ML20042C712

LACBWR Primary Piping & Reactor Vessel Leak Detection Sys Performance

ML20042C712
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	02/28/1971
From:	Steele T DAIRYLAND POWER COOPERATIVE
To:
Shared Package
ML20042C714	List: ML20042C712 ML20054H166
References
DPC-851-21, NUDOCS 8206230061
Download: ML20042C712 (15)

Text

{{#Wiki_filter:* DPC-851-21 LAC 8WR PRIMARY PIPING AND REACTOR YESSEL LEAK OETECTION SYSTEM PERFORMANCE T. A. Steele Health & Safety Engineer 4 l l i Dairyland Power Cooperative 2615 East Avenue South La Crosse, Wisconsin 54601 Feb ruary, 1971 206230061 820614 DR ADOCK 05000409 l PDR

DPC-651-21 NOTICE This report was prepared as an account of work sponsored by the United States Government. Nolther the United States nor the United States Atomic Energy Commission, nor any of their omployoos, nor any of their contractors, subcontractors, or their employoos, makes any warranty, express or Impiled, or assurr.os any legal liability or responsibility for the accuracy, completeness or usefulness of any information, apparatus, produce or process disclosod, or represents that its uso would not in f ringo privately-owned rights. Ii

DPC-851-21 LIST OF FIGURES Figure Pace 1 Lower Reactor Cavity Continuous Air Monitor Acti vi ty vs. Events in October 1969............. 3 2 Lower Reactor Continuous Air Particulate Activity as a Function of Time.................. 5 3 Lower Cavity Filter 30 Minute Decay Gross Bota, Gamma..................................... 6 4 Lower Cavity Filter 24 Hour Decay Gross Beta, Gamma..................................... 7 5 Co.7-nntration of Tritium in Primary Water and Lower Cavi ty as a Function of Ti me.......... 8 6 Lower Cavi ty Ca l cu l ated Leak Rate............... 9 l III l

DFC-851-21

SUMMARY

Tho LACOWR leak detection system has boon described in Amendment No. 3(I) and its Addendum (2) to the LAC 8WR Operating Authorization DPRA-6. The resul ts f rom the system havo boon reported in the LAC 8WR Monthly Reports during operational periods since April,1969(3). The system detected the prosonco of the leak in the foodwater nozzlo which occurred in October of 1969(4,5) This loak was subsequently located in November,1969(6). Oper-ational data obtained sinco May, 1970, is examinod and discussed. Description of System The system has boon designod to detect small leaks which could originato in the LA_COWR reactor vossel and associated pressure piping in the contain-mont building. This is accomplished using throo air particulato monitors which monitor air surrounding di f feront portions of the primary system. These monitors aro constantly running and a continuous roadout is available for immediato and subsequent analysis. The areas monitored are the containmont building atmosphere, the forcad circulation pump cubiclo exhaust and the lower or upper reactor cavities. Analysis of the data f rom those air monitors will glvo an Indication of the general area of a very small leak in the stoam gonoration system. In addition to the air particulato monitors, samples of moisturo in the air are collected from the various areas for tritium analysis. Since those samplos havo to be collected and subsequently analyzed, this data ylolds a wook to week trond of the Integrity of the steam generation system. The constant air monitor, which is used to detect any leakage insido of the biological shielding (lower cavity), draws 5 cfm of air from the lower reactor cavity through the cavity vent. The lower section of the reactor vossol, associated foodwater and forced circulation piping and the steam lino piping are located in this air volume behind the biological shleid. The lower cavity is pressurized with 10 cfm of instrument air which main-tains the cavity at 0.1 inchos of water positive with respect to the contain-mont building to provont any in-loakago to the cavity f rom outsido sources. Through selectivo valving, the upper reactor cavity can be monitored as desired to detect any leaks in the reactor head flango and instrument piping. Tho upper cavity is not pressurized but it is negative with respect to the lower cavity by 0.6 inchos of water. The constant air monitor which is used to detect any leakago in the contain-mont building can be located at any of the four floor elevations. This monitor will detect leakago f rom primary system piping outside of the biolog-leal shield and pump cubicles and inside of the containment building. The constant air monitor which is utilized to detect leakago in the forced circulation pump cubicles draws its samplo f rom the exhaust duct of the DPC-851-21 <;bicles. Sinco the exhaust air from the cubicles also includes air from

• he containmont buliding, the activity contribution from the containment building has to be considered in the analysis of this data. This monitor has detected soveral valvo packing leaks in the forced circulation pump cubicios.

