ML20052C131
| ML20052C131 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 04/29/1982 |
| From: | Longenecker J ENERGY, DEPT. OF |
| To: | Check P Office of Nuclear Reactor Regulation |
| References | |
| HQ:S:82:024, HQ:S:82:24, NUDOCS 8205040404 | |
| Download: ML20052C131 (37) | |
Text
_ _ __ _ -. _ _ _ _ _ - _ -- _
Department of Energy Washington, D.C. 20545 APR 2 9 1982 Docket No. 50-537 HQ:S:82:024
,. q\$l!.; 7 '
s f/ . ~.]
Mr. Paul S. Check, Director [jI CRBRP Program Office Office of Nuclear Reactor Regulation q .,.232 m m
' Q U.S. Nuclear Regulatory Commission Qr Washington, DC 20555 k(\A'
],@/2
\n O
Dear Mr. Check:
967 7 g C-RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION - CHEMICAL AND MECHANICAL ENGINEERING
References:
- 1. Letter P. S. Check to J. R. Longenecker, "CRBRP Request For Additional Information," dated February 26, 1982
- 2. Letter P. S. Check to J. R. Longenecker, "CRBRP Request For Additional Infcnnation," dated March 23, 1982
- 3. Letter J. R. Longenecker to P. S. Check, " Responses to Request For Additional Information - Chemical Engineering,"
dated April 21, 1982 This letter formally responds to your request for additional information contained in References 1 and 2.
The enclosed responses CS 281.6, CS 281.8, CS 281.9, and CS 281.12, along with the responses previously submitted in Reference 3, complete the Project's response to your questions from Reference 2.
Also enclosed are responses to questions CS 210.2 and CS 210.6. The remaining questions from Reference 1 will be answered as follows: CS 210.1, 3, 4, 5, and 7 by May 14, and CS 210.8 and 9 by May 28, 1982.
The enclosed responses will be incorporated into the PSAR in Amendment 68 scheduled for May 7, 1982.
Sincerely, kw (v WJL ao DOO/
.1 Jo n R. Longene er, Manager Licensing & Environmental / f Coordination Office of Nuclear Energy Enclosures P205040404 820429 PDR ADOCK 05000537 A PDR
~
2 cc: Service List Standard Distribution Licensing Distribution
Chiestian G210.2 Describe methods and procedures used to evaluate the structural integrity at locations having savers thermal stripping effects.
Basponse
'Ihe response to this question is provided in anended Section 4.2.2 by the addition of subsection 4.2.2.6.
QCS210.2-1
_ Amen _d. 66
4.2.2.6 hermal Striping Evaluation of Reactor Internals 4.2.2.6.1 General Methodology for %ermal Striping Evaluation n e general method used to evaluate components for thermal striping is illustrated in Figure 4.2-130. %e steps involved are as follows:
a) % e maximum potential for thermal striping is identified. Wis potential is the calculated temperature difference in sodium coolant from adjacent l streams. Included in the ta perature difference are 2a uncertainties. l b) Scale model tests are run as appropriate and the esponent thermal I striping factors are measured. S e measured striping factors are cmbined with the maximum striping potential to generate a thermal striping tmperattire time history and define a maximum fluid a T.
c) The thermal striping tmperature time history is used as the boundary temperature in a transient thermal analysis of selected cm ponent. tis analysis calculates the magnitude of the fluctuating difference in the surface tenperature and the mean temperature (Ts 'Dn) .
d) The range of Ts 'Dn is used to determine a stress or strain range which is used in the fatigue evaluation of the structure, i 1
W ere are primarily 4 methods used to evaluate structures for thermal l striping. I lst Method j his is the most simplified and conservative method used. In this method the maximum fluid AT (Hottest fluid Tep. - Coldest Fluid Temp) is compared to an allowable metal AT. (Tep. of the metal surface - metal mean tenp) . %e method for determining allowable metal AT is described in Section 4.2.2.6.4.
If AT fluid < AT metal allowable the structure is adequate for striping.
Disennnien of Conservatism:
here are several factors which cause this method to be conservative.
- 1. Using fluid tenperatures is conservative, since this assumes an infinite film coefficient and neglects the mean temperature effects on strain. !
Since striping is a fast transient the mean tenperature effect will be small for most cases. However, the film coefficient effect can be significant.
- 2. Using the observed maximum and minimum as unbrella tenperatures is i conservative since each cycle would not have this maximum strain rar.ge.
- 3. Assuming a biaxial stress state is conservative for L e e locations such as at corners, where there is no biaxial stress state.
- 4. Conservatism incorporated into the design such as factors on design l fatigue curves and temperature uncertainties.
l 4.2-22sa Amend. 68
_ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ m i
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2nd Method The second method is used when the 1st method does not show the structure adequate. This method is to use the film coef ficient and the materials thermal properties to determine the actual metal surf ace temperature. The metal surf ace aT is then compared to the allowable metal a T (Surf ace-mean) used in method 1. If a Tmetal surf ace <aTmetal allowable, the structure is acceptable f or striping.
Discussion of Conservatism Conservatism is the same as for method 1 but excluding item 1.
3rd Method (Detailed Analysis)
This method is used when methods 1 and 2 are not adequate to show that the structure is acceptable and there is reason to believe it can be shown adequate with this method. This method involves the evaluation of actual temperature date from prototypic tests. In this evaluation instead of using the hottest and coldest fluid temperatures, the fluid temperature data is first converted to metal temperatures by thermal solutions. Then the metal temperature peaks are umbrelied I.e., all temperature peaks between specified values are grouped and given the highest absolute temperature within the group. These umbrella temperatures are used to calculate strain ranges- for each group.
a = Instantaneous coefficient of thermal expansion "r g= a ATmetal v = Poisson's ratio (1 - v) cr = Strain range Where Tmetal = Dif ference between the max and min temperature variation of the metal within the group (conservative).
