ML20024H615
| ML20024H615 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/30/1991 |
| From: | Bond G, Dougher J, Nicholas Trikouros GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20024H614 | List: |
| References | |
| RTR-NUREG-0800 TR-081, TR-081-R00, NUDOCS 9106070037 | |
| Download: ML20024H615 (47) | |
Text
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O 8 OYSTER CREEK NUCLEAR GENERATING STATION i
L TOPICAL REPORT 081 May 30,1991 PREPARED BY:
f J. D. DOUGHER R. V. FURIA L C. PO G. R. TAYLOR 7 '
APPROVED: /' ' / *^
N. G. TRIKOUROS G.R. BOND l
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TR 001 Rev 0 Page 2
_ TABLE OF CONTEtflS SECTION TOP!C EAQg EXECUTIVE
SUMMARY
4 UST OF FIGURES 6 UST OF TABLES 7
1.0 INTRODUCTION
8 2.0 OBJECTIVES 9 3.0 GENERAL DISCUSSION OF POST ACCIDENT RADIOLYSIS 10 3.1 Temperature and Turbulence Effects for Non-Bolling Water 10 3.2 Bolling 12 3.3 Impurities 12 4.0 OYSTER CREEK SPECIFIC IODINE RELEASE AND METAL WATER REACTION 16 4.1 lodine Concentration from LOCA with and Degraded Conditions 16 4.2 Relationship between Metal Water Reaction and lodine Release 18 50 PLANT SPECIFIC OXYGEN CONCENTRATION WITH REVISED G VALUES 23 5.1 Methodology 23 5.2 Results 24 52.1 Oyster Creek Specific Oxygen Generation 24 5.2.2 Oxygon Concentration Following Severe Accidents 24 5 2.2.1 Total Core 24 5 2.2.2 Localized Effects 25 5 2.2.3 lodine Release Without MWR 26 5.2.3 Additional Conservatisms 27 LCP. GEN
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, TR481 Rev.O Pago 3 TABLE OF CONTENTS (cont'd)
SECTlQN LOP 10 fAGE 60 CONCLUSIONS 30
7.0 REFERENCES
37 APPENDIX A: OXYGEN VS TIME CALCULATION METHODOLOGY 39 APPENDIX B: NRC STAFF SAFETY EVALUATION ON NEDO 22155 43 i
TOTAL PAGES 47 f
t LCP. GEN
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TR-081 Rev. O Page ,i EXEClmVE
SUMMARY
P in Enclosure 2 of their 'Clartfication of NRC Staff Position on Hydrogen Mitigation Requirements 10CFR50.44 Oyster Creek Nuclear Generating Station *, dated November 6,1990, the NRC staff questioned the radiolytic oxygen generation rates used in NEDO 22155. The Staff stated that the resu!!s in NEDO 22155 were applicable to pure water or water containing only minimal amounts of impurities and that including the effect of iodine could drastically change the results. The Staff also indicated that post accident hydrogen and iodine concentrations may vary during an accident and are specific for each individual plant.
In order to respond to the NRC Staff's concerns, GPUN has prepared Topical Repor1081, ' Oyster Creek Plant Specific Oxygen Generation Following a LOCA' This report calculates the oxygen concentration in the OC containment as a function of time following a LOCA and conservatively accounts for the hydrogen and lodine concentrations in the containment water. The methodology described in Appendix A of NUREG-0800 (USNRC SRP Section 6 2.5), ' Combustible Gas Control in Containment *, is utilized except that the non boiling oxygen generation rate is calculated as a function of dissolved lodine and hydrogen.
An Oyster Creek plant specific lodine concentration was calculated for both the base case LOCA and for a more severe LOCA event in which core cooling is degraded such that a metal water reaction 5 times that of the base case LOCA occurs. The latter case results in iodine releases that are 300 times more than the base case LOCA. A plant specific fuel heatup calculation, with and without degraded ECCS performance, was periormed to determine fuel rod temperatures, metal water reaction rates and the number of failed fuel rods. The iodine releases were determined by comparing the calculated fuel centeriine temperature for the failed fuel rods against NUREG/CR 2367, " Updated Best Estimate LOCA Radiation Signature'.
The results of the evaluation show that for the lodine and hydrogen concentrations that would be expected as a result of a relatively severe event, such as a LOCA with degraded ECCS perfortnance, the cxygen concentration inside containment would remain below the 5% oxygen flammability limit. For very severe events in which 30% of the LCP. GEN l
i i
THet I Rev. O Page s core lodine is released and 40% of the core undergoes metal water retction, the flammable limit is not roached for about a year. Postulated events in which significant amounta of lodine are produced without substantial metal water reaction ars not credible.
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i LCP. GEN -
TR-081 Rev.O Page 6 UST OF FIGURES FIGURE .TjTLE PAGJ 31 ORNL EXPERIMENT NO.11 14 32 ' ORNL EXPERIMENT NO.10 15 41 OYSTER CREEK POWER DISTRIBUTION HISTOGRAM 19 42 FAILED PIN MWR AND IODINE RELEASE AS A FUNCTION OF TEMPERATURE 22 51 OXYGEN CONCENTRATION VS. TIME FOR BASE CASE LOCA 28 5-2 OXYGEN CONCENTRA*.lON VS. TIME FOR DEGRADED LOCA 29 53 OXYGEN CONCENTRATION VS. TIME FOR LODINE = 1.4% : MWR = 2.24% 30 54 OXYGEN CONCENTRATION VS. TIME FOR IODINE = 10% : MWR = 15% 31 55 OXYGEN CONCENTRATION VS. TIME FOR IODINE = 20% : MWR = 30% 32 54 OXYGEN CONCENTRATION VS. TIME FOR IODINE = 30% : MWR = 40% 33 5-7 OXYGEN CONCENTRATION VS. TIME FOR IODINE = 4.26% : MWR = 5.36%34 LCP. GEN
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4 . e TR-081 Rev.0 Page 7 UST OF TABLES
~ TABLE 1[ILE PAQg 41 OYSTER CREEK IODINE RELEASE DURING BASE CASE LOCA 20 42 OYSTER CREEK IODINE RELEASE DURING LOCA WITH DEGRADED 21 CORE COOLING 51
SUMMARY
OF RESULTS 35 l
l LCP. GEN i
. TR481 Rev. O Page a 1.0 INTRODilGT1QN On November 6,1990, the NRC issued a letter to GPUN entitled, ' Clarification of NRC Staff Position on Hydrogen Mitigation Requirements - 10CFR50.44 Oyster Creek Nuclear Generating Station *, (Ref.1 1). The letter had two enclosures: Enclosure I stated the St:Ts position on BWR Mark I compliance with the reguiations in general; and Enclosure 2 was a Safety Evaluation on the BWR Owner's Group methodology for determining the oxygen generation rates by radiolytic decomposition (NEDO 22155, Ref.1-2). The Safety Evaluation disagreed with the NEDO report on the radiolytic gas generation rate for bolling and non-boiling conditions The data which the NRC Staff used was based on an experiment conducted by ORNL (Ref. 31) for pue water and a theoretical model developed by BNL for water contaminated with iodine (Ref. 3-2).