Since the radiation levels do not permit routine entry to this aroa, this monitor is useful in the early detection of water leaks which could originato in this area. Amondmont tio. 3(1,2) discussos the sensitivity of the system to detect very smal l leaks. The reported sensitivity for detecting a reactor water phase leak is 1 ml/ min. The lower detection limit for a steam phase or foodwater leak is reported to be 10 ml/ min. Those sensitivities have been accopted as being adequate for the protection of the reactor and associated piping f rom a gross f ailuro generating f rom a very small leak. Dotorminat!on of a Very Small Leak in the Primar/ System in October,1969 This system dotected thgrosence of a very small leak in the feedwater system in October, 1969 The leak was located using a pipe smearing technique in tiovember, 1969(6) The detection of this leak led to the repair of the primary system sensitized stainless steel nozzles during the tiovember,1969 to Apri l,1970 parlod(7). Figuro 1 is a plot of the lower reactor cavity air activity vs. time during the period October 9,1969 to October 14, 1969 The data f rom October 9 Indicates typical data f rom the previous operational period. The activity on the filter builds up to an equilibrium value and is then constant. Dur-Ing this period the filter was being crunged daily as is noted on Figure 1. LAC 8WR scrammed on October 10, 1969 from the initial Pressure Regulator (IPR) closing down which caused a high pressure trip. Upon return to power on October 11, 1969, the cavity mnitor increased to 11,000 cpm, a level which had not previously boon reached duris.g 100% power operation. Examination of the filtor removed for analysis on October 12 revealed traces of primary system corrosion pmducts. Figure 1 Indicates the cavity monitor reaching 14,000 cpm and 15,000 cpm on October 13 and 14, prior to Power Test 242. The Power Test 242, Loss of Load from 44% Power, was run on October 14 and sub-sequent to the test, the reactor power was increased to 50 percent power. During the short period of time that the reactor was at 50% power, the cavity nonitor showed an increaso from 7,000 cpm to 25,000 cpm and subsequent anal-ysis of this filter showed corrosion products in increased concentrations f rom provious analyses on 10/12, 10/13 and 10/14 Fmm this data, it can be Inferred that the feedwater nozzle defect first ponotrated the wall folbwing the high pressure trip on October 10, 1969 Subsequent reactor operation between October 11 and October 14 increased the size of the penetration. The Power Test 242 (Loss of Load from 44% Reactor Power) significantly increased the size of the penetration. At the timo the reactor was shut down on October 14, the estimated leak rate thmugh the penetration at operating temperature and pressure was less than 2 ml/hr.(6) This example is used to illustrate the sensitivity of the leak CPC-851-21 J,' i j -l i

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DPC-851-21 detection system and how of fective it is in indicating the integrity of the primary system piping. Since this monitor reads out in the control room, the operator has a continuous Indication of the primary system i n tegri ty. Data Obtained from System During Operation in 1970 The 1970 lower cavity constant air monitor air particulate activity versus timo is shown in Figure 2 The only signi ficant change is between 60% power and 100% power operation when the activation product concentration shows an lacrease due to increased noutron flux in the lower reactor cavity. It also can be noted that the reactor cavity is " drying out" during this period as Indicated by the downward trend of the air concentration at a give:. power lovol. Figures 3 and 4 Indicate laboratory analysis of the fixed filter removed f rom the constant air monitor. Figure 3 indicatos the 30-minute decay count air particulate activity. A slight increase is noted af ter the reactor was escalated to 100% power in early September. During the leak in the feed-water nozzio in October,1969, this filter activity count increased a factor of 10 on the air sample removed on 10/14/69 A similar slight increase due to power escalation is noted in the 24-hour decayed filter on Figure 4 These analyses provide a constant check of the cavity continuous air monitor. Gamma-ray spectromotry of the air sample filter serves to identify any now contaminant in the air sample which could indicato a leak. The distribution of isotopos would Indicate the source of the leak, i.e., reactor water, reactor steam or feodwater. During 1970, the leak detoction program was expanded to obtain grab samples of water condensed f rom the air drawn f nom the lower cavity for tritium

analysis, in order to obtain an adequate volume of water for this analysis, a cold trap is utilized. Those samples were taken in addition to utilization of the continuous fixed filtor monitor to yloid better sensitivity for detection of very small steam or feedwater leaks which could originato in the lower cavity.