The number of cycles within each group (n1 ) is projected throughout the i reactor life. The allowable number of cycles (Nd,) for the strain range in the corresponding group is obtained from the applTcable f atigue curve. The total striping fetigue damage Is, Dfattgue = E! ni/Nd f This damage will then be added to the creep-f atigue damage calculated for other transients. If the damage calculated here is within the creep-fatigue damage envelop, Figure 4.2-47A, the striping is considered acceptable.
Conservatism i ltems 3) and 4) from Method 1 There is stilI some conservatism in umbrellaing the temperatures, but not as significant as before.
4.2-227b Amend. 68 w
a . .- .. .. . . b - l n . -. , - - - .
Pige-11,(62-0227)L8,22] - #28 ,
4th Method (Alternate Detailed Analysis)
Again this method is used only when methods 1, 2, or 3 are not adequate to )
show the structure acceptable. This method is used when the geanetry of the '
structure is such that the previously discussed methods are overly {
conservative f.e., thin walled structures where mean temperature of fects can J be substantial, and corners where blaxiality offacts vanish. Either offects tend to reduce the stress and strain.
For this method detailed finite element models are developed to determine the thermal and stress response of the structure for the specific striping environment. The stress and strain ranges determined from the models are then used to evaluate the f atigue damage due to striping.
Conservatism See item 4, from Method 1.
4.2.2.6.2 Maximum Potential Fluid ATs To evaluate the of fects of thermal striping on reactor internals, a first step is to determine the maximum potential for thermal striping that exists at each selected component. This maxinum potential is the maximum temperature dif ference ( AT) between the dif ferent sources of sodium in each region, including 2a uncertainties. For example, the striping potential at a lower shroud tube is the AT between the control assembly (C/A) 2a exit temperature and the hottest adjacent f uel assembly (F/A) +2aexit temperature. The striping potentials are developed using a conservatively high core AT that represents a 115% power condition. This core AT is about 600F higher than exp6cted in the CRBRP. A listing of the maximum potential fluid for dif ferent components is included in Table 4.2-68.
4.2.2.6.3 Maximum Fluid ATs Tii~e maximum fluid AT at a given location is determined empircally as a fraction of the maximum potential fluid AT. Water model testing was perf ormed in 1) a f ull scale mockup of a cluster of seven removable assemblies and in 2) the quarter scale IRFM (Integral Reactor Flow Model) at HEDL. Hot and cold water exiting the core assembly outlet nozzles at appropriate flow rates simulate th9 assembly to assembly AT. Fast response thermocouples near the surf aces of the components models provide the time history of the fluid temperature fluctuations normalized to determine a normalized striping f actor.
The results of these tests provide a measure of the maximum fluid AT at critical locations.
4.2-227c Amend. 68 L% tim >
...a. . _ . , .
7-Ass ==hiv Tests l The 7-assembly test model is a full scale model of the reactor internals l configuration above a cluster of 7 core assemblies including the assembly exit l nozzles. Configurations tested include the following components above the l central of the 7 core assemblles.
l o Lower shroud tube and primary control rod driveline. ,
o Lower shroud tube and secondary control rod driveline.
o Lower shroud tube at core periphery without control rod drivelIne.
o Instrumentation post.
Figure 4.2-131 is a sketch of the 7-assembly test model. The configuration here is with an Instrumentation post. In this configuration the thermal _ __
striping f actors are measured on the Instrumentation posts for the dif ferent combinations of fuel assemblies and blanket assemblies that are below it. ,
Integral Reactor Flow Model - Thermal Strining The integral reactor flow model (IRFM) is a 360 degree 0.248 scale model of l the CRBRP outlet plenum and its internal components which models significant ,
hydraulic and vibrational characteristics of the CRBRP. Sodium in CRBRP is modeled by water in IRFM. An overall view is shown in Figure 4.2-132. Fuel, control, and blanket assemblies are represented by tubes orificed at the l bottom to give the proper flow distribution. The top ends of all core outlet i nozzles are prototypic to produce the proper velocity profile. Four different !
zones (fuel assemblies, inner blanket and control assemblies, radial blanket assemblies, and removable radlal shield assemblies) were provided with i dif ferent specified flows and temperatures and were varied to model striping !
conditions at various core life conditions. l l
The frequency and amplitude of fluid temperature oscillations were monitored on the outlet plenum permanent reactor structures. Thermal striping in these l regior.s are due to temperature differences between the fluids exiting the fuel assembiles, the radial blanket assembiles, the inner blanket assembiles, the l control assemblies and the removable radial shleid assemblies.
In addition to the striping measurements at the reactor components, ;
measurements were made in the outlet plenum using radial traversing probes.
The axial location of these probes is Illustrated on Figure 4.2-133.
4.2-227d Amend. 68 May 1982
Test Results he results of these thermal striping tests are listed on Table 4.2-68 in the form of normalized striping factors. h e factors are a fraction of the striping potential with a factor of 1.0 being 100% of potential. Results frm I the traversing probes in the outlet plenum are illustrated on Figure 4.2-134 ]
in the form of striping AT contours. %e striping levels on all components outside of this region, such as the reactor vessel thermal liner, the suppressor plate, and the primary piping, are within the range of allowables as shown on Table 4.2-69.