Both have shown gas production rates higher than the NEDO assumed values. In this report GPUN will use the NRC recommended model, with consideration of beyond design basis post-accident conditions for both boiling and non-boiling reactor coolant water to ca!culate the Oyster Creek plant specific oxygen concentration. In particular, the iodine release fraction for conditions complying with the 10CFR50.44 requirements for degraded ECCS performance and its impact on the oxygen production rate will be addressed.
LCP. GEN
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TR481 Rev, O Page 9
2.0 ' OBJECTIVES The primary objectives of this report are as folkms: ,
a) To utilize the basic methodology provided by the NRC in NUREG 0000 (SRP Section 6.2.5),
Appendix A (Ref. 51) for calculating combustible gas concentrations in containment, with modifications to account for the effect of dissolved lodine and hydrogen on the radiolytic generation rate (G value),
b) To develop a value for the concentration of lodine in the containment water following a large break LOCA with a degradation of the ECC syst?rn such that the resulting metal water reaction (and resulting hydrogen release) is 5 times that resulting from a base case LOCA (without ECCS degradation).
c) To determine the Oyster Creek plant specific containment oxygen concentration as a function of time for the degraded ECC system perfG.:.ance condition evaluated and for more severe accidents as woll.
d) To show that inerting is effective in preventing a flammability condition in the containment following a relatively severe accident in which lodine is released from the core and significant metal water reaction occurs.
+
LCP. GEN
TR481 Rsv.O Page to l
1 3.0 GENERAL DISCUSSION OF POST ACCIDENT RADIOLYSIS For post accident radiolytic decomposition of water, Regulatory Guide 1.7 recommends that G(0j )=0.25 be used for both boiling and non-bolling .:onditions, it is known that tnis value is overly conservative (Enclosure 2 to Reference 1 1) and < hat many factors will affect the G value. For instance, temperature has an effect on the rate of decomposition. When water is nn-bolling, higher temperature usually reduces the radiolytic gas concentration. However, as the water approaches boiling, higher turbulence increases the gas production rate, and the G value increases. At the point of boiling, most dissolved gases in the water are stripped out and a higher G-value is appropriate. The presence of impurities, such as dissolved fission products, which come from post-accident fuel failure, have a strong effect with respect to increasing the magnitude of the G value.
3.1 Temoerature and Turbulence Effects on Non-Boilino Water The ORNL (Ref. 31) data shows the hydrogen partial pressure against integrated dose rate for a number of experimental cases. The experiments simulated various representative BWR core flow rates (100 gpm,1000 gpm and 10,000 gpm), and used different cover gas compositions (air or 5%Oj/95%Nj ) and temperatures (6S' C,95' C and 12S' C). The water was distilled so no impurity consideration are involved.
The test data generally concluded that:
- a. From 6S' C to 95" C, the G-value decreases with temperature. It turns around when temperate s is increased to 125' C (still non-boiling under pressurized condition),
- b. Inhial G(H2 ) varies from 0.1 to 0.3 for BWR representative core flow rates from 100 gpm to 10,000 gpm. A higher pumping rate corresponds to more turbulence and thus less recombination.
LCP. GEN l , _ , ,_ , _. . _ _ . - _ -
TR481 Rev, O Page 11 c, Radiolytic gas pressure reaches an equlllbrium (G =0) in each case. At equilibrium, the dissolved hydrogen will recombine with any oxygen molecules produced by radiolysis. The not production is zero.-
The Staff's Safety Evaluation (Enc. 2 to Ref.1) states that for pure water (no lodine), it was determined experimentally that with no dissolved hydrogen and no boiling G(02 )=0.08. This
- conclusion appears to be based on ORNL Case No.11, which involved 95% N, and 5% 0, gas over distilled water at 6S' C and a flow rate corresponding to 10.000 gpm in a BWR (Fig 3.1). G(02 )
becomes zero when the hydrogen's concentration reaches 2.5 cc/kg corresponding to an equilibrium partial pressure of 0.16 atm (Ref. 3 3). This was used as an argument that the G-value should be significantly greater than zero for pure water under non boiling conditions. However, the high equilibrium pressure is mainly caused by the high pumping rate during the experiment (15 cnf / min) corresponding to 10,000 gpm in a BWR under forced flow conditions. Higher turbulener removes free radicals faster and thus reduces recombination. For post accident BWR conditions, all pumps are tripped and a natural circulation condition is in effect in the core. The flow rate under these circumstances is closer to Case 10 of Reference 3.1 (see Fig. 3 2); from which we can derive G(03 ) becoming zero at a hydrogen partial pressure of about 0,04 atm (4% hydrogen in containment) or a concentration in water of about 0.6 cc/kg (Ref. 3-3).
