ML20071J405
| ML20071J405 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 04/21/1982 |
| From: | Longenecker J ENERGY, DEPT. OF |
| To: | Check P Office of Nuclear Reactor Regulation |
| References | |
| HQ:S:82:020, HQ:S:82:20, NUDOCS 8204270246 | |
| Download: ML20071J405 (12) | |
Text
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Department of Energy Washington, D.C. 20545 Docket No. 50-537 e
HQ:S:82:020 9
RECElygg APR 211982 gh APR261982s $
Mr. Paul S. Check, Director sazawaar Yr
"* **ea CRBRP Program Office Office of Nuclear Reactor Reg:^1ation 9
U.S. Nuclear Regulatory Comission f
Washington, DC 20555 4
to
Dear Mr. Check:
RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION - CHEMICAL ENGINEERING
Reference:
Letter, P. S. Check to J. R. Longenecker, "CRBRP Request For Additional Information," dated March 23, 1982 This letter formally responds to your request for additional information contained in the reference letter.
Enclosed are responses to questions CS 281.1, 2, 3, 4, 5, 7, 10, 11, and 13 that will also be incorporated into the PSAR in Amendment 68, scheduled for April 30.
The remaining questions from the reference letter (CS 218.6, 8, 9, and 12) will be provided under separate cover by April 23.
Sincerely.
1 John R. Longenecker anager Licensing & Environmental Coordination l
Office of Nuclear Energy Enclosure l
cc:
Service List l
Standard Distribution Licensing Distribution l
dd s.
8204270Ng
(
A
82-0184 Ouestion M 81.1 f4.2)
Degradation of CRBR fuel cladding may be caused by (a) selective leaching of nickel and chromium due to soditn exposure, (b) formation of double oxides and intergranular attack by sodium corrosion, (c) loss of mechanical strength due to loss of carbon and nitrogen, or (d) fission product attack on the cladding internal surface.
In your cladding strain and Cumulative Damage Function Analysis of the cladding integrity, you assumed uniform cladding wastage.
Provide additional cladding performance analysis including the effect of localized attack and the loss of strength due to interstitial transfer and the formation of a ferrite layer on the cladding surface, to d eonstrate that the cladding integrity will be nuintained under normal operating and design basis accident conditions.
Response
As described in Reference 58 of PSAR Section 4.2, the CDP procedure incorporates those phenomena which either degrade the strength of the cladding or which reduce its effective load bearing cross section. Specifically, the following comg>nent effects are considered as explicit functions of the prevailing local environmental conditions:
(a) Sodium corrosion (i.e., the physical loss of cladding to flowing sodium).
(b) 'Ihe formation of the ferrite layer at the cladding / sodium interface (i.e., a substrate devoid of substitutional alloying elements, e.g.,
Cr and Ni).
(c) Fission product attack at the inside surface of the cladding.
(d) The change in the cladditej's bulk content of interstitial elments (i.e., C and N).
In application, the change in the interstitial content is taken to affect the cladding's mechanical properties; the other phenomena serve to reduce the effective load bearing cross section of the cladding.
Since the above phenomena have already oeen factored into the cladding performance calculations, additional analyses are not required.
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0281.1-1 Amend. 68 DwD1 MX:R -
o Ouestion N 81.2 (4.2)
In the PSAR, you state that the design objective of the primary control rod system (PCRS) is to achieve a service life of minimum of 17,000 ft. travel and 732 scrams. To inprove the resistance to wear and friction, carbide coating was applied on certain PCRS component surfaces. Describe the design criteria and limits established for maximum carbon transfer in the primary coolant system due to the presence of carbide coating. Include in your analysis the problem of carburization of the fuel cladding surface which leads to cladding enbrittlement due to the use of chromita carbide on the coupling head of PCRS.
Response
%e principal use of chromium carbide coatings in the CRBRP is on renovable assenbly load pads, as described in PSAR Section 4.2.1.2.2 (p. 4.2-48). %is coating is used on fuel, blanket and control assenbly ducts. Reference is also made to the possible use of chrome carbide in the secondary driveline/ control assenbly coupling in PSAR Section 4.2.3.1.7 (p. 4.2-254),
but the potential surface area exposed to flowing sodita in this application would be small in comparison with the load pad surfaces. %is coating is not used in primary driveline couplings.
%ere are no design criteria and no " limits" identified or established for carbon transfer in the primary coolant system that relate specifically to carbide coatings present on certain parts of the PCRS. % e rationale for not considering carburization from this source as a significant issue is given below.
%e concern over the possibility that carbide coatings would produce carburization of core ccznponent arises, not because of the chromium carbide itself, but largely because of free carbon that may be present in the binder material.
l Free carbon is present in the raw material fed into the detonation gun.
