i . I
L Technical Paper Once-Through Steam Generator (OTSG) Materials and Water
RDTP 74-9
F. J. Pocock Scientist Research 81 Development Division Alliance, Ohio
D. F. Levstek Principal Materials Engineer Nuclear Power Generation Division Lynch burg, ‘Virginia
Presented to XI X Nuclear Conference Rome, Italy March 21 -22, 1974
-.-.
Babcock & Wilcox
DISCLAIMER
This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency Thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.
DISCLAIMER Portions of this document may be illegible in electronic image products. Images are produced from the best available original document.
1
ONCE - TH ROUG H STEAM GENE RATOR (OTSG) MATER I ALS AND WATER CHEMISTRY
F. J. Pocock, Scientist Research & Development Division, Alliance, Ohio
D. F. Levstek, Principal Materials Engineer Nuclear Power Generation Division, Lynchburg, Virginia -
Presented to XIX Nuclear Conference, Rome, Italy March 21 -22, 1974
INTRODUCTION
Sankovich and McDonald(’ described the performance of a once-through steam gen- e ra tor (OTSG) a t the XVI Nuclear Con- gress i n 1971. The f i r s t Babcock & Wi lcox Company Nuclear Steam Supply System incorporating these steam gener- a tors went i n t o service l a s t year a t the Oconee Station of the Duke Power Company near Seneca, S o u t h Carolina i n the United States .
In this paper, we are summarizing the materials and water chemistry research resu l t s associated with the development of this steam generator. Me are a lso i n c l u d i n g a summary o f water chemistry data acquired d u r i n g preoperational tes t ing and power operation t o date .
These data confirm the operational prac- t i ca l i t y of the nuclear once-through concept u s i n g vo la t i l e water treatment and high purity condensate demineralized feedwa t e r . The resul t i ng cycle and steam generator cleanliness w i 11 be of i n t e re s t i n view of reported t u b i n g ma- t e r i a l f a i lu re s in Europe and the United States . These f a i lu re s are apparently associated with the concentration of corrosive chemicals on the steam gener- a t ing tube surfaces.(2) Our d a t a reveal t h a t such an event i s unlikely when these steam generators are operated w i t h i n specified water conditions.
The materials and water chemistry re- search program associated w i t h the development o f the OTSG included the following:
1.
2 .
3 .
4.
5.
Water qual i ty speci f i ca t i on development
Materi a1 s behavior w i t h i n these water quali ty specifications and d u r i n g r e a l i s t i c temporary excur- sions
Water treatment and p u r i f i ca t i on research t o establ i sh control parameters and capabi: i t i e s d u r - i ng system excursions
Deposition and/or concentration behavior of dissolved and sus- pended sol ids in the steam generator when operating w i t h i n and o u t of specifications
Corrosion product removal (chemical cleaning).
The de ta i l s of each of these programs and ear ly operational d a t a of the Oconee Nuclear Plant have been reported previously a t ari us forums i n the United States . Y3-7?
2
OTSG SYSTEM DESIGN
PLANT DESCRIPTION
Oconee U n i t I Nuclear Steam Supply Sys- tem contains a pressurized light-water moderated and cooled reactor w i t h once- t h r o u g h steam generators (OTSG's). High-pressure h i gh-temperature (- 2200 psia and 600"F, or 150 atm and 316°C) water c i rcu la tes t h r o u g h a closed loop and transfers heat from the reactor core t o the steam generators where steam a t about 900 psia (-61 atm) and 594°F (-312°C) i s produced t o dr ive the turbogenerator. Oconee I i s the f i r s t commercial power reactor t o pro- duce superheated steam. T h i s system produces 885 MWe ne,t a t 100% power.
PRIMARY SYSTEM
The Oconee I primary system consists of a reactor vessel , two main coolant
Q U E N C H W (WD)
K E Y '
CORE F L O O D I N G SYSTEM
COOLING WATER SYSTEM
5 b I s
OH - D E C A Y H E A T R E M O V A L SYSTEM
HP - H I G H PRESSURE INJECTION SYSTEM
M U - M A K E UP A N D P U R I F I C A T I O N SYSTEM
T E R WD - WASTE DISPOSAL SYSTEM
REACTOR COOLA DRAIN TANK
DRAIN HEADER STEAM GENERATOR (WD)
STEAM GENERATOR
STEAM OUTLET STEAM OUTLET FEEDWATER INLET FEEDWATER INLET
OR COOLANT PUMP LET DOWN 'LINE I
HIGH PRESSURE
(HP) (MU) SAFETY INJECTION
COOLING WATER
SEAL WATER (MU)
loops feeding two OTSG's, a pressurizer t o maintain the reactor coolant i n a subcooled condition, and four cen t r i f - ugal pumps t o return the coolant from the steam generators t o the reactor vessel. Figure 1 i s a schematic dia- gram of the primary system.