4.2.2.6.4 Stripina timits for Alloy 718 and stainla== steels 304 and 316 Many reactor internal components experience strain controlled cyclic 6
deformations in excess of 1 x 10 cycles. Frmtheavailablegatiguedata, reference 182, the allowable strain or stress range for 1 x 10 cycles or more approaches the endurance limits of each of the concerned materials. %ese i allowable strain or stress limits can be converted into limits on the metal I surface to mean temperature range (AT total) . We allowable striping AT metal l can be conservatively determined from the endurance stress or strain ranges i for a fully constrained biaxial stress case, e.g. thick flat plate. he following relations are used; AT metal " II~") ^ r (a) (stress formula) ai AT metal = (1-v) A C r (b) (strain formula) i where AT metal - Metal surface temperature minus the mean temperature See Figure 4.2-130, Section C Aa r
= endurance stress range (253)
Ac r
= endurance strain range og
= Instantaneous coefficient of thermal expansion v = poisson's ratio E = modulus of elasticity 4.2-227e Amend. 68
_- ---- ------------ % ncme
h is limit represents one of the most conservative cases as discussed in Method 2 in 4.2.2.6.1. If the striping AT metal en the ocuponent is below this limit, no further analysis is necessary.
In some applications the alternating stress or strain is ace-maied by mean stress or strain. S e mean stress or strain effect is detrimental and the allowables should be adjusted accordingly.
For 304 and 316 stainless steel the design endurance strain ranges can be obtained frm Figure 4.2.47B, provided that the metal temperatures do not exceed 1100 F. We striping AT metal limits are given in Table 4.2-69 with selected mean strain effects.
Se allow 718 fatigue carve is shown in Figure 4.2-48. % e striping stress limits are developed from the endurance stress8 range of the up-to-date data provided in Reference 183. W e high cycle (10 or more cycles) design fatigue strength curve with no mean stress is shown in Figure 4.2-49. Corrections for mean stress effects should be made by calculating a reduced allowable alternating stress intensity, s , as follows:
a s'=s a a 200,000 - Smax 200,000 - 2Sa where s = the lowest allowable alternating stress instensity a
fra Figure 4.2-49 for the metal ternperature range of interest, psi s
max.
= the maximum calculated stress intensity during the stress cycle, psi s = material yield stress for the metal tmperature range y of interest, psi n e above equations for as ' are used with the following restrictions:
A. > or 2 s s If 2 s#'_
s s=Ysandt$e>equ"a$Y6nisnotused.
B. sy and 2 s, 1 smax., the equation is directly C.
y 2 sa<8y, use smax. " 8y in the equation.
If s ,, > s and
' by applying equation WestripingATmetallimitsareobtainedforanys$rainedbiaxialcasewith (A). W e allowable AT metal range for a fully cons zero mean stress is shown in Figure 4.2-135 for allow 718. To obtain the allowable AT metal fra Figure 4.2-135 the metal ternperature used is the value within the striping range that yields the lowest allowable AT metal range.
(We data base includes ASIM grain sizes of 5 through 10 with an average grain size of 7) 4.2-227f Amend. 68 t- --
The ef fects of the f abrication processes and service environment on the structual integrity of the involved component shall be considered. The effect of grain size on the behavior of Alloy 718 shall also be considered. Some of the UlS liner plates cannot be purchased with an ASTM Grain size finer than 2, theref ore a reduction in the allowable stress range is required. The allowable metal ?T for a required reduction of 35 is shown in Figure 4.2-136 which does not include mean stress effects. When other grain sizes are used the allowable must be adjusted accordingly. Data used to determine the grain size ef fects on the endurance limit are from References 4.2-186 & 187.
I i
l 4.2-227g Amend. 68 May 1982 l
4.2.2.6.5 a==arv of hrmal Stripina namn1ts ne thermal striping evaluation methods an results are sumarized in Table 4.2-68 for the permanent reactor internals. We general conclusion from this table is that all the permanent reactor internals are acceptable for their thermal striping environments. As cited previously and indicated by Figure 4.2-498, the striping values for all reactor components outside the Upper Internal Structure and above the Horizontal Baffle are below the allowable limits for stainless steel.
4.2.3 neactivity control Svstama I
% e mechanical designs of the reactivity control systems consist of the primary control rod systm (PCRS) and the secondary control rod system (SCRS).
Each mechanical system consists of a control rod drive mechanism (PO H for primary, SCRDM for secondary) mounted on the top of the reactor vessel closure head; a control rod driveline (PCRD for primary, SCRD for secondary) connecting the mechanism to the absorber in the core region; and a control asserrbly (PCA for primary, SCA for secondary) located in the core region. We control assembly consists of a movable absorber pin bundle called the control rod and outer duct asserrbly.
Wese control rod systes perform the following functions upon signal from the Instrumentation and Control Systems:
- 1. Provide the primary shutdown (scram) system for the off-normal conditions. f l
- 2. Provide normal operational control. ]
- 3. Provide normal reactor startup and shutdown reactivity control.
- 4. Provide additional margin for control in the event of any anticipated reactivity fault.
Secondarv Control Rod System
- 1. Provide the secondary shutdown system for off-normal conditions.
- 2. Provide reactor shutdown independent of the primary system.
- 3. Provide additional nargin for control in the event of any anticipated reactivity fault.
4.2-228 Amend. 68 w 1
178. A. L. Ward and L. D. Blackburn, " Ductility and Strength of FFTF/CRBRP Structural Materials Irradiated in Various Spectra, Interim Report",
REDL-TME-78-51, August 1978. (Availability, USDOE Technical Information Center).
