No IGSCC<

•

60

1

/ /

E

50

~

40

rn/ / •

/

30 /

20 10

u

/

// --

/

/

rn

rn

[]

Key • Dresden-2 tests D Laboratory tests

D

(6 tests) /

.~1 -0.4

-0.2

DI

I

I

0

Corrosion Potential (Vsh,) Figure

20

Corrosion potential versus percent IGSCC curve for sensitized Type-304 tested using the SSR technique.

0.2

W. BILANIN et al.

60

- 0.06 -0.10

-

reduction

-

Plant power

--~ - 0 . 1 4 -0.18 uJ _ 0 . 2 6

~

,~---,,,~"~ ~

~

~,,,,,k A

Protect!on

~

r~

]

[1 -

- 0.30 - 0.34 320 280 Q240 200 120 160 40 8o

L 0

~

-

0

100

200

L.... ~

300

400

Test Time (h)

Figure

21

ECP and 02 concentration versus test time for sensitized Type-304 specimen 4-84 (see table 4-1) tested at Dresden-2 using the SSR technique.

rates and PWR water chemistry hydrogen pickup rates, the loss of cladding ductility due to hydrogen embrittlement could be severe. Another area of uncertainty was determined (Johnson, 1983; Cox, 1983) to be related to the hydriding of those fuel elements that would be first exposed to a normal water chemistry environment for one or two reactor cycles before being exposed to an HWC environment. 3.3.2 Fuel surveillance program objective. The objective of the fuel surveillance program is to detect, identify, and assess any material characteristic that could degrade the performance of fuel assemblies operated in commercial BWRs in which hydrogen is intentionally added to the reactor feedwater. Of particular interest, are the effects of hydrogen on oxidation and hydriding of the Zircaloy components, as well as on crud deposition rates and characteristics. 3.3.3 Program description. Four lead test assemblies (LTAs) of carefully characterized fuel components were inserted into the Dresden-2 reactor of Commonwealth Edison Co. at the beginning of the first cycle of hydrogen addition (cycle 9). The LTAs include Zircaloy-2 and -4 with precharacterized ranges of corrosion behavior that are representative of those usually found with cladding batches used in BWR fuel fabricated during the past few years. The precharacterization of the Zircaloy materials was made on the basis of a hightemperature, high-pressure test whose results tend to correlate well with in-reactor corrosion performance (Cheng et al., 1985). In addition, three fuel bundles that have been or will be exposed to various combinations of normal and hydrogen water chemistries have been selected for postirradiation examinations. All three bundles are made of fuel components for which no or few precharacterization data exist; the basis for the selection lies in their power histories, which are similar and among the most severe for the Dresden-2 reactor. Table 2 outlines

the ongoing fuel surveillance program.

ltydrogen water chemistry lor BWRs T a b l e 2.

Outage/ End of Cycle

Fuel

BWR/HWC Cycles

3/0

8

(Cycles 9

0/i (LTA) (Cycle 9)

I/2

None

2 U, 1 G d

1 Bundle Visual Oxide Crud

12 U, 4 Cd 12 U, 4 Cd 4 U, 2 Cd

1 Bundle Visual Oxide Crud

12 U, 4 Cd 12 U, 4 Cd 4 U, 2 Gd

2 U, 1 C d

1 Bundle Visual 8 U, 4 Cd Oxide 8 U, 4 Cd Crud 4 U, 2 Cd

2 U, i Cd 2 Water Rods 7 Spacers

Same as 3/0 .........................

8,9,10)

0/2 (LTA) (Cycles 9,10) ii

Hot Cell Examination

Site Examination

None

2/i (Cycles 7,8,2)

(Cycles

program.

6,7,8)

3/0

i0

surveillance

61

0/3 (LTA) 9,10,11)

Same as 0/i (LTA) ...................

Same as 0/i (LTA) ...................