The materials in contact with the pr i - mary water are corrosion res i s tan t alloys such as Alloy 600 fo r the OTSG tubes, Zircaloy 4 cladding fo r the fuel elements, and unstabilized aus ten i t ic s ta in less s tee l cladding f o r the reac- tor vessel, pressurizer, and main piping. The primary faces of the OTSG tubesheets are weld clad with a mate- r i a l corresponding t o the specification for Alloy 600. Alloy 600, Zircaloy 4, and s ta in less s teel - comprise about 75, 20, and 5% respectively of the total surface area exposed t o the reactor coolant.
These materials -
FIGURE 1 REACTOR COOLANT SYSTEM
SECONDARY SYSTEM dra ins back t o the condenser h o t well when desirable (see Figure 2 ) .
The Oconee I preboiler cycle, which i s s imilar t o the cycle employed on once- through fossi 1 boi 1 e r cycl es , i s shown schematically i n Figure 2. The major secondary system piping i s made o f carbon s t e e l . The condenser and low- pressure feedwater heaters are tubed with s t a in l e s s s tee l ; the high-pressure feedwater heaters and the turbine mois- ture separators and reheaters are tubed w i t h carbon s t e e l . Alloy 600, as previously mentioned; the OTSG shroud, shel l , support p la tes , and tubesheets a re carbon s t e e l .
The OTSG tubing i s
During system s ta r tup , the OTSG's and the turbine are bypassed t h r o u g h the recycle feedwater cleanup system (see Figure 2 ) . This system i s capable of recirculat ing 25% of fu l l load flow t h r o u g h the condensate demineralizers t o meet s tar t -up feedwater qual i ty c r i t e r i a .
During power operation, a1 1 condensate is purified by the Powdexa condensate demi neral i zi ng sys tem. About 30% of the steam is e i t h e r used fo r reheating or extracted from the turbine fo r feed- water heating purposes. The heater drains , with the exception of the lowest pressure heaters, a re normally pumped forward w i t h the feedwater, thereby bypassing the condensate demineralizing units. There are , how- ever, provisions fo r dumping the heater
A t f u l l power the boi ler feedwater sys- tem provides about 11 million pounds (-4,994,000 kilograms) of water a t 460°F (-238°C) per hour t o the steam generators.
Dur ing plant outages (for refueling and/ or maintenance) the steam generators are stored wet under a nitrogen blanket w i t h hydrazine and ammonia-treated feedwater qual i ty water. During wet storage, water i n the steam generator is r ec i r - culated by means of the system shown i n Figure 3 t o ensure proper mixing and dis t r ibut ion of treatment chemicals t o a1 1 surfaces.
OTSG SYSTEM CHEMISTRY
REACTOR COOLANT SYSTEM CHEMISTRY
The reactor coolant chemistry spec i f i - cations and averaged data obtained during preoperational tes t ing the f i r s t year. of power operation are shown i n Table 1 .
Boron, i n the form o f boric acid, i s used as a soluble poison i n the reactor coolant t o h e l p control the operat ion of the reactor by improving the power d is t r ibu t ion i n the core.