Assuming a degraded core condition with 5 times the 10CFR50.46 calculated metal water reaction (2.24% MWR), the initial hydrogen concentration in the Oyster Creek containment is calculated to be about 4% (Rei,3-3). This partial pressure of hydrogen under non boiling conditions was shown above to result in a G(Q)=0.0 at equilibrium. The ORNL data case 10 thus supports the NEDO assumption of G(02 )=0 for non boiling if no iodine was assumed in the post accident water, i
LCP. GEN
t TR481 Rev. O Page 12 Since the presence of iodine in post-accident reactor coolant cannot be ignorod, the pure water G-value data will not be used in the Oyster Crook plant specific calculations presented in Soction 5 of this report 3.2 Boiling Bolling strips dissolved gases out of the liquid phase so that the maximum decomposition will proceed. Equilibrium betwoon the atmosphere and the liquid will not occur during boiling. It is conservatively assumod that post accident boiling will last 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in the Oyster Crook combustible gas concentration c'ilculation (consistent with NEDO-22155). A G(Os )=0 225 will also be conservatively assumed for this entire duration. Enclosure 2 to Reference 1 states that the maximum values of G(De ) for 5% MWR and 30% lodino release are between 019 and 0.22.
3.3 Imourities The presence of Impurities such as post-accident fission product lodino in the reactor water may affect the decomposition rate. A static model is usod by the NRC (Ref,3 2) as follows:
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VR481 R31. O Page 13 For a small lodine concentration in the water (<10' gm-mole /l), G(Hg ) decreases very quickly to zero as the hydrogen concentration in the water builds up. However, for a moderate lodine concentration ( [1] > 10' gm-mole /l ), which corresponds to greater than 2% of the total core lodine being released to the water, the G value remains positive and would increase the long term oxygen build up in the containment.
LCP. GEN
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TR481 Rev.O Page 3e 4.0 OYSTER CREEK SPECIFIC IODINE RELEASE AND METAL WATER REACTION 4.1 lodine Concentration from LOCA and LOCA with Dearaded Conditions Based on the NUREG/CR 2367 fission product release data (Ref. 41) an Oyster Creek specific lodine re;oase concentration was calculated for both base case LOCA and LOCA with degraded core cooling. The degradM core cooling case is defined as having a metal water reaction (and resulting hydrogen generation) that is 5 times the amount calculated for the base case LOCA used in this evaluation. The metal water reaction rate used in the base case LOCA was somewhat greater than ,
those calculated pursuant to 50AS(b)(3) becausi of the ECCS code used in the evaluation. The released lodine concentrations are calculated in this section to be 1.80E49 and 5.44E47 gm mol/l for the base case LOCA and the LOCA with degraded conditions, respectively.
The Oyster Creek core wide metal water reaction based on compliance with 10CFR50,46 is 0.448%
The 10CFR50.44 criterlon for degraded core conditions is the larger of: 1) five times the amount of hydrogen calculated in demonstrating compilance with 10CFR50.46, or 2) for the amount that would result from reaction of all the metal in the outside surfaces of the cladding cylinders surrounding the fuel to a depth of 0.00023 inch. A fivefold increase in hydrogen corresponds to a fivefold increase in the metal water reaction which would be 2.24% The metal water reaction due to the reaction of 0.00023 inches of the cladding surfaces is 0.77% Therefore, the five time increase in MWR is the criterion used for Oyster Creek in determining the degraded core condition.
A fuel heat up calculation was performed to determine fuel rod temperatures. MWR and ti number of failed fuel rods during a LOCA based on an end of cycle (EOC) core conditions. Using the l NUREG/CR 2367 lodine release rate, the total lodine concentration was calculated along with a core wide metal water reaction.' This served as a basis from which the degraded core cooling case could i
be evaluated. For the degraded core cooling case, it was assumed that the initiation of emergency core cooling was delayed and flow rates were less than Appendix K requirements. The fuel heat up 1 LCP. GEN
TR 081 R:v.O Page 17 calculations were redone with reduced ECCS flow and Itcrating on the time for delayed core cooling until the metal water reaction increased by a factor of 5. An lodine concentration was calculated for the degraded core cooling case using the resulting fuel rod temperatures and Reference 41 fission product release data.
The heat up calculations were performed using the HUXY code (Ref. 4 2). The HUXY code has been approved to perform 10CFR50 Appendtx K calculations for the ANF fuel loaded in Oyster Creek. The HUXY code does not calculate the mechanical response of the cladding during a LOCA.
However, it does allow a temperature input which, when exceeded, falls the fuel rod and calculates a MWR based on both an inner and outer cladding surface as per Appendix K. Current licensing analyses (Ref. 4 3), show that a fuel rod will perforate at nodal exposures exceeding 19.0 GWD/MT when the peak clad temperature (PCT) exceeds 1600 F. Dividing by an approximate axial exposure peaking factor of 1.25, this translates to a bundle average exposure of 15.2 GWD/MT,' For bundles having exposures less than 15.2 GWD/MT, a fuel failure temperature of 2500 F was used.
A core power distribution histogram (Figure 4.1) of number of bundles versus radial power was developed is, bundle exposures below and above 15.2 GWD/MT. An EOC case was used for conservatism to maximize the number of higher exposeo fuel Dunoles. Another conservatism was to place all fuel bundles in the peak radial power group, for the high and low exposures, at their MAPLHGR limit. The heat up calculations were repeated for the low and high exposures, for each of the radial power factors indicated, and for the base case LOCA and LOCA with degraded core cooling conditions.
The results of the calculations are summarized in Table 4.1. The base case LOCA calculations result in 6288 failed fuel rods out of the 33,600 rods in the core, and a core wide MWR of 1.16%. Both of these values are greater than the Appendix K results due to the additional conservatisms used in this LCP. GEN
T M B1 he Rev.O Page 1g analysis. The degraded cooling case results in 17744 fallod fuoi rods and a core wido MWR of 5.85% (an increase of a factor of 5 04 over the base case LOCA case). The iodine release rate was calculated for each group of bundles for a gNen radial power factor based on the calculated fuel temperature. A fuel rod lodine concentration of 0.486 gms por fuoi rod, which corresponds to a high power rod, was conservatNoly used for all failed fuel rods in the coro. The average core lodine concentration is 0.389 g/ fuel rod (Ref. 4 4). In addition, the loMne release rato for a failed fuel rod was conservatively based on the limiting (hottost) axial nove in the rod.