(Cycles

A combination •

provide

for the following

activities:

Site examination -----

•

of site and hot cell examinations

Visual assessment of the overall bundle condition. Visual determination of nodular corrosion coverage. Eddy-current determination of oxide thickness. Crud scraping.

Hot cell examination -----

Metallographic Metallographic Hot extraction Crud analysis.

determination determination determination

of oxide thickness and nodular of hydrogen distribution. of hydrogen content.

oxide coverage.

Although significant HWC-induced changes in the corrosion behavior of the Zircaloy would be apparent during the site examinations, information on Zircaloy hydriding can not be properly quantified except by destructive examination in hot cells. Components examined include urania (denoted "U" in Table 2) and urania-gadolinia (denoted "Cd" in Table 2) fuel rod claddings (Zircaloy-2), spacers (Zircaloy-4), and water rods (Zircaloy-2). As shown in Table 2, no examinations were conducted prior to the start of cycle 9, at which time hydrogen was first added; however, a three-cycle bundle discharged at the end of cycle 8 was selected as the reference bundle (designated as 3/0, i.e., three cycles in normal water chemistry and none in HWC). The latter was examined at the end of cycle 9 together with a 2/I discharged bundle, and one of the four 0/i LTAs. At the end of

62

W. BILANIN et al. cycle i0, another discharged bundle (1/2) and one of the four LTAs (0/2) will be examined. At the end of cycle Ii, corresponding to three cycles of hydrogen water chemistry, the four (0/3) LTAs will be available for detailed examination. 3.3.4 Results after one cycle of hydrogen addition. During the first cycle with hydrogen addition, the oxygen concentration in the recirculation piping was kept below 20 ppb during 80% of the time the reactor was operated at more than 25% of nominal full power. Therefore, the fuel was exposed to a well-established HWC environment for about 80% of the time, while during the remaining 20% of the time, the actual reactor water chemistry varied between the limits defining normal and hydrogen water chemistries. As presented in Table 3, the 3/0 reference bundle, a 2/1 bundle, and one of the LTAs were first examined nondestructively at the reactor site. On the basis of the poolside examination, a number of fueled and unfueled components were selected, shipped, and destructively examined in hot cells. Oxidation. In a typical BWR environment, oxidation of Zircaloy produces (i) a thin layer of zirconia (ZrO2) that uniformly covers the Zircaloy surfaces in contact with water, and (ii) "nodules" that result in locally thicker patches of zirconia (Figs. 22 and 23). As exposure time increases, the nodules may grow and coalesce, resulting in uniformly thick oxide layers. Present results (Cheng et al., 1985; Cheng and Blood, 1986) indicate that the overall amount of oxide is, as expected, significantly lower on the LTA rods than it is on the three-cycle rods. In particular, measured uniform oxide thicknesses are only i-3 ~m on the LTA rods, whereas they are 8-12 um on the three-cycle rods. Hydriding. In a typical BWR environment and with good fuel manufacturing hydrogen pickup by the cladding mostly results from the oxidation reaction:

practices,

Zr + 2H20 + ZrO 2 + 2(l-x)H 2 + 4xH, that is: zirconium (main constituent of Zircaloy) + water ÷ zirconium oxide + hydrogen released to the coolant + hydrogen retained in the Zircaloy. The hydrogen pickup denoted by "x" is about 0.I for Zircaloy-2 and -4 under typical BWR conditions.

Table 3.

Poolside

inspection at Dresden-2

at the end of cycle 9.

Hydrosen Exposure Condition

2/i

o/i

1

i

I

30.3

32.7

8.4

Fuel rods inspected (Visual, oxide thickness)

16

16

13

Fuel rods scraped for crud

6

6

8

3 (1)

3(1)

3(i)

3/0 Bundles

inspected

Bundle exposure GWd/MT

Fuel rods removed Bundle hardware

removed

Notes: (I)I urania-gadolinia

rod: 2 urania rods.

(2)2 water rods; 7 spacers.