179. A. L. Ward and L. D. Blackburn, " Ductility and Strength of FFTF/CRBRP Structural Materials Irradiated in Various Spectra, Second Interim Report". HEDL-TME-79-19, July,1979 ( Availability, USDOE Technical Information Center).
180. L. D. Blackburn, " Strength and Ductility of Fast Reactors Irradiated Austenitic Stainless Steels", HEDL-TME-79-21, August 1979 181. TID-26666, " Nuclear Systems Materials Handbook", Volume 1, Property Codes 2206 (E-1) and 3304 (E-1), Hanford Engineering Development Laboratory, 1974.
182. A. Batenburg, H. O. Lagally, R. J. Simko, " Coefficient of Dynamic Friction via Analog Computer Data Reduction", ASLE Paper 80-LC-8B-1, August 1980.
183. T. T. Claudson, " Quarterly Progress Report: Irradiation Effects on Reactor Structural Materials, February - April,1973, "HEDL-TME-73-74, pp. 53-60, May 1973 ,
184. KEI-64-80, Letter from D. D. Keiser to C. E. Gilmore, DOE /10,
" Preliminary Alloy 718 ASME Code Case Package, "May 6,1980.
(Proprietary DOE) 185. R. A.- Leisure, "Effect of Carbon and Nitrogen and Sodium Environment on the Mechanical Properties of Austenitic Stainless Steels",
WARD-NA-94000-5, December, 1980.
186. C. R. Brinkman, " Material Data Base Used for Design and Elevated Temperatures for Long Term and Cyclic Loading", CRBRP/NRC Meeting on HTS Materials and Structures, April 6-7, 1982.
187. Huntington Alloys, Inconel Alloy 718 product brochure, 10M 10-73, T39, 1973 eReferences annotated with an asterisk support conclusions in the section.
Other references are provided as background information.
4.2-330a Amend. 68 May 1982 t
i TABLE 4.2-68 REACTOR INTERVALS THERMAL STRIPING POTENTIALS AND STRIPING FACTORS
' ~
MAX. POTENTIAL NORMALIZED STRIPING MXIMAM MAXIR M BASIS FOR CDeFONENT LOCATION MATERIAL FLUID T SOUR FACTOR TEST BASIS FLUID T KTAL T ACM PTMllLITY 291'F 1:F/A to .79 7-Assembly 230 222 Method #2 UlS instr oentation 1-718 Post B/A 343 F 1 above .58 7-Assembly 199 - Method #1 1-718 345'F F/A to C/A I 7-Assembly 345 270 Method #4 Lower Shroud Tube Upper Shroud I-718 170*F Shroud Tube- 1 Analysts 170 - Method #1 ,
Tube Outlet Plenum Lower Support I-718 343*F 1 above .41 7-Assembly 142 - Method il Plate Liner Skirt Liner 1-718 3430F 1 above 41 IRFM 142 - Method #1 Skirt Ring 11-718 251 2: Radial .62 1RFM 156 -
Method #1
- Horizontal Blacket Avg.
Liner to Peripherial F/A Chimney inlet l-718 343 1 above 45 IRFM 155 - Method #1 Chimney Outlet 1-718 343 2 above .27 BRFM 93 -
Method il IVTM Port Plug I-718 251 2 above 49 IRFM 123 - Method #1 Liner IVTM Port Plug l-718 251 2 above .49 IRFM 123 - Method #1 Lower end IVTM Port Plug 316SS 251 2 above .24 IRfN 60 - Method #1 Above I-718 Seismic Keys 31655 251 2 above .23 1RFM 58 - Method #1 Liner 1-718 251 2 above .27 IRFM 68 -
Method il sheer web 31655 251 2 above 46 1RFM 115 88 Method if Upper Support 31655 343 1 above .15 1RFM 51 - Method #1
+
4.2-459a Amend. 68 }
May 1982
TABLE 4.2-68 (Continued) .
l MAX. POTENTIAL NORMALIZED STRIPlNG MAXIMJM MAXIMJM BA$l$ FOR (DR)NENT LOCATION DMTERI AL FLUID T SOUR FACTOR TEST BASIS FLUID T METAL T ACM PTMllLITY . .
Upper Support 31655 343 1 above .35 lRFM 51 - Method il Plate I-718 326 F/A to Core .61 IRFM & In- 200 - Method #1 Core Liners *
- interstitial terstitial Former Structure Flow Flow Test .!
31655 Leakage to 48 IRFM 70 - Method #1 ,
Horizontal Plate 145 Baffle Outlet Plenum I-718 300 Fl&SA to .83 1RFM 248 192 Method #2 ':
Fl&SA Nozzle 7 Outlet Plena
.~
Method #1 - Most conservative, all cycles are umbrellaed under the maximum fluid T, and this fluid T is compared to the allowable metal T.
Method #2 - All cycles are umbrellaed under the maximum metal surf ace T, and this metal T is compared-to the allowable metal T.
Method #3 - The striping data peaks which are above the endurance strength T are umbrellaed under increments of T ranges, and .
the f atigue damage for these Increments are determined and projected for the reactor life. This fatique damage is added to the creep-f atigus for other transients, then crepared to the damage envelope.
Method #4 - A detalled finite element analysis (time history) of the component for its striping environment is performed.
' Evaluation in Process P
4.2-45 %
Amend. 68 May 1962
\
l l
'3BLE 4.2-69
. ALLOWABLE STRIPING MERL 'ITNPERATURE RANGE KRSIRAIN MEAN 304 EFFECIS AND 316 (< SS INCLUDIF) 1100"F .