DEMINERALIZERS
LOW PRESSURE
FIGURE 2 SECONDARY SYSTEM FLOW PATH
EMERGENCY
FEEDWATER LINE T 7-L RECl RCULATING
LOWER T ~ B E - S H E E T DRAIN
FIGURE 3 STEAM GENERATOR WET LAY-UP SYSTEM
4
PPM B
PPM 'L,
pH B 77'F I-25'CI
' PPM O2
PPM CI'
PPM F-
H2 cdKGlSTP
TOTAL GAS, cc/KG/STP
TABLE 1 REACTOR COOLANT WATER QUALITY
POWER OPERATION
<2100
0.2-2.0 0.35 0.35
4.68.5 8.0 6
0.1 (MAXI <O.l <o. 1
0.1 (MAXI <0.05 <O.ffi
0.1 (MAXI 0.07 0.05
15-40 15 15
1W (MAXI 40 35
SUSP. SOLIDS, PPM
PPM N2H4
7 L i t h i u m hydroxide C 7 L i O H ) i s the chosen pH additive. present since i t i s formed in the reac- t o r coolant by the reaction of 1oB ( n , ~ 1 ) 7Li and i s always present when the coolant contains boric ac id . . Lithium hydroxide i s more e f fec t ive i n maintaining an alkal ine pH a t e le - vated temperatures than i s a vo la t i l e additive such as the ammonium hydroxide (NH4OH) additive used by some i n the past . T h i s vo la t i l e additive was used because of concerns about the e f f ec t s of concentrated nonvolatile a1 kal ies on Zircaloy corrosion; however, experi- ence has shown t h a t this e f f ec t is n o t of concern w i t h shim reactors. Figure 4 shows the behavior of ammonium hy- droxi de a d d i t i v e s a t reactor cool a n t operating temperatures.
I t i s naturally
<O.l' 0.025 0.015 <O.OlO
- t - - t 0.1-1.0 0.5t
0.25
6.2
0
(0.05
<0.05
-20
38
5.4
-0- loo0 W M H$03 --- 10 PPM NH3 + 1oM) PPM H 9 0 3 ---- 2 PPM L, t lorn PPM ~ $ 0 ~
77 Mo 400 600 800 OF
25 93 205' 316 427
TEMPERATURE
5.6 5.9 6.9 7.2
Table 2 and Figures 5a and 5b show t h a t the stronger a lka l ies give be t te r gen- eral corrosion protection t o aus ten i t ic alloys ( s ta in less s tee l and Alloy 600) . This protection reduces surface metal release as well as long-term r a d i a t i o n levels . The l a t t e r reduction i s , of course, important from the maintenance standpoint. The low level of suspended sol ids (corrosion product oxides) a t Oconee confirms our choice of the non- vo la t i le additive lithium hydroxide.
PURE WATER
12lEST.l
2
TABLE 2 CORROSION RATE A T 572OF (-300°C) (mg/dm2 - mol(')
2
9
2
SOLUTION CHEMISTRY
BORON AS H$Os PPM
ALKALI (Li. KI. M
AMMONIA, PPM
I N W N E L
TYPE 304 STAINLESS STEEL
1500
27
5
1500
l o 4
5
2
1w4
4
2 ~
CORROSION RATE OF IO MGIDM~-MO = o woo55 IN /YR 1 0 . ~ 0 1 4 CM/YRI
The l i m i t a t i o n s on dissolved oxygen, f luoride, and chloride a re well under- stood because the i r e f f e c t on the s t r e s s corrosion behavior of usten- i t i c alloys has been s tudied.f l0) Hydrastine i s used t o preclude the e f fec ts of dissolved oxygen a t low temperature condi t i ons , whi 1 e hydrogen reacts with oxygen t h a t has been formed r a d i o l y t i c a l l y from water a t reactor coolant operating conditions.
Total gas concentrations a re 1 imi ted for operational reasons associated w i t h preventing inadvertent gas accumula- t ions in the control rod drive system located above the reactor.
SECONDARY SYSTEM CHEMISTRY AND MATERIALS BE HAV IOR
The water chemistry specifications for the f i n a l feedwater supplied t o the steam generator d u r i n g both preoper- ational tes t ing and power operation are contained in Table 3. included i n t h i s table are spec i f i - cations t h a t apply t o the boi ler water a t b o t h preoperational levels and <, 1% power opera t i on.
Also
5
PREOP AVG VALUE
9.4
0 45
50
20
< 10
< 1
< 2
20
150
I -1 AMMONIATED- .- BORATED - LITHIATED---
POWER OPERATION AVERAGE VALUES
JUN-NOV NOV-MAR
9 4 9 5
0 5 0 4
10 < 10
< m < m
< 10 < 7
< ' < ' < 2 < 2
15 < 1012-3)
32-50 E-1W
0 1WO 2OW 3000 4000 5000 6WO TIME, HOURS
BRIGHT ANNEALED
I I I I I 0 1000 3000 5ooo
TIME, HOURS
a. CORROSION OF WROUGHT SPECIMENS- I N b. CORROSION OF A L L O Y 600 IN 600°F ED, L ITHIATED, AND BORATED ( - 3 1 6%) BOR ATE D WATER ( 1 2)
FIGURE 5 DATA FOR METAL RELEASE RATE IN PRIMARY COOLANT
TABLE 3 FEEDWATER QUALITY
pH 0 770F (-25OC)
CATION CONO.