(2)

Hydrogen water chemistry for BWRs

Nodular oxide 20 #m

Figure

Figure

22

23

Photomicrograph showing the uniform and nodular oxide formed on one of the LTA spacers.

Appearance of a UO2 rod from the LTA (rod diameter: -12.55 mm) at lO0-in from the bottom of the rod; oxide nodules are readily visible as white spots.

Present results (Cheng and Blood, 1986) indicate that hydrogen levels in all the Zircaloy components examined at the end of cycle 9 are low and correlate well with the overall extent of oxidation. Therefore, the presence of hydrogen in the coolant has not resulted in a detectable increase in hydrogen pickup by the Zircaloy material. Crud Deposition. Excessive crud deposition accompanied by the formation of tenacious crud deposits can be very detrimental to the performance of the Zircaloy cladding (Marlowe et al., 1985). Present results

(Martin et al., 1985) yield the following observations:

The iron surface concentration in the crud deposits represents more than 95% of the total surface concentration of all metallic elements detected in the deposits for all three bundles; The total measured crud deposits, and the distribution and magnitude of cobalt-60 deposition are roughly equivalent for the 3/0 and 2/i bundles; however, the 2/I bundle compared to the 3/0 bundle has a higher percentage of tightly adherent deposits and higher concentrations of copper in the tightly adherent deposits; The loosely and the tightly adherent deposits in bundle 0/I are lower by factors of two to four than those on the two three-cycle bundles; also the copper, nickel, cobalt, zinc, and manganese surface concentrations are much lower in the 0/i bundle than they are in to the three-cycle bundles. Overall, crud deposit characteristics are in the normal indications of any adverse impact on fuel performance. 4.

PROCRESS TOWARD IMPLEMENTATION

range, and there are no

OF HWC AT OTHER U.S. BWRs

Based on the favorable results obtained in the verification program at Dresden-2 during Cycle 9, interest has been growing in implementing hydrogen water chemistry at other BWRs in the United States (Cowan et al.~ 1986). This section describes activities that have been undertaken to facilitate more widespread adoption of this new technology.

63

64

W. B[LANIN et al. 4.1

Assessment of the impact of radiation field increases associated with HWC

As mentioned in Section 3, the use of hydrogen water chemistry increases the N-16 radiation fields in and around the plant. As part of the overall HWC study, the impact of the field increases on the radiation dose to members of the general public and plant personnel in restricted and unrestricted areas was assessed for 19 BWR sites in the United States (Burley and Anstine, 1986). Computer model projections and dose rate measurement data from Dresden-2 were used to develop an analytical technique for using site-specific data provided by the utilities. Predictions of off- and on-site dose rates were made for site configuration. Major input data for each site included the amount of shielding, and the steam activity inventory on distance and orientation of key facilities from the turbine, shielding between these facilities and the turbine.

each turbine building and the arrangement of equipment, the turbine deck, the and the supplemental

Comparisons of predicted dose rates with measured data from several sites showed that, with accurate input data, the analytical technique is capable of calculating dose rates within a factor of 2. On the basis of predicted dose rates, the study projected that: Radiation dose rates at nine sites would be well below the regulatory limits for the three groups of concern--the genera[ public, personnel in restricted areas, and those in unrestricted areas of the site. At five BWR sites, dose rates would exceed 50% of the regulatory limit for at least one group. At these sites, selected mitigating actions to reduce exposure may be desirable to accommodate operation with HWC. At the remaining five sites, dose rates would exceed the regulatory limit for one group and approach it for another. Mitigating actions would be necessary for these sites if HWC was implemented. The major cause of the projected higherthan-desired dose rates was lack of adequate turbine-deck shielding. The radiological effect of implementing HWC should be minima[ at most BWR sites and manageable at others, where the addition of turbine-deck shielding or the relocation of temporary buildings should keep dose rates within regulatory limits. It should be noted that this study provides only scoping evaluations of radio[ogical effects, and quantifying the effects requires actual measurements of dose rates during short-term HWC mini-tests such as those discussed below. 4.2