Allowable cmean Ac r .
(in/in) (in/in) Surface Range ( Fp) no mean .0014 84 0002 .0013 78 00045 .0012 72 0009 .0011 66 00165 .0010 60 0025 .0009 54 Allowable Striping Metal Temperature Range for 304 and 316 SS Including Mean Strain Effects 0
(< 1100 F) .
1 4.2-459c Amend. 68 mvdlcm
l l
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TIME (a) SODIUM STREAM MIXING (b) MIXING ZONE SODIUM TEMPS .
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CYCLES T- e_ A
.(d) THERMAL FATIGUE ANALYSIS (c) TRANSIENT TEMPS IN STRUCTURE Figure 4.2-130. Thermal Striping Evaluation Steps l
f 70441 4.2-633 -
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TEMPERATURE ('F) i, i
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l Figure 4.21 . Allowable AT metal Range for Alloy 718 for a Fully Constrained
- 1. Biaxial Case (Grain Size 5 and Fir.er)
I 7044 7 i
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i 4.2 639 l
l i
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150 -
>= .
<3 180 1 1000 1100 1200 TEMPERATURE ('F)
(s Figure 4.213{. Allowable AT metal Range Vs. Temperature for Alloy 718 (Grain Size of 2) for a Fully Constrained Biaxial Case 7044-8 . ,
4.2-640 ii
Question CR710.6 Describe testing or analysis performed which assures the leakage will not occur at 2-1/2 Cr - 1 Mo threaded closures in steam generator bolted !
connections during elevated temperature service. l 1
Response
The design of the steam generator modules has been modified to include integral welded steamheads in place of the bolted configuration. Furthermore, the design of the steam generator to piping interf ace has been changed to an all welded system which deletes the need for bolted flanged spool pieces at the inlet and outlet of both the evaporators and superheaters. Therefore, no additional analysis or testing to assure a leak-free system is required.
The PSAR will be modified to reflect the welded steamhead in a future amendment.
QCS210.6-1
Ouestion CS 281.6 in the CRBR primary and intermediate sodium piping system, Fe, Cr, and Mi are dissolved f rom the high temperature regions and deposited in the lower temperature regions because of super-saturation. included in this process of mass transf er is the formation and decomposition of various transition metal and sodium double oxides. Deposition of these mass transfer and corrosion products may cause flow restrictions and loss of heat transport efficiency of heat exchangers. Describe the criteria and bases in your analyses of mass transfer and deposition of corrosion products in the CRBR primary and lHX sodium systems to assure necessary system flow and heat transfer. Include the instrumentation and detection system which will alarm when these limits are exceeded.
Resoonse:
Primarv Sodlum Ploing System l The lHX is designed with an ef fective tube length of 24.21 feet which is 5.12 f t or 33% greater than required f or naninal operation, included in the 33%
excess allowance is a 9% factor for heat transfer degradation due to the deposition of mass transfer products on the primary side of the unit.
Corrosion products go into solution in the core region of the reactor either by direct dissolution or by the formation of soluble oxide canplexes. As the coolant flows through the cooler regions of the system It becomes super- l saturated with respect to these corrosion products and they precipitate out.
Precipitation is expected to occur in the IHX. The deposits will result in some degradation of the overall heat transfer co-ef ficient. This potential problem wss recognized in the early 70's and work was performed to determine the magnitude of this ef fect. A summary of this work is given below.
Two tube-in-shell heat exchangers were available from an ongoing corrosion program. The first had operated in an all stainless steel system with a hot leg temperature of 13250 F for 0.84 years. The second operated in a T304ss/Incoloy 800 system for 1.5 years with a hot leg temperature of 1100 and a high, (28 ppm by analgamation) oxygen level. Heat transf er measurements were made on these heat exchangers and compared with similar measurements on new heat exchangers of identical design. The percentage change in heat transfer coef ficient was determined f or each set of readings.
l QCS281.6-1 I __ _ --- - - - - - - -
The exposure mnditions experienced by the heat exchangers were excessive in that the first one operated with a high maximum hot leg temperature (1325 F) and the second wIth a high oxygen level. Equivalent operating times at r.eactor operating conditions (1100 F mean T ) were calculated at 5.3 years and 11.4 years. Itwasjudgedthatthedep@Sitthicknessesreached equilibrium. Additional increases in thickness are prevented by flow induced shearing of the friable deposits.
The qverage deposit heat transfer resistance was calculated as 8.4 x 10-5 h-ft' F/ BTU.2 The overall heat transfer coef ficient of the IHX design was 1190 BTU /h-f t F. Adding the deposit resistance gives a calculated degradation of 5%. This 9% value was used for the FFTF lHX design and in the CRBRP design.
The offect of corrosion products on pressure drop and flow blockage is not addressed in section 5.3 of the PSAR. Flow blockage is addressed in section 15.4.1.3. Flow blockage in the Core Assembly in 15.4.1.3.1. ' Prevent i on - and -----
Detection' C, (Corrosion Products).
Deposition induced pressure drop in the IHX is not considered to be a f actor because of the large flow cross sections on the shell side.
The effects of corrosion product deposition in the PHTS will result in a very gradual change, if any, in system performance. The PHTS performance is continuously monitored and critical performance parameters are calculated by the plant computer from temperature, flow rate and pressure drop sensors in the PHTS system.
Monitoring sensors in each loop include resistance temperature detectors at the inlet and outlet of the IHX, pressure sensors in the hot and cold legs, and a flow meter.