TOTAL SOLIDS, PPB
PPB S O 2
PPB O2
PPB N2H4
PPB Pb
PPB c u
'1W PPB FOR PREOPI
FW - SPEC 9.3-9.6
< 0.5
<50
< 10'
20
< 7
< ' < 2
4TIONAL T I
STEAM GENERATOR WATER - PREOPERATIONAL
PREOP.
pH 0 77OF (-25OCI
PPM CI-
PPM Na 20 0.25
In an OTSG feedwater enters the bottom of the tube bundle a t saturat ion tem- perature. Nucleate boi 1 i n g follows and the qual i ty of the steam-water mixture increases gradually t o 100% steam. The saturated steam i s then superheated as i t flows over the t u b i n g i n the upper portion o f the steam generator. As feedwater enters and flows upward through the st,eam generator, steam qual i ty increases rapidly and contam- inants such as sodium hydroxide and
sodium chloride dissolve completely i n the steam when the i n i t i a l concen- t ra t ion i n the feedwater i s w i t h i n water qual i ty specif icat ions.
A t power operations, both research and confirmatory f i e l d data show tha t only the monitoring of the f inal feedwater i s required t o assess boi ler water con- di t ions. Concentrations of corrosive chemicals such as sodium chloride and sodium hydroxide do n o t occur i n an OTSG over normal power operating ranges as they do i n recirculat ing steam gen- erators (see Figures 6 and 7 and Table 4).
TABLE 4
10-2E-73
1Cb2E-73
ANALYSES OF BOILED WATER* AT SHUT-DOWN OR TRIP AFTER EXTENDED POWER OPERATION
< 0.05
'BOILER WATER ANALYSES NOT NECESSARY DURING WWER OPERATION IF FEEOWATER IS WITHIN LIMITS. STEAM SOLUBLE SODIUM SALTS IN BOILER WATER DURING EXTENDED WWER OPERATION
SHUT.WWN SAMPLES TAKEN TO CONFIRM NO CONCENTRATION OF
tANALYTICAL METHOD DETECTION LIMITS, INDICATES ZERO TO POSSIBLE TRACE
6
1000
,,,I
a
1 w O
100 lx >. k I! 3 5: 10
1.a
- - O0 - ARC DATA
0
,” a” 0 AT - 7OOOF I371OCI AND 1542 PSlA (105 A T M ) ( l 3 I - AA
- (-716 - 1076°F1(14’ A - A TEMP RANGE 380 - 58OoC
- 0 AT SATURATION TEMP!”’ A
A
0 SODIUM CHLORIDE
. A SODIUM HYDROXIDE A
0 600 700 800 900 1000 1100 1200
STEAM PRESSURE. PSlA
34 41 48 54 61 68 75 82 -STEAM PRESSURE, ATM
FIGURE 6 SOLUBILITY OF SODIUM SALTS IN SUPERHEATED STEAM
Research i nves t i ga ti ons were a1 so made of the deposition behavior i n the steam generator o f a mixture o f soluble s a l t s and simulated suspended corrosion prod- ucts a t extremely h i g h concentrations i n the feedwater. The concentration chosen was 10,000 ppb, or 200 times the allowable normal upper l imi t of 50 ppb. T h i s h i g h concentration was chosen so tha t deposition patterns and chemical cleaning s tudies could be made in a reasonable research time-span. F i g - ure 8 shows the d is t r ibu t ion o f iron oxide (Fe 04) and sodium s a l t s (as the oxide Na2 a ) .
1 24‘ (-7.3MI
0 11 22 32 43 54 65 -GMIM2
AVG. WT. OF Fe304 AND N a p ,
FIGURE 8 DISTRIBUTION OF IRON OXIDE (Feg04) AND SODIUM OXIDE (Na20) IN LABORATORY OTSG
0.01 - 30 50 3 10
STEAM PRESSURE , PSlA X 10’
2 0 6.8 20.4 34.0
STEAM PRESSURE.“ATM X 10
FIGURE 7 SOLUBILITY OF SODIUM CHLORIDE
These data were obtained by shutting down the 1 aboratory steam generator dry under nitrogen seal and determin- i n g the weight d i s t r i b u t i o n and compo- s i t ion of .the deposits. Corrosion product deposition occurred primarily in higher qual i ty regions w i t h the heaviest deposits approximating the location o f departure from nucleate boiling. n o t dissolved i n the steam deposited predominantly on the superheated tube surfaces beyond the 100% qual i ty region.