4.2 f,g.lationshio Between Metal Water Reaction and lodine Relgasp The analysis discussed in Section 4.1 providos an estimate c4 tne coro wido MWR and lodino release for degraded core conditions. The treatment of the MWR was based on the parabolic rate law of Baker and Just and the lodine release was determined using NUREG/CR 2367, Figure 4-2 is a plot of the MWR and lodine release as a function of pin contoriino temperature for a failed fuel pin. As can be seen, if the pin conteriine temperaturo increases, both porcent MWR and lodino release also increase. As the temperature increases to 1600' C, the lodine rolesse approaches 30% while the MWR approaches 70% of total.
1007 C represents a limiting fuel pin condition for the degraded core analysis reported in Section 4.1 While a few pins may approach this limiting condition, the majority of the fuel will remain well below this temperature. The inset in the upper left corner of Figure 4-2 lists the degraded core analysis results for percent of total core MWR and percent of total core lodino release. The lodine release and MWR percentages reflect the fact that for the degraded core condition, only half the pins fail and the centerilne temperatures of most of the pins are well below 1607 C.
LCP. GEN s -
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l L Oyster Creek Iodine Release During LOCA with Normal Core Coolinq j !
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RADIAL FAILED BUNDLES TOTAL C/L . FUEL 99tR IODINE IODINE 'f EIPOSURE PONER RODS / FAILED TEMP C RELEASE' CONCENTRATION' I GWD/MT BUNDLE RODS j RATE ON MOL/L l t
<15.2 1.48 0 76 0 1200 3.5 0.00035 0.OOE+00 I L
<15.2 1.3 0 48 0 1080 1.8 0.00024 0.OOE+00 I t
<15.2 1.2 0 32 0 1010 1.1 0.00017 0.OOE+00 I
<15.2 1.1 0 20 0 960 0.7 0.00013 0.OOE+00 . '
>15.2 1.3 52 12 624 1160 6.1 0.00031 2.79E-10 'I
>15.2 1.2 32 112 3584 1060 3.2 0.00022 1.14E-09 3 i
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>15.2 1.1 30 40 1200 990 2.0 0.00015 2.60E-10 i
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>15.2 1 11 80 880 930 0.9 0.0001 1.27E-10 [
l >15.2 0.9 0 32 0 860 0.3 0.00006 0.OOE+00'
, >15.2 0.8 0 108 0 815 0.2 y 30004 0.OOE+00
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i TOTALS: 560 6238 1.80E-09 i i
TOTAL CORE % IODINE RELEASED = 0.0046 )
TOTAL CORE % MWR = 1.16 i.
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FIGURE 4-2 Page 22 FAILED PIN MWR AND LODINE RELEASE AS A FUNCTION OF TEMPERATURE PERCENT OF TOTAL 80 - -
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1 DEGRADED CORE RESULTS l o
Percent of Total Core /
MWR lodine Release /'
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O 500 1000 1500 2000 PIN CENTERLINE TEMPERATURE (deg C)
--- % M W R + % lodine Release
- TR 081
- Rev.0 Pag) 23 5.0 PLANT SPECIFIC OXYGEN CONCENTRATION WITH REVISED G VALUES 5.1 Methodoloav The Oyster Creek plant specific oxygen concentration was calculated for a variety of lodine and MWR assumptions. The metriodology described in Appendix A of the NRC Standard Review Plan (Section 6.2.5), NUREG4800, ' Combustible Gas Contrd in Containment' (Ref. 51), was used except that G values are calculated as a function of dissolved lodine and hydrogen (Ref. 3-2) (see Section 3.3):
G(H2 )=0.45 2.7/(1 + kl [l]/kh [H] '5.1) where G(Hg ) = net hydrogen generation rate, molecules /100 ev
[1]
iodine molar concentration in the coolant (H) = hydrogen molar concentration in the coolant kl,kh
reaction rate constants for the adverse lodine reaction and the favorable hydrogen reaction in radiolysis suppression For the first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of the LOCA, G(H,) is given its maximum value (0 45) n in boiling.
Thereafter, G(it) is calculated from Equation 5.1. The dissolved hydrogen is calculated from Henry's Law; PH2= KH*tH1 (5.2) where KH = the Henry Law constant for hydrogen in water PH2 = the pressure of hydrogen in the gas phase LCP. GEN
18421
' Rev.0 Page 24 The details of the application of these equations to the specifics of the OCNGS are gNon in' Appendix A.
5.2 Besults 5.2.1 Dvster Creek S22fd!!C_OEY20!LCODCtntrg11gn The results of the HUXY analysos discussed in Soction 4.0 showed that for a base case LOCA, the lodine concentration will be 1.80E.9 g-moles / liter which corresponds to a total core lodino release of about 0.0046%. The total core % MWR for that case was 1.16%. The containtnent oxygen concentration for this event would only increase by about 0 25% as can be soon in Figure 51.
For the degraded ECCS case analyzed, the total core iodine release was 1.4% and the calculated % MWR was 5.85% (a factor of 5.04 increase over the base caso LOCA case).
The oxygen concentration in containment for this case would not increase from the initial value. This is depicted in Figure 5 2. Figure 5 3 provides the results for the same lodino release case (1.4%) with a MWR of 2.24% (5 times the 0.448% calculated for 10CFR50.46).
- Again, a 5% containment oxygen concentration is not reached.
5.2.2 Oxvoon Concentration Followino Severe Accidents 5.2.2.1 Total Coro in this section, lodine and MWR assumptions more severe than those calculated l
specifically for Oyster Creek in Section 4,0 are evaluated with respect to expected oxygen concentrations. These analysos are being performed to address the roloase of larger amounts of lodine up to and including 30% of the total core lodino inventory. The release of such large fractions of the total core lodino inventory would require that all of the core L al rods achlove substantial fuel centerlino l
temperatures (Ref. 41). Fuel rods achieving such high contoriino temperatures LCP. GEN
_ . _ . . .. - . _ _ ~_ _ __ _m . . . - _ - -
' TR481 l . Rev.0 ;
Pags 25 I
would also be undergoing substantial metal water reaction. The relationship between these parameters was discusseo in Section 4.2 of this report and will form the basis of the cases analyzed herein.