The performance evaluation for the PHTS is identified in the system procedures and includes:
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! QCS281.6-2 l I
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. l nws: tion m?81.8 Provide the design criteria and bases that demonstrate wastage allowance of the CRBR steam generator tubes, caused by sodium-water reaction products, is acceptable. We analysis should include Na water reaction tenperature and other major variables in the small water leak situation.
Response
he steam generator tube wastage allowance and provisions to accormnodate tube leaks are discussed in the revised PSAR Section 5.5.3.11.4. It should be noted that section 5.5.2.3.4 discusses the function of the tube sheet baffling as wastage baffles. This provides tube protection in the most likely location for leaks.
QCS281.8-1 l Amend. 68
! 17vLMKn
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As a final level or protection against tube leaks in a steam generator, the steam generators and the IlfrS are being designed to withstand the effects of a large soditan water reaction (fMR) . S e ASME Code categories being applied in the design of the steam generators and IHTS piping and components for the large W R event are given in Table 5.5-10.
Se design basis leak (DBL) for the GBRP was selected based upon examination of the physical processes which exist for leak initiation and growth. Se conservatism of this postulated DBL will be confirmed through the IUIR test progran (Ref.12) .
'No types of tests have been reported which provide information on the leak growth mechanism - small scale tests which model effects of a W R on materials, and large scale tests which model a large water leak in a model of a steam generator. Smaller scale sodite-water reaction tests have been done to develop an understanding of the effect of a SWR on neighboring tubes in a steam generator. W ree mechanisms have been identified for leak growth:
self-wastage, impingement, and overheating (mechanical danage from pipe whip, although extranely unlikely, would be considered another mechanism, as discussed later in this section). g f 9 tage has been shown to occur for very small leaks in the range of 10 lb/sec (Ref. 13). Se process is depicted in Figuge 15.3.}.3-1. % e result of this process is a leak size of the order of 10 to 10- lb/sec. which can produce wastage on another tube in the vicinity of the leaking tube.
Wastage can occur on the outside of a steam generator tube from a leak in another tube in the vicinity. 'Iwsts of this mechanism have typically been done by using a water jet directed through sodium to a target material sample.
Water injection rates of approximately 10 4 lb/see to 1 lb/sec have been tested. S e wastage mechanism results in erosion of the target material at maximum rates of 0.001 to 0.007 inches per second (Ref.14, 29) . Se wastage I rate is found to be a function of the water injection rate, tube spacing, CRBRP steam generator tube at these rates could cause a secondary water leak fran the penetration. However, this would require at least 20 secu-4s to penetrate the 0.109 inch thick tube wall assuming an initiating leak of the proper characteristics to produce maximum wastage.
S e size of a secondary water leak resulting from wastage is difficult to quantify since wastage tests are typically done in materials samples rather thanpgessurizedgubes. Se wastage areas observed in tests have ranged fonn 0.1 in to 1.5 in . Failure areas corresponding to the highest observed wastage areas would result in water leak rates corresponding to that of a double-ended guillotine tube failure. However, the entire wastage area would not be expected to blow out. W e wasted areas are typically pit-shaped with the area of the pit decreasing with depth. It would be expected that the small area at the bottom of the pit would fail, yielding a return water leak which halts the wastage. B erefore, while the size of a secondary failure caused by wastage is difficult to predict, it is expected to be analler than the leak rate corresponding to a double-ended guillotine failure.
5.5-24a Amend. 68 Mar 1982
5.5.3.11.4 cmnatibility with ennlants he decarburization kinetics of 2-1/4 Cr-1 Mo base metal and subsequent strength loss are slow enough so that the selection of 2-1/4 Cr-1 Mo steel as steam generator tubing can be made with only a small design stress adjustment.
2-1/4 Cr-1 Mo steel creep rupture properties are insensitive to carbon content util the level drops below 0.03% carbon. W is low carbon level will not be reached during the thirty-year tubing design life. W e bulk carbon loss is predicted to be 0.04% at the 965 F design temperature based on experimental decarburization data (Reference 5). Since the initial carbon content is 0.07
<%10.11, the carbon content will not drop below 0.03%.
A minim m wall thickness of 0.109 inches has been specified for the CRBRP steam generator tubes. Of this wall thickness, a minimum thickness of 0.07/
inches has been specified for strength with the balance of 0.032 inches allocated for total corrosion allowance of the sodium and water sides. %e steam generator turbine material is 2-1/4 Cr-lMo steel.
% e possibility that wastage fra a small sodium water reaction (or, more likely, rapid pressure rupture) would cause propagation of a leak to adjacent tubes has been considered in the definition of the design basis steam generator leak event. (See PSAR Section 15.3.3.3)
A small sodim water reaction event that does not result in replacement of the affected steam generator module may also produce wastage of steam generator tubes. We wastage near the leak site (combined with other corrosion) may be equal to or greater than the 0.32 inch corrosion allowance so that tubes will have to be plugged.
Based on the results of the Large Leak Test Rig (LUIR) tests (Reference 29),
this amount of wastage is expected to affect only those eighteen tubes within 2 rows of the leaking tube and only a portion of those tubes. Based on the same test series, wastage of tubes beyond the 2nd row is expected to be significantly less than the 0.032 inch corrosion allowance. %is is due to the large excess of sodium present which dilutes the SWR reaction products.
After each LU1h/NR test the steam generator tube wastage was measured using a boreside ultrasonic testing (W) device. Following the LLTR Series I program of 5 WR tests and 1 inert gas test, the test steam generator was destructively examined and the wastage directly measured. Excellent agreement was found between these post-test measurenents and the inter-test W measurements. Similar W equipnent will be anployed for CRBRP steam tube inspection and would be sensitive down to 4 mils wastage on a routine basis.