The soluble so l ids t h a t were
OTSG MATERIALS BEHAVIOR
To provide more assurance tha t Alloy 600 is the best material for the OTSG, B&W has conducted an extensive t e s t program. Results t o date , as shown i n Table 5 , confirm our confidence in the selection of Alloy 600 for the OTSG. There has n o t been a s ingle f a i l u r e in any o f the t e s t s which simulate actual conditions i n the OTSG.
7
TABLE 5 OTSG MATERIALS RESEARCH DATA
PARAMETRIC AUTOCLAVES
CONTAMINANT AUTOCUVES
TATIC
ENVIRONMENT
OTSG OPERATING ENVIRONMENT
OTSG OPERATING ENVIRONMENT WITH TEMPERATURE VARIATIONS
Pb. N d H . NE1
HIGH PURITY WATER
~ ~~ ~
RESULTS
NO DISTRESS UNDER OTSG SIMULATED CONDITIONS IN OVER 28.W HOURS
NO DISTRESS AT OTSG OPERATING TEMPERATURE IN OVER 1 B . W HOURS TWO HIGHLY STRESSED SPECIMENS FAILED IN 15-1B.W HOURS AT 6WoF (-3U'%l
FAILURE IN HIGHLY STRESSED SPECIMENS CONTAMINATED WITH F% IN 4 - 5 . W HOURS NO OTHER DISTRESS NOTED IN OVER 2o.m HOURS WITH LOWER STRESSED SPECIMENS - NO FAILURE I N NaOH OR NlCl
NO DISTRESS IN OVER 17 WO HOURS
I t i s evident i n our research program, as well as i n the work of others, t ha t under cer ta in conditions Alloy 600 i s susceptible t o s t ress-assis ted inter- granular corrosion. The three most important fac tors controll ing the sus- c e p t i b i l i t y of this material are (1) stress leve l , ( 2 ) environmental condi- t ions , and ( 3 ) metallurgical conditions.
STRESS LEVEL
Our data and t h a t of others show t h a t a h i g h t ens i l e stress level i s required for stress ass i s ted intergranular crack- i n g of Alloy 600.
Coriou, for example, has reported the lowest applied stress tha t produced stress corrosion cracking (SCC) o f Alloy 600 was 75% of the tes t mater ' a l ' s ambient temperature y ie ld strength. 116) Our tes t data further support this posi- t ion. With gross lead contamination, Alloy 600 specimens exhibited SCC i n 4000-5000 hours when stressed t o 125% of their ambient temperature y ie ld strength i n an otherwise normal OTSG operating environment. Specimens s t ressed t o 90% of the ambient temper- a ture y ie ld strength w i t h the same gross contamination and i n the same tes t auto- clave have now surpassed 20,000 hours exposure without f a i lu re .
The OTSG is unique i n t h a t i t i s sub- jected t o f u l l furnace stress r e l i e f g t 594 t o 621 "C (-1101 t o 1150°F) a f t e r a l l fabr icat ion i s completed.
This heat treatment reduces the level o f residual s t r e s s and thereby the to ta l applied s t r e s s i n the t u b i n g i n areas such as the expanded zone in the tube- sheet. Aside from this reduction i n s t resses , i t now appears tha t the Alloy 600 microstructure produced by s t r e s s r e l i e f is a l so desirable from the s t a n d - point of resistance t o SCC.
ENVl RONMENTA L CONDITIONS
All instances t o date of Alloy 600 steam generator tube f a i l u r e caused by SCC have been ascribed t o upset secondary water chemistries, usually involving s igni f icant concentrations of f r ee caus- t i c . S t ress corrosion cracking i n reportedly "pure water" s t i l l seems t o be a 1 aboratory "curi osi ty" . Using refreshed (flow-through) autoclaves and model boi lers w i t h condensate deminer- alized feedwater and vo la t i l e treatment under OTSG operating conditions has not resulted i n the f a i l u r e o f any heat- treated 621 "C ( - 11 50°F) materi a1 i n our 1 aboratory except f o r one "U" - bend specimen t h a t was cold bent a f t e r heat treatment.