Analysis of the results of HWC minitests at U.S. plants

Hydrogen water chemistry minitests are performed primarily to determine the hydrogen addition rate required to suppress IGSCC in recirculation lines and to assess the steam line radiation levels associated with that hydrogen addition. Minitests usually last for a few days during which hydrogen is injected into the feedwater at various rates, and measurements are made of the oxygen concentrations and corrosion potentials of stainless steel in the recirculation system water and of steam line radiation levels (and dose rates at many site and off-site locations). In some minitests, slow strain rate tests are made with sensitized stainless steel specimens in recirculation water to confirm that IGSCC is suppressed. To date, minitests have been performed in five U.S. BWRs in addition to Dresden-2, namely, Peach Bottom-3, Pilgrim, FitzPatrick, Duane Arnold, and Hatch-l. In order to perform a minitest, the plant must first purify the reactor water to the point that conductivity levels are less than 0.3 ~S/cm. (As discussed earlier, suppression of [CSCC is achieved only in low conductivity water even with hydrogen injection.) In the plants tested, the reactor water conductivity levels were 0.3 ~S/cm or less before tests were started. In each plant, conductivities decreased further when hydrogen was added. Oxygen dissolved in the water in recirculation lines is an important factor for IGSCC in the pipes. The aim of hydrogen addition is to suppress the dissolved oxygen content to low levels. Oxygen concentrations measured in recirculation water lines as a function of hydrogen concentrations in the feedwater (i.e., a function of H 2 addition rate) are shown

Hydrogen water chemistry for BWRs

65

in Fig. 24. Hydrogen addition decreases the recirculation oxygen in all the plants, as desired; however, the amount of decrease is very plant-dependent. The corrosion potential of stainless steel is felt to be the primary indicator of the propensity for IGSCC in the water. Laboratory and plant tests show that IGSCC does not occur if the corrosion potential is below -0.23 volts (SHE) in high-temperature water of conductivity less than 0.3 uS/cm (see Fig. 20). In all the minitests, moderate hydrogen addition could depress the corrosion potentials below the critical value (-0.23 volts). The relative degree of decrease of corrosion potentials in the various plants is roughly parallel to the decrease of dissolved oxygen, which is in accord with laboratory information about the relationship between corrosion potential and dissolved oxygen. Several series of in-reactor materials tests have overseas BWRs. The results are consistent with those IGSCC of sensitized stainless steel is prevented when 230 mV SHE and the reactor water conductivity is <0.3

been performed in domestic and of laboratory tests and confirm that the stainless steel ECP is
Fig. 25 shows the results of constant extension rate tests (CERT) at five plants. These tests were performed in autoclaves fed with recirculation system water during plant power operation. The reactor water conductivity was <0.3 pS/cm in each case. The test results are presented as plots of the average IGSCC propagation rate measured in a CERT on a sensitized Type 304 stainless steel specimen (i.e., IGSCC depth divided by test duration) versus the stainless steel ECP. The resolution of the test technique is about I0 -° mm/s so tests in which no IGSCC was observed are plotted at that rate. It is apparent that IGSCC is stopped below -230 mV SHE, in agreement with the laboratory results summarized earlier. Fig. 25 shows the relationships between stainless steel ECP and hydrogen addition rate observed in short-term tests at six plants. The ECP measurements plotted in this figure were made during full-power plant operation in autoclaves fed by recirculation system water. It is apparent that the stainless steel ECP decreases as more hydrogen is added to the feedwater and that values well below -230 mV SHE are achievable. It is also apparent that each plant behaves differently. These plant-to-plant differences arise as a result of the indirect way in which the added hydrogen affects the ECP.