Figures 5.4 through 5.6 show the oxygen concentration profiles for 10%. 20% and 30% total core lodine release. The 10% lodine analysis (Fig. 5-4) shows oxygen concentration for a % MWR of 15%. The results indicate that it will take about a year to reach 5% oxygen concentration. Figure 5.5 shows the 20% lodine release results with a % MWR of 30%. Again, oxygen concentrations of 5% do not result in less than approximately one year. Similar results can be seen in Figure 5.6 for the 30% lodine release and a % MWR of 40%
The selection of the % iodine /% MWR ratios was based upon the results depicted in Figure 4.2 which shows that for a given fuel temperature condition that results in a
% lodine release, the % MWR for that condition will be su'ostantially higher than the corresponding % lodine. A conservative ratio of 1.5 or less was used for each case analyzed.
5.2.2.2 Localized Effects This section is addressing the concem that, in the event of a LOCA, a small fraction of the core might become overheated. It is assumed that this might occur from a hot spot resulting from local coolant flow starvation as a result of: 1) delivery of less l
than planned cooling to a localized area, or 2) local flow blockage it is further l
r assumed that the expected MWR will not occur at any time even though such an assumption is not credible, i The following conservative assumptions are being used:
LCP. GEN
_ _ _ _ _ _ - - - .- . . . . . . - . _ . - .~ . _ _ _ - .
. . TR481
. Rev. O Peg 3 26 a) Fuel centerline temperatures in affected region reach 1600* C (30% iodine release).
b) Size of affected region is 10% of core (56 bundles).
c) % MWR for affected region is 1%*.
d) Remainder of core as per degraded LOCA cas,e of Section 4.1.
This is conservative since higher percent MWR would produce less oxygen than the assumed case because of the suppressing effects of hydrogen on G(0,), The porcent MWR would be about 70% for the affected region.
These assumptions result in an overall total core lodine Inventory release of:
(1.4%) * (0.9) + (30%) * (0.1) = 4.26%
The total core % MWR is:
(5.85%) * (0.9) + (1.0%) * (0.1) = 5.36%
The oxygen profile resulting from this condition is shown in Figure 5.7. The results show that a 5% oxygen concentration in containment for this non-credible assumption will not be reached for approximately three months.
5.2.2.3 lodine Release Without MWR lodine release without a comparable MWR is not credible. Even if a blockage of cooling water to a small region of the core is assumed as the basis for limiting the MWR, eventually either cooling will occur or fuel melt will result. Melt progression l will only cease when cooling is re-established. When this occurs, MWR will also occur. The requirement of 50.44(h)(1) is that the degree of degradation is not sufficient to cause core meltdown. This implies that cooling is established, and this cooling of hot fuel must result in significant MWR.
! LCP. GEN
. .- TR 081 '
, Rev. O Pag 3 27 At TMI 2 where cooling was uruvallable such that significant core heatup ocetared, a significant MWR resulted from the eventual re establishment of cooling water.
Even if complete core melt were to occur, significant MWR would occur when the melt material comes in contact with water inside containment.
52.3 Additional Consefyall$m$
a) The NRC model used in this report (Section 5.1 and Appendix A) is very conservative (overpredicts G value) when it is applied to low impurity (low lodine and hydrogen concentrations) cases. The reason for this is the assumption of an initial G(He ) of 0.45 which is then allowed to decrease. In the Zittel experiment for pure water (Ref. 31), the G value never exceeds 0.3. This difference contributes to a larger radiolytic gas production rate and higher oxygen concentration in i
containment.- For lodine concentrations less than approximately 10' g-molos/ liter, li would be more appropriate to apply the Zittet results, b) The calculation herein assumes that the precipitated ZrO, from the metal-water reaction would occlude 10% of the water and that inis water would have a G(H2 )=0.45. The NRC in Reference 3 2 uses a value of 1% rather than 10% for this effect. The model herein would thus overpredict the radiolytic gas production slightly as a result of this.
LCP.CEN ;
i 4
1 s
1 FIGURE s-1 i BASE CASE LOCA 4.0 r- ,
i
_ l 5 I '
3.9- -d i
. r, -
l l 6 3.9, ii j
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w"
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FIGURE s-2 r
IODINE =1.4%; MWR=5.85% , '
i 3.6 i 3.6 .
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- z -
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33 ........ .........
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- T1ME (DAYS) 'E ?to6 i ca O - L C
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RGURE s IODINE ==1.47.; MWR=2.247.
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. . TRoal
- R:v O Pag) .3 TABLE 5-) l EUMPd&B. Y QE.BESULIS
.GASE DESCRIPTION % IODINE E '16'8 TIME TO SEQXYDEfL(Q&y.Sj Base case LOCA 0.0046 1.16 > 1000 LOCA with Degradod ECCS 1.4 5 B5 > 1000 ,
1 LOCA with Degraded ECCS 1.4 2.24 > 1000 Sovere Accident 10 15 300 I
Severe Accident 20 30 > 1000 Severe Accident 30 40 > 1000 Localized Effect 4.26 5.36 90 LCP. GEN
-. _ , , . . - . . , . . . _ , , . . ~ . . . . , _ _ _ _ _ _ . - . _ . . . _ . _ - . , . . _ , . - . - . , _ . - . _ ,
TR481 l l Rev.O Pags % 4 4
4 1
i 6.0 MWS10flS R The following conclusions can be reached as a result of the analysos discussed in the report:
a) The non-bolling G(Oh ) is not zero with the presence of discolved kdino, but the effect of the i dissolved lodino in the containment water on the G(0;) is offset by the offect of the dissolved hydrogen resulting from the initial metal water rocction and from radiolysis. ;
b) For both a LOCA and a LOCA with sovoroly degraded ECCS periormanco, the oxygen concentration in the Oyster Crook containment would not reach $%
c) For sescro accidents in which 30% of the total core lod;no is releasod, i o., NRC assumption of fuel rod conterline ternporatures of 1007 C over the entito cole, the oxygen concentration would not reach 5% in less than approximately one year for a conservatively low 21rconium water reaction rate of 40%. For 2irconium water reaction rates greator than 40%. It would take even longer to reach 5%.
d) Thero is no cndiblo mechanism by which substantial amounts of coro lodino can be released without a substantial amount of metal-water reaction.
e) Even in the ovent that flow is blocked to a small fraction of the core following a LOCA, oxygen >
concentration in containment would not reach 5% for several months.
s I
LCP. GEN l
, , 7Roal 4
Ret 0 Page n j 7=0 REFERENCES 11 Stevt.n A. Varga, OSIPC Letter tu R. L Long, GPUN, November 6, itKX) 12 DWR Owners Group Report, ' Generation ard Mirigation of Combustible Gas Misture in (norted Mark i Containments', GE NEDO 22155,1982.