S e plant will be shutdown on the basis of a confirmed leak. At shutdown the affected mit will be examined, using helium sniffing te:hniques to locate the leaking tube (s) and W techniques to determine amount o'. wall thinning by wastage in surrounding tubes. Leaky tube (s) will be plugged; plugging of adjacent tubes will depead on the extent of wall thinning as determined by W.
Se crib for tube plugging will be specified in the FSAR.
5.5-32c Amend. 68 r?wKTM
Pagek(82-0227)[8,22] - #28 ,
. l Because of the conclusive test results mentioned above and because tubes will be volumetrically inspected and plugged based on actual wastage, quantitative analysis of wastage as a function of ta perature or other variables has not been performed; nor will it be usd as a basis for design or operating specifications.
No specific protection is required for protecting Type 304 SS or 2-1/4 CR-1 Mo steels against intergranular attack, stress-corrosion or general corrosion, provided that specified sodiun purity is maintained.
In water or steam, carbon steel and 2-1/4 Cr-1 Mo steel are susceptible to caustic gouging and possibly caustic stress corrosion cracking. Maintaining the feedwater and steam drum purity levels as stated below will prevent these forms of localized attack. For normal operation other than start-up conditions, the feedwater and steam drum purity will be specified as follows:
Steam Feedwater Tururities Feedwater Drum suspended Solids Pm -
0.1 Dissolved Oxygen PPM .005 -
l Pm 0.1 Silica Iron as Fe Pm .01 -
Copper as Cu Pm .002 -
l Hydrazine Pm .005 .015 -
Chlorides Pm -
.015 Sodium PPM .001 .006 Sulfate Pm -
.015 l pH @ 77 F 8.8-9.2 8.8-9.2 Condgetivity (After Cation Removal) 0.2 1.0
@ 77 F micro-mho/cm Limited duration operation with impurity levels above specified limits is allowable for periods not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in special instances. 'Ibese special instances are defined to include condensate polishing system perturbations, such as those imediately associated with a termination of regeneration.
Corrosion impurities may enter the feedwater system through condenser leakage and/or poor makeup water. 'Ib guard against damage from such sources, the feedwater and steam drum water are maintained at levels within stated limits by full flow deineralization and continuous steam drum drainflow (blowdown) at a nominal rate of 10% of full power steam flow (See Section 10.4.7).
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5.5-33 Amend. 68
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27 J. C. Amos, et.al, " Evaluation of LLTR Series II Test A-3 Results."
General Electric Advanced Reactor Systems Department, November 1980, Prepared for U.S. Department of Energy under Contract No.
DE-ATc3-76SF70030, Work Package AF 15 10 05, WPT No. SG037
- 28. J. O. Sterns, " Metallurgical Evaluation of the Modular Steam Generator (MSG) after LLTR Testing," ETEC-78-12, Sept.1978.
29 D. A. Greene, J. A. Gudahl and P. M. Magee, "Recent Experimental Results on Small Leak Behavior and Interpretation for Leak Detection",
CONF-780201, Vol. 1, paper No. 12 (First Joint U.S./ Japan LHFBR Steam Generator Seminar), February 1978.
' References annotated with an asterisk support conclusions in the Section.
Other references are provided as background information.
5.5-35b Amend. 68 Mag 1982
tumatlan CS 2A1.9 Describe the sample and instrument readings and the frequency of measurenents ;
that will be performed to monitor the feed water purity and need for ,
condensate cleanup system demineralizer resins and filter replacement. State ;
the chemical limits and precaution to be taken to protect steam generator l tubes against excessive corrosion and deposition. Also, provide the basis of establishing the chenistry limits.
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Response
PSAR Section 5.5.3.11.4 presents the feedwater and steam drum purity i established to protect the steam generator tubes against excessive corrosion j and deposition. PSAR Section 5.5.3.11.4 also adds additional information !
relative to monitoring and controls. h following major factors provided the basis for establishing the chenistry limits: .
j
- 1. Because of the relatively low evaporator recirculation ratio in CRBRP, j it was recognized early in the program that the OtBRP water chemistry !
limits would need to be similar to those limits which extensive j experience in the fossil fired boiler and nuclear steam generator !
industries with once-through designs had shown to be required. }
Basically, this requires the use of all volitile treatment (AVT) consisting of a pH adjustment agent (typically anonium hydroxide) and an oxygen scavenging agent (typically hydrazine). 'Ihe concentration i of AVT agents is controlled in the feedwater to minimize corrosion in both the feedwater train and in the evaporator recirculation loop. ;
'Iherefore, the then existing industry AVT chemistry requirenents were '
established as the basis for CRBRP chemistry control. l 1
- 2. h se chemistry requirements were further refined to address the ;
particular needs of OtBRP relative to materials, i.e., because of the 90/10 copper-nickel condensor, a 0.002 ppn copper concentration was f specified. 'Ihis low limit minimizes the potential for transport of copper to the evaporator tube internal surfaces where it would cause excessive tube corrosion. :
- 3. 'Ihe low recirculation ratio in the evaporators results in DNB in the !
2-1/4 Cr-1 Mo evaporator tubes. 'Ihis requires close control of the '
sodium ions to prevent stress corrosion cracking problens and close control of the chloride and sulfate ions to prevent "under deposit" I corrosion. For example, the feedwater sodium ion concentration is l maintained at 0.001 ppm maximum to achieve 0.006 ppn maximtzn in the I
recirculation loop. Similarly, the chloride and sulfate ions are -
maintained at low values in the feedwater to achieve a 0.015 ppn i maximum for both species in the recirculation loop.