The moisture f rom the lower cavity is analyzed for tritium concentration along with primary water, condensed steam and foodwater. Analysis of the primary water, condensed steam and feodwater for tritium concentration has indicated nearly the same concentration at a speci fic time. A plot of the primary water tritium concontration and lower cavity moisture tritium as a function of time is shown on Figure 5 Using an activity balance relationship (II, the size of a leak source can be calculated. Using the tritium data obtained from the lower reactor cavity and the primary water, the calculated size of the source is shown on a plot of rato as a function of time during operation in 1970 in Figuro 6 Examin-ation of this plot Indicatos a decreasing or constant leak rato as a function o f ti me. DPC-851-21

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DPC-851-21 Discussion The calculated leak rate is smaller than the reported sensitivity of the system for any source, hence, it is very di f ficult to reach a firm con-clusion as to the exact source. In any case, since the calculated leak rate is constant, the integrity of the steam generation system is such that there is no undue risk f rom continued reactor operation. Several postulations can be reached from this data. Since the distribution of primary water Isotopes is not seen in the analysis of the constant air monitor spectrum of isotopes, it is concluded that primary water is not the source of the tritium detected in the lower cavity. This leaves two other possible sources; steam or feedwater. Neither can be positively identi fied f rom the analyses of the spectrum of Isotopes present. If the calculated size of the source were 10 to 30 times larger, positive identification could be made if an actual leak existed. Another postulation can be made as to the source of the tritium. This is the di f fusion of tritium through the primary system piping and reactor vessel into the lower reactor cavity. The di f f usion of tritium through fuel cladding has been reported in the literature (8) To determine if di f fusion is a possible mechanism, two independent calculations were made(9,10), These calculations used Fick's Law and a modi fication of the permeability equa-tion to determine the possible amount of tritium which could permeate the piping and rn;ctor vessel into the lower reactor cavity. These calculations indicated that it would be possible for up to 60 times as much tritium as detected to permeate the piping and vessel walls. It would not be feasible 'to calculate the minimum amount possible due to the unknown effect of the condition of the internal surf aces of the coolant system. Assuming that the diffusion model is the mechanism by which the tritium is entering the lower cavity, the increase in tritium concentration in the lywe{ reactor cavity between 60% and 100% power should follow the relationship. (Conc. of tritium in Vessel) 1/2 Conc. of tritium in cavity at 100% power / at 100% power [ Conc. of tritium in Vessel) 1/2 Conc. of tritium in cavity \\ at 60% power / at 60% power i f Figure 5 is examined, it can be shown that this relationship is indicated by the data. Conclusions The LACBWR leak detection system is capable of detecting very small primary system leaks as was shown in October of 1969 Operation since repair of the sensitized feedwater nozzles has indicated no unexplained increase in the continuous air particulate lower reactor cavity monitor. DPC-851-21 4 Sampling for tritium during 1970 has indicated its presence and calculations indicate no increase in the size of the tritium source during 1970 operat;ons. The di f f usIEn of tritium through ths piping in the lb~weh reactor cavity is ~ ' ~ a probable mechanism and it can be shown that di f f usion characteristics are indicated by the data, The presence of tritium in the lower cavity presents no undue risk to the a health and safety of the public or LACBWR personnel f rom continued power ope ration. Continued surveillance will provide assurance that no small leaks in the steam generation system go undetected. Acknow ledgements The author wishes to acknowledge the assistance of R. Craig, J. Shannon, D. Larson and L. Krajewski, who collected and performed analysis of the samples referred to in the report. t 1 i l 4 l _11_

~ \\ DPC-851-21 REFERENCES (1) Amendment No. 3 to Application for Transfer of Provisional Operating Authorization DPRA-6 for LACOWR, April 16, 1968, Dal ryland Power Cooperative, April 16, 1968. (2) Addendum to Amendment No. 3, Dal ryland Power Cooperative, March 17, 1970 (3) LACBWR Monthly Reports, ACNP-69504, ACNP-69505, ACNP-69507, ACNP-69508, ACNP-69510, ACNP-69511, DPC-851-11, DPC-851-12, DPC-851-13, DPC-851-14, DPC-851-16. (4) LAC 8WR incident Report 69-1, Dal ryland Power Cooperative, November 7,1969 (5) LAC 8WR Monthly Report for October, 1969, ACNP-69511, Allis-Chalmers Manuf acturing Company, Atomic Energy Division. (6) LAC 8WR Monthly Report for November, 1969, Section 4.3, DPC-851-1, Dalryland Power Cooperative. (7) LACBWR Primary System Sensitized Nozzle Safe-End Replacement Program - Summary Report SS-561, United Nuclear Corporation, April 10, 1970 (8) Jacobs, D. G., Sourcos of Tritium and its Behavior Upon Rolease to the Envi ronment, TID-24635, December,1968. (9) Rothman, S. J. and Westlake, D. G., Calculation of the Maximum Amount of Tritium that can Dif fuse Through the Wall of Piping Containing Primary Water into the Cavity of the LACBWR, Unpubilshed Memo, Argonne National Laboratory, December 14, 1970. (10) Epstein, Leo J., Tri ti um Lovels in the La Crosse Reactor, Unpublished Memo, Argonne National Laboratory, December 17, 1970 (11) Murdock, J. F., Private Communication, i l l ! __}}