31 ORNL TM 24t2 Park Vill, ' Design Conskforations of Reactot Containmerit Spray Gyr. toms . Part Vill Eolling Water Reactor Radidysis Studios'. H E. Zettel, October 1970.
32 Memo from K.1. Parczewski to Victor Donaroya, *Raddysis of Coolant Water in Millstono l',
June 23,1982.
33 GPUN Cali ..iation C1302 243 5450-000, 'OC Post LOCA Hydrogen ard lodine Concentration in the e Containmont and Gas Ptoduction Rate by Radiolysis', June 1991.
41 Updated Dost Estimate LOCA Radiation Signature,' NUREG CR 2307, D. D. Thayer, 42 HUXY: A Generatind Multipod Heatup Code with Appendix K Heatup Option XN CC 33(A)
July 28,1975.
41 ' Oyster Creek NGS SAFER /CORECOOL/GESTR LOCA Loss of Coolant Accident Analysis,' August 1987.
44 GPUN Calculation C1302 220-5411236, ' lodine Holoaso During LOCA with Dogradod Core Cooling".
May 1991, 51 NUREG 0000, USNRC Standard Review Plan, Section 6 2.5, ' Combustible Gas Control in Containment', Appondix A.
A1 NUREG4000, USNRC Standard Review Plan: Section 0 2.5 Combustible Gas Control.
A-2 US NRC Memorandum, K.l. Parcrewski to Victor Borwoya Chief, Chem. En0rg Branch, Div. of Energy, 'Radiolysis of Coolant Water in Millstone 1.
A-3 GPUN Calculation C1302 2404340 005, 'Radiolysis Attor LOCA at OCNGS', May 1991.
LCP. GEN
.- .-. . _ - . - _ _ . - . - - .~.
> e TR481 Rev O Pag) 3H APPENDIX A
.0X10DLYS llMLC&QV1ARQtLMERIQDOLOG.Y
.L!ilBQRLICILQD The CRP (Ref.1) Soction 6.2 5 calcutations are used with modifications as noted.
DISCUS $10B in the event of a loss of coolant accident (LOCA), hydrogen ard oxygen gases will be generated within the Oyster Creek reactor containment by:
- 1. Metal. water reaction invcdving the ritconium fuel cladding and the reactor coolant, producing free hydrogott
- 2. Radiolytic decomposition ct 'he post accident emergency cooling solutions. producing both oxygen ard hydrogen if a sufficient amount of hydrogen is generated, it may react with the O, present in the containment atmosphere or, In the case of inotted containments, with the oxygen generated fe!!owing a LOCA.
The extent of Zirconlum water reaction and associated hydrogen production deporKis strongly upon the course of events assumod for the accident. Analytically the reaction can be described by; Zr + 2H,0 - Zro, + 2B, I lb Zr - 0.043956 lb H, 1 lb Zr - 0 021970 lb-mole Hj Therefore, one pound of reacted zirconium will produce 0 021978 pound moles of free hydrogen. Assuming the perfect gas relationship, this is equhtalent to 8.4860 scf/lb Zr:
l l V - MRT V = 0,021978(10.71)(530) /14.7 (Standard conditions taken as 14.7 psia,537 R)(77 F)
V = 8.4866 rict/lb Zr.
1 l
l LCP. GEN
'
- VR401
- Rev.O Page 39 i
The total amount of hydrogen produced is based on the amount of reacted tirconium The computer program, to maintain a dogree of generality, allows the roaction percontage to be specifiod as an input quantity. The expression 1 .
2 used is:
J WG = (022)(WZr)(6)
! where WG = pound moles of hydrogon generated WZr = weight of rirconium fuel element clad 6,e zirconlum water reaction fraction
, The rate of gas producilon from radiolysis depends upon the power decay profilo and the amount of fission products released to the coolant. The radiolytic hydrogen production rate at time (t) is given by;
- Y " #' Ff" ' -._ _P ,OfHy,gggg),ggg))
Su( t) = (D) (N) -100 ( B) ( N) 100 r 8 where S,(t) = hydrogon production rate, Ib mde/ soc P - operating reactor power level, MWt B = conversion factor,454 gm. mole /lb molo N ?
= Avogadro's number,0 023 x 10 ) molecules /gm molo Q = radiolytic hydrogen yloid in core, molecules /100 ev
((t) - gamma ray fission product energy absorbed by coro coolant, ov/ soc MWt 4 - radiolytic hydrogen yield in solution, molecules /100 ev See below for definnion of G(H,)
((t) = energy absorbod in coolant outside Coro due to fission products cissolved in coolant, ov/ soc MWt The quantity ((t) is defined by:
((t) = (Q K(t)
LCP. GEN
. . . ~ -_ - . -
. . TR481
, . R*v.O
- Pag) ao
. t 4
l 1
where (f,( = fraction of fission product gamma enerCy absorbod by coolant in core region
= 0.1 ;
i ft(t) a gamnu energy production rate, ev/(Sec-MWl)
Similarly, E,(t)is defined by:
1
((t) = (t, , ,( F4 , , (t) + t H (t) i where (f,, ,), = fractior - atal solid fission product energy absorbed in coolant outside core !