Recirc Oxygen Concentration (ppb)

100.0

__•••

• PB-3 • PIL

10.0

1.0

1

I

PI'I

I-

r

0.2

0.6 1.0 1.4 1.8 2.2 Feedwater Hydrogen Concentration (ppm)

Figure 24

Recirculation line oxygen concentration as a function of feedwater hydrogen concentration for four U.S. plants. (The arrows indicate addition rates that depressed the stainless steel corrosion potential below the IGSCC protection potential.)

26

66

W. BILANIN et al. The ECP of stainless steel in the recirculation system is determined principally by the local concentrations of the oxidizing impurities (02, H202 and others) originally formed by radiolysis of water in the core. Equilibrium is not achieved in BWR coolant circuits and the concentrations of these substances in the recirculation system will vary from plant to plant (whether or not hydrogen is added) because of design differences that affect the rates ofproduction in the core and the rates of decomposition and recombination as the coolant circulates from the core exit to the recirculation system. Because decomposition and recombination rates are slow in the absence of radiation, the design differences of principal importance are believed to be those which affect the radiation exposure of the coolant as it passes through the annulus between the core shroud and the vessel. These differences include: i.

Core power dentify (affects the radiation field in the annulus),

2.

Annulus physical dimensions annulus), and

3.

Coolant flow rate through the annulus (affects the total radiation dose experienced by a given amount of coolant).

(affects the degree of self-shielding

in the

The addition to the feedwater of a reducing agent like hydrogen is expected both to reduce the net rate of production of oxidizing species in the core and to increase the rates of decomposition and recombination in the annulus. Thus, hydrogen addition should reduce the concentrations of oxidizing species in the recirculation system and, consequently~ reduce the stainless steel ECP. However, the magnitude of the effect observed will depend on the design variables listed above and is therefore expected to be plant-specific. At present, the observation that ECP decreases more rapidly with hydrogen injection at Fitzpatrick than at Pilgrim (Fig. 25) can be qualitatively understood but quantitative predictions are not yet possible. Consequently, the hydrogen addition rate required to reach the desired ECP of <-230 mV SHE must be determined empirically for each plant. It should be noted that the hydrogen addition rate required to reach a particular ECP is a function of both power and core flow and that cycle-to-cycle variations have been observed in long-term tests. Accordingly, frequent (or, preferably, continuous) monitoring of the ECP is necessary to assure that IGSCC immunity conditions are maintained in the recirculation system. The short-term hydrogen addition test at Hatch-i deserves special mention because it is the only case to date in which the stainless steel ECP could not be maintained below -230 mV SHE for a sustained period. In the Hatch-i test, hydrogen injection initially reduced the stainless steel ECP to values below -230 mV SHE but the ECP subsequently drifted back to higher values and became insensitive to hydrogen injection. This result is thought to reflect the oxidizing effect of the significant concentration of soluble copper (~25-30 ppb) in the reactor water at Hatch-l. It has been shown in loop tests that the presence of 30 ppb Cu ++ can shift stainless steel ECPs upwards by as much as 250 mV. As this is a significant shift compared to the downward shift achievable by hydrogen injection (Fig. 25), it is clear that special attention to the control of reactor water copper will be required in plants with copper alloy condensers and filter/demineralizer cleanup systems to ensure that stainless ECPs are maintained in the IGSCC immunity range during hydrogen water chemistry operation. Hydrogen addition to feedwater causes increased gamma radiation in steam lines, which leads to higher site radiation levels. As part of a minitest, radiation levels are measured at many locations. Steam line radiation levels as a function of hydrogen added are shown in Fig. 26. (Because of the very short half-life of the gamma radiation (~7 see) and because of shielding effects, these values are very dependent on measurement location.) The increase of the relative radiation in the steam line at the measurement point is also plant-specific but in a different way from the other changes mentioned above. Other related water chemistry quantities were measured in each minitest and found to be affected by hydrogen addition. The changes observed tend to be plant-specific and are not discussed further here. A general model relating plant behavior to hydrogen addition is not yet available.