METALLURGICAL CONDITIONS
The variable of metal 1 urgi cal condi - t ions has generally been ignored i n Alloy 600 SCC studies i n the past . Increasing evidence, however, indicates tha t controllable variables such as heat treatment can have a profound e f f ec t on the SCC of Alloy 600. Coriou i n his most recent paper reports t ha t comnercial a l loys heated for one hour a t 700°C (-1292°F) d i d no t crack i n tes t of 10,000 hours dura t ion i ( - 662°F) uncontaminated water. Nhen solution annealed, the same al loys m r e subject t o SCC i n the same 350°C (-662°F) water t e s t .
'71 ??Ooc
Results t o date
8
in our test program substant ia te these findings since no specimen given the OTSG s t r e s s re1 ie f treatment (600°C or -1112°F) a f t e r t h e i r fabrication has exhibited SCC a f t e r t e s t duration for more than 28,000 hours. Further- more, no SCC f a i lu re s of Alloy 600 t u b i n g i n a steam generator have been reported when a f u l l s t r e s s r e l i e f has been appl i ed.
Material condition apparently also has a s ign i f icant e f f ec t on SCC of Alloy 600 i n contaminated aqueous environ- ments.
Wagner, Schenk, and Spahn report tha t sensi t i zi ng treatments a t 400°C ( - 752°F) and less produce a def in i te ly adverse e f f ec t in aerated, 5% NaOH t e s t s a t 325°C (-617°F). However, a heat treatment a t 600°C (-1112°F) pro- duced no increase in suscept ib i l i ty unless the material had previously received a high-temp ra ture , solution- anneal i n g treatment.718) Another study, using capsule specimens contain- ing deoxygenated, highly concentrated caust ic solutions a t 300" produced s imilar r e su l t s .
In another s e r i e s of U-bend t e s t s in deoxygenated 10 and 50% NaOH solutions a t 316°C (-601"F), both cold work and heat treatment a t 650°C (-1202°F) were found t o improve t h c rrosion resistance o f Alloy 600. 7 7 20
REMOVAL OF DEPOSITS
Regardless of how pure the feedwater, some transporting of impurities i n t o the steam generator will occur. Con- taminants t h a t a re not soluble i n steam are principal ly corrosion product oxides. Blowdown has proved ineffec- t i ve in removing these materials from a l l large nuclear steam generators; therefore , chemical cleaning will be periodically required t h r o u g h o u t the l i f e of the equipment.
A chemical cleaning research program was conducted a t the B&W Alliance Research Center ( A R C ) a t Alliance, Ohio, consisted of evaluating various so l - vents f o r t he i r a b i l i t y t o dissolve magnetite (Fe304), the principal cor- rosion product oxide in a l l s tee l cycles, without s ignif icant ly corroding carbon s tee l and Ni-Cr-Fe alloy s t am
The e f f ec t of cleaning solvent on the localized high flow areas o f a steam generator and the best method for t he i r application was also investigated. Table 6 shows the resu l t s o f impinge- ment-corrosion of solvents in Ni-Cr-Fe alloy t u b i n g during simulated additions and/or reci rcul a t i on .
generator s t ructural materials. ( 21 7
TABLE 6 CHEMICAL CLEANING TESTS
'SOLUTIONS CONTAINED EITHER 0.296 RODINE 31A OR DOW A145 INHIBITORS. THE pH WAS OBTAINED WITH NH40H. AND ALL TESTS WERE STATIC, CONDUCTED AT 2OOOF l-93°CI.
EACH TEST CONTAINED 2.27% FejOq
'SOLUTIONS CONTAINED EITHER 0.2% RODINE 31A OR DOW A145 THE pH WAS
TESTS WERE CONDUCTED AT A VELOCITY OF 14 FPS OBTAINED WITH NH40H. 1-4.3 MKECI IMPINGEMENT ON A'CARBON STEEL COUPON AT 200OF 1-93OCI
The solvent used in these t e s t s a lso contained suspended and dissolved sim- ulated corrosion products consisting of 0.16% Cr2O3, 0.16% NiO, and 2 .27% Fe304. The suspended material was added t o investigate i t s abrasiveness when introducing or recirculating the solvent. Velocities t o 15 fps ( -4 .6 m/sec) with impingement on the tube samples (contained in a 14-tube bundle) and a temperature of 200°F (-93°C) were used, perature was chosen because of inhibi tor l imitat ions; the 15 fps
The 200°F tem-
9
velocity was the maximum calculated a t any location, based on minimum flows o f -0 .2 fps (-0.06 m/sec) i n the steam generator when water enters th rough ports a t the bottom and ex i t s th rough ports a t the t o p . The pene- t r a t ion r a t e per 8 hours ranged from 0.0003 t o 0.002 mils. The to ta l expo- sure time was 78.5 hours fo r the t e s t described, or the equivalent of about 10 chemical cleanings over steam gen- e ra tor service l i f e .