During the coating process, noch of this is oxidized and lost to the 0
atmosphere. Tests for 2000-hours in vacuum have demonstrated that at 1160 F, the CiCr ratio in the coating is sufficiently high to favor the stabilization of C.
Similar exposures in sodium, however, produced CRnC6 as the ibtcarbidepresumablythroughsomeinitiallossofUiefreecarbon.
pr e
QCS281.2-1 tasA @@
After each exposure several carbide phases were present; in the former case 3
accomted for only 10% of the carbide total, while in the latter it w
t 70% of a M23 6'-M C,- g C2 mix.
C 73 Although definitive experiments have not been conducted to determine whether or not significant core c a ponent carburization is possible through carbon release fr m the chrm i m - carbide coated regions of the PCRS, the following statements may be made:
o
'Ihe surface area of the PCRS carbide coatings is a negligible fraction of the total surface area exposed to sodium in the primary systen. Carbon release from these areas would not be discernible fr m the general carbon release within the core.
o Carbon loss fr m the coatings is not likely to be at a level that would increase the carbon activity of the sodim to a level that would produce carburization on the fuel cladding. '1he coating contains a substantial amount of free chr mium from the nichrome binder and continued breakdown of carbides and other types is more likely to produce additional E
ithin the 3-mil thick coating, than to result in significant carbon n
r81 se to the sodium.
General Reference 0281.2-1:
G. A. Whitlow, R. L. Miller, S. L. Schrock, and P. C. S. Wu, "Sodim Cmpatibility Studies of Low Friction Carbide Coatings for Reactor Amlication", Corrosion V30, No.12 p.p. 420-426 (1974).
1 Ocs281.2-2 Amend. 68 n m astsuwa
82-0184 Question N>81.3 (4.2)
Carbides formed by 2 and Ti in the alloy 718 are thermodynamically more stable than the chranium carbides formed in the type 316SS fuel cladding.
Provide analysis that the presence of the Inconcel 718 in the OBR core region would not lead to decarburization and subsequent loss of strength of the fuel cladding.
Response
It is true that from a thermodynamic standpoint 2 and Ti carbides will form in preference to Cr carbides; however, there is essentially no " free" W or Ti in Alloy 718. 'Ibese elenents are already tied up in a very stable form as 2C, N 2 or as/and J" the coherent precipitates (Ni3 (Al Ti 2) and Ni 2, x
respect vely) responsible for strengthening the alloy 0
'Ihe fuel cladding exposed to hot spot tenperatures (1250-1300 F) is expected ~ -' '- ~ ~
to experience decarburization to surface levels of about 50 gin. 'Ibe rate of carbon loss from the cladding will be controlled by the brcakdown of M C in 22 6 the 316SS which is virtually independent of the sodium carbon activity at the low activities expected in CRBR. 'Ihe effect of adding another carbon sink, to the already large sink available (i.e., cold leg regions, cold trap, IHX, mass transfer deposits) would have a negligible effect on carbon loss from the cladding.
'Ite supportive data on expected sodium carbon activities and decarburization rates can be found in the Nuclear Systens Materials Haniook.
In conclusion, Alloy 718 cannot be considered as a carbon sink in the CRBR primary heat transport systen that would in any way influence decarburization of the fuel cladding.
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i 0281.1-2 Amend. 68 April 1982 l
82'-0184 rk=ntion CS 281.4f5.3)
On page 5.3-16 of the PSAR you state that sodium leak tests have shown corrosion rates of steam generator tubes by Na-water reaction product; e.g.,
0 NaOH, to be 0.12 mils /hr at 1050 F,1000 vppn H O and 1.2 v/o or less, and 2
this accelerated corrosion from the presence of water vapor sodium is acceptable in that propagation of a leak from corrosion at this rate will not significantly affect plant capability to safe shutdown and maintain safe shutdown conditions. Provide the technical basis and analysis for the above statement, and denonstrate the validity of leak before break criteria in your analysis.
Response
a) '1he steam generator tubes are not addressed in the section referenced.
Information regarding the steam generator tubes is found in PSAR Section 5.5.
Additional information regarding the effects of sodiunt-water reaction products on the steam generator tubes can be found in the response to Question 281.8.
b) 'Ihe quote from Section 5.3 on corrosion rates deals with the primary heat transport system. WARD-IH)185, " Integrity of Primary and Intermediate Heat Transport system Piping in Containment" provides the basis for the Project's conclusion that rapid catastrophic failure of the in-containnent HTS piping need not be considered in the plant design basis. Section 2.2 provides the Piping Integrity Rationale; Project standards for Heat Transport Systen Piping are sunnarized in Section 3.1.5, and Section 5.5 and provide additional information dealing with piping corrosion due to leakage. 'Ihe rationale for the leak before break concept is addressed throughout WARD-D-0185.