H y d r o g e n w a t e r c h e m i s t r y for B W R s

67

0.2 0.1 0 >

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Figure 25. Stainless steel corrosion potentials in the reactor water at six BWRs as a function of feedwater hydrogen concentration.

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Figure 26. Steam line radioactivity observed in hydrogen mini-tests and at Dresden-2, (The arrows indicate addition rates that depresssed the stainless steel corrosion potential below the IGSCC protection potential.)

68

W. BILANIN et al.

For BWRs characterized by the presence of high concentrations of copper in the reactor water, a key issue is the impact of the HWC environment on the formation and deposition on the fuel elements of adherent copper-rich cruds. As expected from thermodynamic data and suggested by the results from several tests of short duration (Oskarshamn, 1985; Peach Bottom-l, 1985; Hatch-l, 1986), the less oxidizing environment promoted by the addition of hydrogen favors the reduction of the soluble copper ions into species having a lower solubility. The latter have a greater tendency to deposition; therefore, this effect is potential]y detrimental to fuel performance. However, the addition of hydrogen also results in lower oxygen concentrations throughout the BWR system, and therefore to a reduction of the general corrosion rate of copper alloys. This could decrease the input of copper to the reactor core, which is potentially beneficial to fuel performance. The overall impact on fuel performance will depend on the relative magnitude of the two effects: reduced copper input versus greater deposition efficiency. The dominant process has yet to be identified. In summary, the tests of hydrogen addition reported all indicate that hydrogen addition can achieve conditions that suppress ICSCC in sensitized stainless steel, but the required addition rates are plant-specific. Further, steam line radiation levels are increased due to the hydrogen addition rates required for IGSCC suppression but to different degrees for the various plants. 4.3

Generic guidelines

for permanent HWC installations

As the success and benefits of HWC have continued to be demonstrated, attention has begun to focus on implementation of permanent HWC installations. By early 1985, Commonwealth Edison was finalizing the design for a permanent HWC installation at Dresden-2. At about the same time, other utilities began considering such permanent installations. Since early studies indicated significant quantities of hydrogen could be required~ BWR utilities recognized the need to evaluate the effects of hydrogen storage. Therefore, the BWR Owners Group for ICSCC Research implemented a program to develop generic guidelines for design, construction, and operation of permanent BWR HWC installations. This program was begun in May 1985 and the guidelines document was formally submitted to the NRC in October 1985 (Guidelines, 1986). This guidelines document is primarily for utility use but it also provides the NRC with a "standard" for use in reviewing HWC installations. The generic guidelines are intended to streamline the implementation process by providing the basis for implementing permanent HWC installations without prior NRC case-by-case review. The guidelines were prepared by a group of people who had broad experience in all the key areas. This group consisted of representatives of EPRI, several utilities, the NSSS supplier, an architectengineer, and a producer/supplier of hydrogen. The document itself is organized in the general format of an ANS Standard. The scope of the document includes the currently available on-site hydrogen and oxygen supply options (i.e., compressed gas, cryogenic liquid, and electrolytic generation) and the delivery system design and controls. Included are guidelines for design, operation, maintenance~ surveillance~ radiation protection, and testing to provide for safe system and plant operation. Conservative methods are provided for considering the hazards of hydrogen and oxygen storage. For example, the guidelines specify separation distances between liquid hydrogen storage systems and safety-related structures which are far in excess of those specified in other related industrial standards (such as the National Fire Protection Association) or by nuclear insurance companies. The document specifically excludes addressing the various methods for verifying the effectiveness of HWC as a remedy for IGSCC and the licensing credit for HWC (e.g., reduced in-service inspection). These issues are to be addressed on a plant-unique basis. This thorough treatment in the guidelines document addresses all key safety issues such that compliance with the guidelines will ensure that installation and operation of this system will not produce a safety concern. The guidelines document is currently in the process of being revised to incorporate NRC comments. 5.