0.01 i f4, ,(t) = total solid fission product energy production rate, ev/sec MWt
( = fraction of lodine isotopo energy absorbed in coolant outside core
= 100% of the fraction of lodine energy released to the coolant ,
H(t) = lodine isotope energy production rate, ev/sec MWt ,
The equations for oxygen generation by radiolysis are identical to those above describing hydrogen evolution except that the yield is one half that of hydrogen. For calculational purposes, the reactor decay profiles (F4(t), Ft , ,(t),
and H(t)) specified by the ANS 5.1 standerd for two year reactor operation have been fittod by several finite ,
exponential series expressions and also incorporated into the program. The resulting equations are:
N, ( t ) = 108 8 ( 5 .1912 e "" ' + 0. 0 7 4 3 e d8"* ' + 0 . 6 5 57 e * * ' 8" + 0 . 4 0 9 8 0 * *" ' ' + . 016 5 0 e ' ' ""*
- H,,, ( t ) = 2 . 0 H, ( C) ;
hr( C) =108 8 ( 0. 619 7 e ** ' n'"" ' + . 3 2 7 9 e *1 8 5'" ' ' + . 0 5 7l 4 e " * '
LCP. GEN r
. .--.+-wAw.. . --,m, ,-w -----,w. . , . .-.r v~w ..,m--,--wy.-.. .-- --- . . - - ,,.-,,.y.--y-.,-r---,--..+~# wr.,.,-yw. -.--.-w-,me
_ _ . . _ . _ _ _ . _ . _ _ _ _ _ _ _ _ _ . . _ _ _ _ _.~. _ ._. . . _ _ _ . . . -
, . . TR481 Pw. 0 Pags 41 where t = time a'ter reactor shutdown, sec, Between 400 ard 4 x 1(f sec, the equations overpredict the standard curve by 20% The equations urderptedict the standard curve soon after shutdown. However, thl3 does not serioutJy affect the results due to the short time perKd involved. The equations are equivalent to the afterheat decay curve in BTP ASB 9,2 over the times of interest for post-accident hydrogen generation it should also be noted that this formulation overpredicts the radiolytic j
hydrogen generation by a small amount due to a " double-counting
- of the gamma energy of those fission products 4
assumed to be released from the fuel rods.
G(H,-) is taken as
= 0 45 during bolling
= 0.45 2.7/(1 + ki'[l)/kh2'(H]) in non bolling water (A 1) where kl = 10101/(m's) (liter / mot see) kh2 = 3071/(m's)
(1) = dissolved lodine, mol/ liter
[H] = dissolved hydrogen, mol/ liter We assume (as per A. O. Allen, Ref. A.2)
- 1. Water system consists of suspended ZrO2 and dissolved lodine in water.
- 2. Water included in the porous ZrO2 particles continuet to boll.
- 3. The fraction of water (and decay energy) absorbed in the ZrO2 is fz, and this is the same in both the core and torus (fz-0,10).
l-l l
LCP. GEN 1
. . TR 081 R:v.0 Pegs n The dissolved hydrogen at any time is calculated from
[H) = Ph2/KH (A2) where Ph2 e partial pressure of hyd' ogen in the total gas volume, (wetwell + drywoll), psia KH = Henys Law constant, psla/(mol/l)
A basic computer program performs these calculations. The hydrogen inventory is calculated by step-wise i integration of the hydrogen production rate, the Ph2 is calculated assuming a perfect gas in the drywell, the [H]
calculated from Equation A.2 and the G value for the nex1 time increment is calculated from this (H) value and A.2 to repeat the cycle.
The percent oxygen at each step is calculated from '
% oxy = 100% mox/mox + mh+ mn) where mox = total moles oxygen (original inventory plus 0.5 times the mols hydrogen produced radiolytically) mh = total moles hydrogen (radiolytic plus Zr/ water reaction mols) mn = total moles nitrogen originally present The radiolytic hydrogen formed during boiling (the first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> LOCA) can be calculated analytically since the G is independent of time, and the decay energy expression integrates to a sum of terms in the form B'(1 EXP( C*t)) with b and c constant and t = 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
i 4
)
i LCP. GEN
~
l l-
. . TR-031
- Rw. o Pag) 43 i
APPENDIX B SAFETY EVALUATION BY THE OFFICE OF NUCLEAR RgAQJQR REGMLATION
.GENEBALILECTRIC COMPANY'S METHODOLOGY FOR DETERMINING RAJES OF GENERATION OF OXYGEN BY RADIOLYTIC DECQMPOSITION (NEDO 22155) l l
LCP. GEN
. _ . . _ , _ -. . . . . . _ - ~ , _ . _ . _ _ . - _ _ _ _ _
e
,, ,g
,,d y ',
k g UNITED STATES NUCLEAR REGULATORY COMMISSION '
, <8 W ASHINotoN, o C 70f,65
\, ' ' /
ENCLOSURE 2 SAFETY EVALUATION BY THE 0FFICE OF NUCLEAR REACTOR REGULATION GENERAL ELECTRIC COMPANY'S METHODOLOGY FOR DETERMINING RATES OF GENERATIONS OF OXYGEN BY RADIOLOYTIC OECOMPOSITION (NE00-221SS)
In June 1982 General Electric (GE) issued the subject report containing a description of the methodology for determining rates of generation of oxygen by radiolytic decomposition of water in the inerted Mark I containments. In this report, GE assumes that after an accident water in the containment will boil for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> only. During this time it will undergo radiolytic decomposition with oxygen generated at the rates corresponding to G(02)=0.1, Where G(0 2 ) is a number of molecules of oxygen generated by 100 ev of radiant energy absorbed.
This value was based on the results from the measurements of the hydrogen evolution rate in the offgas systems during normal (boiling) operation and during refueling shutdowns and confirmed by the experiments performed in the KRB Nuclear Power Plant.
For radiolysis of water beyond 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, when boiling ceases, G(0 2)=0 was assumed and consequently there was no net generator or radiolytic oxygen.
This last assumption was based on the analytical results obtained by Knolls Atomic Power Laboratory (Reference 1) and by Argonne National Laboratory (Reference 2) in connection with the Three Mile Island accident. The values of G(02 ) in the GE report differ considerably from the value of G(02 ) in Regulatory Guide 1.7 which for both boiling and non-boiling cases recommends G(04 )=0.25. However, this value is not based on any specific mechanism of radtolysis but is chosen to bound all possible cases and consequently it tends to overpredict the rates of generation of radiolytic oxygen. In 1982 an extensive effort was undertaken by the Northeast Utilities and by the NRC in connection with the Millstone 1 licensing action to determine a more realistic method for calculating rates of radiolytic oxygen generation. In performing this task the staff was assisted by a consultant from BNL. The results of this effort: have indicated that G(0 )2 is not a constant parameter but varies with the amount of hydrogen dissolved in water and with the concentrations of certain impurities, most notable: among them iodine. Since concentrations of these substances may vary with time and may be different for different accidents, the true value G(02 ) should be expressed as a function of these variables.