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CS281.4-1 A mend. 68
Question CS 281.5 (5.3. 9.3)
'Ihe heat transport system liquid metal (Na and NaK) chemistry is selected to minimize corrosion and to ensure the fuel cladding integrity and to prevent radiation leakage.
In the PSAR, you state that periodic analysis of the liquid metal chemistry is performed to verify that the Na and NaK quality meets the proposed specification.
'Ihe liquid metal purity is maintained by the use of cold t. raps.
Describe the frequency and chemical and radiochemical analysis to be performed for liquid metal analysis and the criteria for cold trap replacanent to ensure the liquid metal quality meets the proposed specifications.
Response
Sodium PirIS and IHTS sodium samples will be taken by the multipurpose sampler (MPS) for laboratory chemical analyses. Chemical analysis will be performed monthly with the following exceptions:
1.
During periods of anticipated changes (initial fill; refueling, maintenance activities) in impurity levels, the frequency will be increased.
2.
When it has been established that an elenent impurity level is no longer changing after several years of plant operations, the frequency of analysis may be reduced.
Indication of oxygen and hydrogen concentrations by the plugging meter technique will be obtained routinely once a day, or if there is no apparent change over a long period of operation, at a decreased frequency.
'Ihe determination of oxygen, hydrogen, and carbon concentrations is by the equilibration method in which metal tabs or wires are exposed to flowing sodium in the MPS for a time sufficient to establish equilibrium with respect to the impurities. Subsequent measurenent of the impurity concentration in the wire or tab is used to determine the impurity concentration in the sodium.
Uranitun and plutonium concentrations in liquid metal are determined fluorometrically and by alpha assay, respectively, using a vacuum distillation residue. Tritium concentrations are measured using liquid scintillation counting techniques.
W concentrations of various elenents (Ca, Co, Cr, Cu, Fe, Li,_Mg, Mn, Mo, Ni, K, Cs, Rb, Si, B) are determined using standard techniques of atcanic absorption or flane emission spectrophotcanetry.
Several of the analysis are performed with a sodium distillation residue.
CS281.1
.m
April 1982
% e current basis for developing CRBR procedures for chemical and radiochemical analyses of liquid metal is RDT Standard F 3-40T which contains typical analytical procedures.
Forced Circulation Sodium Cold TraD Reolaca*=nt Criteria
%e criteria for forced circulation cold trap replacement is an increase in cold trap 4p at design flow. mis limiting 4p is currently estimated to be 22 psid. When this limiting Ap is reached, cold trapping may be continued at decreased flows provided impurity limits are not exceeded.
tiaK It is expected that no sampling will be necessary because the diffusion cold traps have been sized to clean up the systen initially and to remove the oxygen and hydrogen introduced when each system is opened four times during the life of the plant. If operating experience indicates leakage or additional contamination, a decision would be made at that time to take a sample. Provisions will be available for obtaining NaK samples from the NaK storage vessels.
If this sample indictes that the cold trap is not maintaining acceptable oxygen and hydrogen concentrations, a decision may be made to replace the diffusion cold trap.
Chenical analysis of the NaK samples would be performed in the laboratory with methods similar to those for sodium. %ere is no plan for radiochemical analysis of the NaK.
CS281.2 me
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82-0184 Danmnse to NRC Onantion N81.7 Questien Provide a Failure Mode and Effects Analysis (FMEA) to demonstrate that the sodium dup subsyste of the OBR steam generating system is designed such that no single failure of the isolation and dump equipnent. causes the loss of shutdown heat removal capability, including that a single failure does not cause two sodium dap valves in the same dump path to open.
Response
'Ibe IIns has three independent loops, any one of which can transport all the resi&al and decay heat from the reactor to the steam generator systen. 'Ihere are two normally closed valves in series on each dump line. A single failure of one dump valve actuator would not cause the loss of an IHTS loop. Failure of both dump valve actuators, causing the valves to open, would be necessary in order to lose an IHTS loop. CRBRP is designed to acce==nrhte failure of two sodium &mp valves in series and retain the capability in the remaining loops to perform its shutdown heat removal function.
A failure modes and effects analysis performed on the IHTS has considered the following sodium dump valve failure modes:
o Failure to open or close due to solenoid failure or loss of compressed air.
o Opens falsely due to operator error or I&C failure.
o Loss in integrity of the sodium boundary at the valve.
'Ihe sodium dump valve control for each loop is separate and independent.
Within each loop the control of the upstream sodium dump valves is independent of an IHTS loop, interlock features are provided on the dump valve operation.