CONCLUSIONS

Results to date support the conclusion that an effective remedy for pipe cracking in BWRs can be developed based on hydrogen injection to reduce the stainless steel corrosion potential combined with careful control of water quality. Two years of experience at

Hydrogen water chemistry for BWRs

69

Dresden-2 indicate that it is practical to implement HWC in an operating power plant and that long-term protection against IGSCC can be obtained. The only significant adverse impact of HWC found at Dresden-2--a substantial increase of the operating radiation fields--has proved to be manageable. Concerns about potential effects of HWC on fuel performance have been considerably allayed by the results of fuel examinations after one cycle of HWC operation. The plant-to-plant variability observed in the minitests needs to be understood to give assurance that plant-specific "hydrogen water chemistry" specifications based on minitests will in fact provide protection against ICSCC over the longer term. In the meantime, periodic empirical demonstrations of HWC effectiveness are recommended for plants adopting hydrogen water chemistry. REFERENCES Andresen P. L. (1983) The Effects of Aqueous Impurities on Intergranular Stress Corrosion Cracking of Sensitized 304 Stainless Steel, EPRI NP-3384, November. Andresen P. L. and Indig M. E. (1982) Corrosion, vol. 38, no. 531. Burley E. L. et al. (1982) Oxygen Suppression in Boiling Water Reactors--Phase 2 Final Report, DOE/ET/34203-47, NEDC-23856-7, October. Burley H. H. and Anstine L. D. (1986) Radiological Effect of HWC in the Environs of U.S. BWRs, EPRI NP-4416, February. Cheng H., Ikemoto R. N. and Stumph F. R. (1985) Dresden-2 EOC 9 Site Examination, EPRI RPI930-10, November. Cheng B. and Blood R. E. (1986) Postirradiation Examination of Fuel Components After One Cycle of Hydrogen Water Chemistry in Dresden-2, EPRI RPI930-10, July. Cheng g., Levin H. A., Adamson R. B., Marlowe M. O. and Monroe V. L. (1985) "Development of a Sensitive and Reproducible Steam Test for Zircaloy Nodular Corrosion," Proceedings of the 7th International Conference on Zirconium in the Nuclear Industrx, Strasbourg, France, June. Cheng C. Y., Gamble R. M., Taboada A., and Turovlin B. (1980) USNRC NUREG 0313, Rev. I, July. Childs W. et al. (1983) Plant Materials Program: NP-2879-SR, February.

Progress, June 1981 to May 1982, EPRI

Cowan R. L., Gordon B. M., Kiss E., Sundberg L. L. and Adamson R. B. (1985) "Hydrogen Water Chemistry Operating Experience," Proceedings of Post SMiRT Conference, Ispra, Italy, August. Cowan R. L., Indig M. E., Kass J. N., Law R. J., and Sundberg L. L. (1986) Water Chemistry of Nuclear Reactor Systems 4, volume i, 29-36, British Nuclear Energy Society. Cox B. (1983) Effect of Hydrogen In~ection on Hydrogen Uptake by BWR Fuel Cladding, EPRI NP-3146, RP1930-5, June. Davis R. B., Indig M. E. and Weber J. E. Analysis of Dresden-2 SSRT and ECP Data, to be published. Davis R. B. and Indig M. E. (1983) "The Effect of Aqueous Impurities on the Stress Corrosion Cracking of Austenitic Stainless Steels in High Temperature Water," Corrosion'83, Paper #128, Anaheim, California. Dillman g. H., Head R. A. and Liu C. C. (1985) BWR Coolant Impurity Study, EPRI NP-4156, August. Ford F. P. (1982) Mechanisms of Environmental Cracking in Systems Pecular to the Power Reactor Industry, EPRI NP-2589, September. Franklin D. C., Ceh[ S. M., Machiels A. J. and Santucci J. (1985) LWR Core Materials Program: Progress in 1983-1984, EPRI NP-4312-SR, October.