In general, an increase of concentration of hydrogen in water results in a decrease of radiolysis d" to promotion of recombination reactions. On the-other hand an increase o. iodine concentration tends to promote radiolysis by destroying free radicals which are required for the recombination reactions to proceed. The highest rate of oxygen generation is achieved when G(02 )=0.22, 1
- - - - _ - - ~ ~ - _
TR 061 Iw v , o j 2 Iwge 45
.hich is the highest theoretical limit for gamma radiation. This occurs when l water is completely free of oissolved hydrogen, or when the concentrations of dissolved iodine are extremely high. However, in most cases G(0 2
) will be lower and at certain concentrations of hycrogen and iodine the rates of radiolytic dissociation and recombinations reactions may become equal resulting in G(0 3)=0 and no net generation of radiolytic oxygen. During the boiling regir'e hydrogen will be stripped by vapor bubbles and it is expectea that G(02 ) will be higher ;
than in non-boiling water, ,
Quantitative evaluation performed by the stat' was based on the model developed by the BNL censultant (Reference 3) and on the experimental data from ORNL (Reference 4), For pure water (no iodine) it was determined experimentally that with no dissolved hydrogen and no boiling G(0 2)=0.08. However, when under non-boiling conditions the concentration of dissolved hydrogen reached 2.5 cc/kg of water, corresponding to equilibrium hydrogen pressure of 0.16 atm.,
G(02 ) became zero and generation of radiolytic oxygen stops. This finding contradicts the information in the GE report where G(0z)=0 was assumed for all non-boiling cases.
For water containing dissolved iodine no applicable experimental data were available and the staff calculated G(0 ) 2corresponding to the maximum creoible iodine concentration in water using the BNL model. Since all iodine in the containment water comes from failed fuel, an accident had to be postulated '
which would result in a release of this amount of iodine. In such an accident fuel was assumed to fail by oxidation of Zirconium cladding and hence, in addition to released iodine, additional hydrogen was produced. Concentrations of both these substances had to be considered in calculating G(0 ).2 The accident considered consisted of a LOCA in which 5 percent of fuel cladding was oxidized by reaction with steam producing failure of all fuel rods and overhetting of the core, but without initiation of fuel melting. This case represented maximum degradation of core allowed by 10 CFR 50.44(d)(1) and 10 CFR 50.46(b)(3). The analyses performed by Sandia (Reference 5), based on the experimental work on fuel rods from the H. B. Robinson plant, have indicated that for this type of accident 30 perc1nt of total fuel iodine inventory was released. The released iodine consisted of the initial gap inventory and of the iodine diffused from the overheated fuel. Assuming that all the released iodine was dissolved in water and using plant parameters corresponding to a typical BWR with Mark I containment, the iodine concentration in water was determined to be 1.11 E-5 moles / liter and the partial pressure of hydrogen in the ccntainment 0,12 ata. This partial pressure corresponds to an equilibrium concentration of 1.9 cc hydrogen /kg of water. Inserting this value of iodine concentration into the BNL mathematical model a relationship between G(03 ) and partial pressure of hydrogen in the' containment was developed. From this relationship it was determined that for a non-boiling case, when partial pressure of hydrogen was 0.12 atm., G(0 2 )=0.19. -It also found that G(03 )
would not reach zero value until partial pressure of hydrogen in the containment reaches l'atm, For boiling case, when hydrogen is stripped from the solution, G(0 )3 would be slightly higher, somewhere between 0.19 and 0.22. <
% a e 9
Tu-081 iw v , o 3
twp .m These values differed consicerably from those in the NE00-22155 report. The main difference was probably due to the GE results being applicable to pure water or to water containing only minimal amount of impurities. Including the '
effect of iodine, which would be released during certain types of LOCA, could drastically change the results.
CONCLUSIONS AND RECOMMENDATIONS
- 1. The NE00-22155 report underpredicts generation of radiolytic hydrogen for both boiling and non-boiling cases. This is due to the use of too low values for G(02 ). G(02 )=0.1 for boiling case was based on the measurements made in an environment of zero or low iodine concentrations. G(03 )=0 for non-boiling case was derived from the data calculated by the codes which did not consider effects of dissolved iodine. The results were also in disagreement with the experimental data from ORNL.
- 2. Since G(02 ) is a function of hydrogen and iodine concentrations in the containment water, it may vary during an accident and is specific for each individual plant.
- 3. The maximum values of G(0 2
), calculated with the NRC radiolysis model for LOCA (5% metal water reaction and 30% iodine release) in a BWR with Mark I containment, are G(02 )=0.19 for non-boiling and between 0.19 and 0.22 for boiling cases. They are considerably higher than the values prasented in the General Electric's NE00-22155 report.
- 4. The value of G(02 )=.25 in Regulatory Guide 1.7 is overly conservative.
However, it is not very much different from the maximum values calculated for a LOCA using the BNL model, it is recommended, therefore that until a better understanding of post accident radiolytic decomposition of water is developed, this value should be used for predicting generation rates of radiolytic oxygen in the containment.
REFERENCES
- 1. J. C. Conine, D. J. Krommenhoek and D. Emanual Logan, KAPL Evaluation of Radiolysis Associated with Three Mile Island Unit 2 Incident, dated May 1979.
- 2. S. Gordon, K. H. Schmidt and J. R. Honekamp, An Analysis of the Hydrogen Bubble Concerns in the Three Mile Island Unit 2 Reactor Vessel, Argonne National Laboratory.
- 3. NRC Memo and K. 1. Parczewski to Victor Senaroya, dated June 23, 1982.
4 n a
a s F O ' TR - O ti l l< e v , o 4 Iti g er 47
- 4. H. E. Zittel, Design Considerations of Reactor Consicarations 5 pray Systems - Boiling Water Reactor Acctdent Studies, ORM*TM-2412, Part Vill, Octocer 1970.
- 5. NUREG/CR-2367, Upaated Best-Estimate LOCA Radiaticn Signature, dated August 1981.
Principal Contributor: K. Parczewski Dated: July 6, 1989