IHTS &mp valve controls are located on the Main Ccmtrol Panel with the following features:
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o Push-button controls are located on the Main Control Panel by loop and l
properly designated.
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'Ihere will be separate protective covers over each of the two controir, for each loop to preclude inadvertant operation.
o An interlock requires operation of downstream valves before upstream valves can be opened.
(Interlock bypass capability exists to allow valve l
l cxercising for maintenance and testing.)
W se features preclud2 the simultaneous draining of all three loops.
CS281.7-1 Amend. 68 DwNLJdYR
rs=ntion CS 281.10 (9.1)
Describe the chemistry and radiochemical limits, monitoring frequency, and criteria for cold trap replacanent to ensure the soditan purity in the spent fuel storage pool. Provide the basis for establishing these limits.
Response
'1he limits for sodium purity in the " Spent Fuel Storage Pool" which is known as the Ex-Vessel Storage Tank (WST) are 5 ppn oxygen and 0.4 ppn hydrogen. 'Ihe basis for establishing these oxygen and hydrogen limits is corrosion, which is tenperature dependent. 'Ihe corrosion effects due to these impurity levels for the WSr 0
(operating at 500 F) are comparable to those for 2 ppn 02 and 0.2 ppn 0
H for the reactor sodium (operating at 1,000 F).
2
'Ihe radiochemical limits currently inposed for the WSr sodium are 10 ppb Pu and U.
'Ihese limits are established to minimize potential maintenance problens resulting from contamination of components and to minimize contamination resulting from potential sodium spills.
WSr sodium samples will be taken by the multipurpose sampler (MPS) for laboratory chemical analyses. Chemi.M analysis will be performed monthly with the following exceptions:
1.
During periods of anticipated changes (initial fill; refueling, maintenance activities) in impurity levels, the frequency will be increased.
2.
When it has been established that an elenent impurity level is no longer changing after several years of plant operations, the frequency of analysis may be reduced.
Indication of oxygen and hydrogen concentrations by the plugging meter technique will be obtained routinely once a day, or if there is no apparent change over a long period of operation, at a decreased frequency.
'Ite criteria for forced circulation cold trap replacement is an increase in cold trapa p at design flow. ' Itis limiting A p is currently estimated to be 22 psid. When this limiting A p is reached, cold trapping may be continued at decreased flows provided inpurity limits are not exceeded.
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April 1982 Question CS 281.11 (9.8)
Provide any analysis and any experimental results to denonstrate that the plugging meter, on-line inpurity monitoring device, can provide quantitative results on sodim chemistry. Include correlation of these results with those obtained by the vanaditan wire equilibration device (WED) technique.
Response
Plugging temperature indicators (PTI) have been in use for several years as on-line monitoring devices for sodium systems at EBR II and FF'IF. A PTI of the type planned for use in CRBR was developed and tested by HEIL and is currently in successful operation at the FfW.
Table CS281.11-1 contains references to documents published on various experimental results and other studies. 'Ihese reference h==nts provide the basis for utilization of the plugging meter method to provide sodium chemistry information.
'Ihe PII is utilized to indicate the saturation (plugging) temperature for all the species (impurities) present in the sodium stream which precipitate out at that temperature. 'Ihe plugging ta sture cannot be correlated directly to the concentration of any spec. ic impurity such as 0 r1 However, as long as the plugging temperatura is 2
below the max permissible, then the corresponding oxygen and hydrogen impurities in the soditan stream do not exceed the specified limits of 2.0 ppn and 0.2 rpn, respectively.
TABLE CS281.ll-1 Reference Documents for Plugging Temperature Monitoring of soditan Systems i
1.
HEDL 'IME 73-41, Prototype A@lication Loop " PAL" l yr of Operations, Interim Report, J. J. McCown (3-1973).
2.
HEDL 'IME 74-24, Prototype A@lication Loop " PAL" 2 yrs of Operations, J. J. McCown (3-1974).
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82-0184 n eation N 81.13 7he NRC position and clarifications on post accident sampling capability of light water reactors is presented in Item II.B.3 of NJREG-0737. Provide the design bases and criteria for a systen that will provide post accident sampling and analysis capability for the CRBR which will be equivalent to the functional requirements for light water reactor plants.
BREDQDER Appendix H of the PSAR provides CRBRP's evaluation of and resolution to the requirements delineated in NUREG-0718. NUREG-0718 defines requirenents of NUREG-0737 applicable to Applications for Construction Permits.
Item II.B.3 is addressed in the A[pendix H.
Means to assess the degree of core damage in the event of an accident will be provided. However, the approach to be utilized has not yet been finalized.
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CS281.13-1 Amend. 68
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