70

W. BILANIN et al.

Gordon B. M., Jewett C. W., Pickett A. E., Indig M. E., Andresen P. L., and Niedrach L. W. (1985) Hydrogen Water Chemistry for BWRs, Interim Report, EPRI NP-3959M, April. Gordon B. M., Pickett A. E., Jewett C. ., and Indig M. W. (1983) "Laboratory Studies on Hydrogen Water Chemistry," Paper presented at EPRI Seminar on Countermeasures for BWR Pipe Cracking. Palo Alto, Calif., November 14-18. Gordon B. M., Cowan R. L., Jewett C. W., and Pickett A. E. (1983) "Mitigation of Stress Corrosion Cracking Through Suppression of Radiolytic Oxygen," Paper #50 presented at the International Symposium on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors, Myrtle Beach, South Carolina, August 24. Gordon B. M. (1980) Materials Performance, vol. 19, no. 4, pp. 29-38. Indig M. E. and Weber J. E. (1983) "Mitigation of Stress Corrosion Cracking in an Operating BWR Via Hydrogen Injection," Corrosion '83, Anaheim, Calif., April 18. Indig M. E., Gordon G. M. and Davis R. B. (1983) "The Role of Water Purity on Stress Corrosion Cracking," Proceedings: International Symposium on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors, Myrtle Beach, August. Johnson A. B. (1983) ATR and ETR Corrosion and Hydriding Studies--Relevance to Hydrogen Addition to BWRs, EPRI RP1250-4, January. Jones R. L., Danko J. and Bilanin W. (1985) Boiling Water Reactor Owners Croup Intergranular Stress Corrosion Cracking Research Program: Executive Summary, EPRI NP-4273-SR, October. Kurtz R. J., Shannon D. W., Francis B., Kustas F. M. and Kolhrnstedt P. L. (1983) Evaluation of BWR Resin Intrusions on Stress Corrosion Cracking of Reactor Structural Materials, EPRI NP-3145, June. Lin C. C. (1985) BWR Cobalt Deposition Studies, EPRI NP-4236, September. Ljungberg L. and Hallden E. (1984) BWR Water Chemistry Impurity Studies: Review of Effects on Stress Corrosion Cracking, EPRI NP-3663, September.

Literature

Marlowe M. O., Armijo J. S., Cheng B. and Adamson R. B. (1985) "Nuclear Fuel Cladding Localized Corrosion," Proceedings: Topical Meeting on Light Water Reactor Fuel Performance, Orlando, Florida, April. Martin G. C., Judd C. R., Lewis J. E. and Smith F.R. (1985) Fuel Rod Crud Deposit Examinations - First Inspection. EPRI RPI930-10, November. Roberts J.T.A., Jones R.L., Naughton M. and Machiels A.J. (1985) "BWR Pipe Crack Control Using Hydrogen Water Chemistry Status Report on Dresden-2 Program," Nuclear Engineering and Design 89, p. 505. Ruther W. E., Soppet W. K., Ayrault G., and Kassner T. F. (1983) "Effect of Sulfuric Acid, Oxygen, and Hydrogen in High Temperature Water on Stress Corrosion Cracking of Sensitized Type 304 Stainless Steel," Paper #125, Corrosion '83, Anaheim, Calif., April 18. Guidelines (1986) Guidelines for Permanent BWR Hydrogen Water Chemistry Installations, EPRI NP-4500-SR-LD, March. ACKNOWLEDGMENTS

J. E. L. M.

The authors gratefully acknowledge the contributions made by T. Wojnulewicz and Almer of the Dresden-2 reactor; Commonwealth Edison Project Managers, E. Zebus and Rowley; General Electric Project Managers, R. Cowan, B. Gordon, and R. Adamson; and Anstine of APT. Thanks are also due to former EPRI employees M. Fox, A. Roberts, and Naughton who ably managed portions of the work described herein.