Two of the laboratory steam generators a t the ARC were fouled using a mixture of magnetite (Fe304), f e r r i c hydroxide (Fe[OH]3), and nickel oxide ( N i O ) i n an accelerated t e s t and then chemically cleaned with one of the developed sol- vents using the " f i l l , soak, and d r a i n " method of chemical cleaning. This method was chosen over c i rculat ion methods because of a be t te r d i s t r i b u t i o n of f resh solvent. had been previously cleaned i n another t e s t u s i n g an e a r l i e r version of the solvent and the continuous circulat ion method o f chemical cleaning. Figure 9 shows the r e su l t s w i t h b o t h of the methods. o f material introduced and removed from one o f the laboratory OTSG's a t t h a t time.
The steam generators
Table 7 gives the mass balance
After chemical cleaning, operating t e s t s showed tha t the performance of the laboratory uni t was returned t o t h a t
-KG LB
3. 18
2.72
2.27
0 2 1 8 2
s1 E n 136
LL
0.91
0.45
7
CIRCULATION
/ FILL-SOAK-DRAIN
0 1 2 3 4 5 6
TIME, HOURS
FIGURE 9 IRON DISSOLUTION DURING LABORATORY OTSG CLEANING
of the prefouled condition. Destructive examination did n o t reveal t h a t the sol- vent caused any adverse e f fec ts on the materials used in the construction of the OTSG. The solvent was found t o decompose a t elevated temperatures t o C O z , NH3, and Hz0; there appears t o be no danger of accelerated long-term corrosion i f small amounts of residual solvent remain i n the OTSG on r e s t a r t a f t e r cl eani ng . This method was also shown t o produce less corrosion in carbon s tee l t h a n the continuous circulat ion method because of the shorter exposure period under hi gh velocity conditions . During chemical cleaning, the iron con- centration and solvent composition were
TABLE 7 CONTAMINANT REMOVAL AFTER FOULING TEST measured periodical ly to determine when the cleaning operation was complete.
CONTAMINANT ADDED
4MT. REMOVED 2WoF (-93OCI DEMlN
WATER RINSE STEAM GENERATOR
4MT. REMOVE0 200°F,(-930CI DEMIN WATER RINSE FEEDWATER HEATER
AMT. REMOVED STEAM GENERATOR CHEMICAL CLEANING (DISSOLVED)
AMT. REMOVED STEAM GENERATOR CHEMICAL CLEANING (SUSPENDED1
AMT. REMOVED FEEDWATER HEATER
CHEMICAL CLEANING IDISSOLVED1
AMT. REMOVED FEEDWATER HEATER CHEMICAL CLEANING (SUSPENDED)
TOTAL AMT. OF CONTAMINANT
REMOVED
e. LE (-KG1 NiO. LE (-KG1
11.12 (5041 0.816 (0.3701
0.46 (a211 0.033 (0.0151
0.40 ( a i 8 1 0.029 10.0131
7.85 (3.56) 0.047 (0.0211
0.084 (0.0381 0.006 IO.W)31
1.96 (0.891 0.001 (O.Ow51
0.018 IO.W811 0.001 (0.005)
10.77 (4.891 0.117 (0.0531
- . We highly recommend t h i s evaluation method. Figure 9 reveals t h a t the cleaning process was complete a f t e r 3 . 5 hours of the exposure time for the recirculation t e s t . A longer period was required for f i l l , soak, and drain. Detai 1 s concerning the chemical clean- ing research pr r m are contained in a separate paper. V 1 7 Since copper alloy condensers are present i n the cycles of some more recent plants , current re- search i s al so being directed toward developing a solvent system for remov- ing copper containing deposits.