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i QCS281.9-1 Amend. 68 May 1982
In order to meet the evaporator water chenistry requirements described above, requirenents for condensate systen denineralizer resin regeneration and/or replacanent, continuous monitoring / recording and grap sampling of the Condensate Polishing effluent have been established as follows:
Design Limit Max. Allowable Operation Above Grab Inpurities 5% Power Monitoring Sample Total Suspended 16 ppb None Yes Solids Silica (SiO2) 5 ppb Continuous Yes Iron (Fe) 5 ppb None Yes 1 Copper (Cu) <l.5 ppb None Yes Sodium (Na) 1 ppb Continuous Yes Chloride (Cl) 2.5 ppb Continuous Yes Cation Conductivity 0.2 unho/an Continuous Yes at 77 F QCS281.9-2 Amend. 68 DM MTR
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Because of the conclusive test results mentioned above and because tubes will be volumetrically inspected and plugged based on actual wastage, quantitative analysis of wastage as a function of temperature or other variables has not been performed; nor will it be usd as a basis for design or operating specifications.
No specific protection is required f or protecting Type 304 SS or 2-1/4 CR-1 Mo steels against intergranular attack, stress-corrosion or general corrosion, provided that specified sodium purity is maintained.
In water or steam, carbon steel and 2-1/4 Cr-1 Mo steel are susceptible to caustic gouging and possibly caustic stress corrosion cracking. Maintaining the feedwater and steam drum purity levels as stated below will prevent these forms of localized attack. For normal operation other than start-up conditions, the feedwater and steam drum purity will be specified as follows:
Steam Feedwater Imourities Feedwater Drum Suspended Solids PPM -- 0.1 DI solved Oxygen PPM .005 -
Silica PPM -- 0.1 fron as Fe PPM .01 -
Hydraz ine PPM 005 .015 -
.015 Sodium PPM .001 .006 Sulfate 0 PPM --
.015 pH 8 77 F 8.8-9.2 8.8-9.2 Conductivity ( Af ter Cation Removal) 0.2 1.0 8 77 F micro-mho/cm Limited duration operation with impurity levels above specified limits is allowable for periods not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in special Instances. These special Instances are defired to include condensate polishing system perturbations, such as those immediately associated with a termination of regeneration.
Corrosion Impurities may enter the feedwater system through condenser leakage and/or poor makeup water. To guard against damage from such sources, the feedwater and steam drum water are maintained at lesels within stated limits by full flow demineralization and continuous steam drum drainflow (blowdown) at a nominal rate of 10% of full power steam flow (See Section 10.4.7).
5.5-33 Amend. 68 May 1982
L To determine the feedwater quality, continuous analysers with alarms are provided to sample conductivity, dissolved oxygen, hydrazine, turbidity, pH, sodim , chloride and silica. Continuous samples of steam drum downceer water and periodic samples of drum drain (blowdown) water are monitored for conductivity, sodium, silica and pH. h downcomer continuous sample monitors )
are also alarmed if out of specification conditions occur. h condenser hot-well is monitored for conductivity and sodita ions to guard against condenser leakage. 'Ihe denineralizer effluent is guarded against inpurities break-through by in-line measurenents of silica, conductivity and sodium.
Finally, the feedwater train is monitored downstream of the deaerator for pH and oxygen content to prevent potential corrosion of this portion of the steam systs. An alarm is coupled with the most critical in-line measurenents to signal departure from specified levels.
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I Ouestion CS 281.12 (9.8)
The impurity Monitoring and Analysis System consists of primary heat transport system (PHTS), ex-vessel storage tank (EVST), Intermediate heat transport system (IHTS) sodium characterization, f uel handling cell (FHC), and lHTS Cover Gas Sampling Systems. Provide the chemical and radiochemical limits for the sodium and the cover gas analyses, in addition, describe the method, sampling procedures, and frequency of sampling.
,Essponse The chemical and radiochemical limits for sodium and cover gas systems are listed in Table CS 281.12-1. The methods for sodium sampling are discussed in PSAR Section 9.8. For the required sampling frequency for sodium, see the response to Questions CS 281.5 and CS 281.10. Cover gas sampling methods and frequencies have not yet been finalized. The procedures to be utilized for sampling will be developed prior to plant operation and will be described in the FSAR.
TABLE CS 281.12-1 Sodium imourftv Primarv Intermediate J300[
2 ppm 2 ppm 5 ppm 0.2 ppm 0.2 ppm 0.4 ppm
,U 10 ppb 10 ppb 10 ppb C 0.7 ppm 0.7 ppm 0.7 ppm Other 10 ppm 10 ppm 10 ppm Cover Gas Imourftv Primary Intermediate fyjI flig H
2 10 ppm 8 ppm 8 ppm -
0 2 ppm 10 ppm 10 ppm 25-75 ppm 2
2 100 ppm 15 ppm -
6% vol 4 I 25 ppm 25 ppm -
00 (4 ppm - - -
H 2O 2 ppm 8 ppm 8 ppm 25-75 ppm i Radiochemical limits for the reactor cover gas during continued reactor operation are provided in PSAR Section 16.3.2.3. Cover gas radiochemical limits during ref ueling operations are provided for the reactor in PSAR Section 16.3.10.3.2 and for the EYST and FHC in PSAR Section 16.3.10.3.1.
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QCS281.12-1 Amend. 68