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SUMMARY 2 . "Steam Generator Problems a t PWR's",
Research s tudies and f i e l d data from the f i r s t operational once-through steam generator (OTSG) have led t o the fol 1 owi ng concl usi ons :
1 . Volati le water treatment and 100% condensate demi neral i z i n g of feed- water used f o r OTSG's have proved both desirable and prac t ica l .
2 . Because of t h e i r so lubi l i ty i n OTSG steam, concentrations of dissolved feedwater contaminants, such as sodium hydroxide and/or sodium chloride, do n o t occur i n OTSG boiler water when feedwater purity i s w i t h i n specified l imi t s .
3. In addition t o the obvious bene- f i t s of stress reduction, the A1 1 oy 600 mi cros t ructure achi eved w i t h the full-furnace s t r e s s r e l i e f of the OTSG can be bene- f i c i a l from a s t r e s s corrosion resis tance standpoint. Data from uncontaminated water and abnormal water conditions (including NaOH Contamination) exhibi t t h i s bene- f i c i a l e f f ec t .
4 . Metal re lease ra tes i n the reac- t o r coolant revealed by both research and f i e l d operating data a re unaffected by the microstruc- tural conditions t h a t occur as a r e s u l t of the full-furnace r e l i e f a t 600°C (-1112°F).
5. The OTSG can be chemically cleaned. Research studies and laboratory steam generator chemical cleaning methods used t o remove simulated corrosion products have demon- s t r a t ed th i s f ac t .
REFERENcts
1. M . F . Sankovich and B . N . McDonald, "Once Through Steam Generation", XVI Nuclear Congress, Rome, I t a l y , March 25-26, 1971.
Reactor Safety, ROE-73-B , USAEC , August 31, 1973.
3. 6. N . McDonald and L . E . Johnson, "Nuclear Once Through Steam Gener- a tor" , ANS Topical Meeting - Power Systems and Components, Williams- b u r g , Virginia, September 2 , 1970.
4. J . H. Hicks, N . J . Mravich, and F. J . Pocock, "Nuclear Steam Supply System Water Chemistry Research", American Power Conference, Chicago, I l l i n o i s , April 2 1 , 1971.
5. F . J . Pocock, J . H . Hicks, N . J . Mravich, and M. J . Bell , "Water Chemistry f o r PMR Nuclear Steam Supply Systems" , 10th Annual L i b - e r ty Bell Corrosion Course, P h i lade1 p h i a , Pa. , September 14 , 1972.
6 . R . R . Beach, M . J . Bell , J . H . Hicks, and F. J . Pocock, "Preoper- ational Water Chemistry Control fo r Nuclear Steam Supply System" , American Power Conference, Chicago, I l l i n o i s , May 9 , 1973.
7 . R . M . Koehler, R . R . Beach, D . F . Hallman, and F. 3. Pocock, "Oconee U n i t I - Water Chemistry", ANS Meeting , San Franci sco , Cal i forni a , November 11 -1 5 , 1973.
8. F. J . Pocock, "Comments, Primary Coolant of PWR's", 31st Interna- t ional Water Conference Proceedings , Wi 11 iam Penn Hotel , P i t t s b u r g h , Pa. , October 28, 1970, p . 64.
9. I.I. D . Fletcher, "Primary Coolant of PWR's", 31st International Mater Conference, I;li 11 iam Penn Hotel , P i t t s b u r g h , Pa. , October 28, 1970.
10. A . L . Lowe, J r . and F . J . Pocock, "Stress Corrosion in Steam Genera- t ion Systems", ASME Jo in t Power Conference, S t . Louis, Missouri, September 2 2 , 1971.
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J . Blanchet, H. Coriou, L . Gral l , C . Mathieu, C . Ot ter , and G . Turluer, Preprint G-13 , International Con- ference: Stress Corrosion Cracking and Hydrogen Embrittlement of Iron- base Alloys, Firminy, June 1973.
G . H . Wagner, H. Schenk, and H . Spahn, Prepared Discussion on Paper G-22, International Confer- ence: Stress Corrosion Cracking and Hydrogen Embrittlement o f Iron-Base Alloys, Firminy, June 1973.
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F. J . Pocock and W . S . Leedy, "Chemical Cleaning Research f o r Nuclear Steam Generators", Inter- national Water Conference, William Penn Hotel, P i t t s b u r g h , Pa., November 3, 1971.
