ML14178A485
| ML14178A485 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 06/25/2014 |
| From: | South Texas |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| GSI-191, NOC-AE-14003101, TAC MF2400, TAC MF2401 | |
| Download: ML14178A485 (133) | |
Text
NOC-AE-1 4003101Attachment 1Enclosure 1Enclosure 1 to Attachment 1Supporting resolution of APLAB, CASA Grande-LOCA Frequencies: RAI 2STP-RIGSI191-RAI-APLA-III-2, Rev. 1, "RAI APLA-III-2: Modeling LOCAFrequency and Break Size under DEGB only Breaks", University of Texas NOC-AE-14003101Attachment 1Enclosure 110South Texas Project Risk-Informed GSI-191 EvaluationRAI APLA-III-2: Modeling LOCA Frequency and Break Sizeunder DEGB-only BreaksDocument: STP-RIGSI191-RAI-APLA-III-2Revision: 1Date: May 13, 2014Prepared by:John Hasenbein, The University of Texas at AustinDavid Morton, The University of Texas at AustinJeremy Tejada, The University of Texas at AustinReviewed by:Zahra Mohaghegh, University of Illinois at Urbana ChampaignSeyed A. Reihani, University of Illinois at Urbana ChampaignApproved by:Ernie J. Kee, South Texas Project NOC-AE-14003101Attachment 1Enclosure 1RAI APLA-III-2: Modeling LOCA Frequency and Break Sizeunder DEGB-only BreaksJohn Hasenbein, David Morton, and Jeremy TejadaThe University of Texas at Austin1 RAI APLA-I1-2The statement of RAI APLA-III-2 is as follows:RG 1.174, Section 2.3.4, "Plant Representation," states that PRA results should bederived from a model that realistically represents the risk associated with the plant.NUREG-1829 states that, in general, a complete rupture of a pipe is more likely thana partial rupture. It appears, however, that STP's methodology leads to the oppositeresult (i.e., a rupture of a given size is more likely to be caused by a partial ruptureof a large pipe than a complete rupture of a smaller pipe). Please illustrate the resultsof your method by comparing the frequency of partial versus complete breaks for a setof representative pipe sizes. Please describe whether the methodology described in theSTP pilot is consistent with the assumption of NUREG-1829 or provide justification foran alternate approach.2 IntroductionWe modify the analysis in the hybrid LOCA document of Pan et al. (2013) to handle the case inwhich welds only experience double-ended guillotine breaks (DEGBs). We refer to the resultingmodel as the DEGB-only model. We compare these results with the results of Pan et al. (2013),which we refer to as the continuum break-size model. We limit reintroduction of the notationdeveloped in Pan et al. (2013), except as it is modified for the DEGCB-only situation. As in ourearlier analysis, we maintain consistency with NUREG-1829's (Tregoning et al., 2008) initiatingfrequencies. First, we use the bottom-up conditional weights derived from Fleming et al. (2011),which is the approach we recommend. (Using the bottom-up weights, in conjunction with the con-tinuum model and NUREG-1829's frequencies, constitutes the hybrid method that is implementedin CASA Grande.) Then, we study the results of employing a top-down approach, which allows usto highlight the effect on DEGB probabilities of using the bottomn-up methodology.1 NOC-AE-14003101Attachment 1Enclosure 1Let j -J index STP break sizes. As before we compute the probability of a LOCA being incategory j (catj) using the formulaP [catj Frequency[LOCA > catj] -Frequency[LOCA > catj+1]Frequency[LOCA > cat] ,(1)where Frequency[LOCA > catj] is consistent with NUREG-1829. By consistentwith NUREG-1829we mean the following: We model Frequency[LOCA > catjJ as a random variable using a boundedJohnson distribution, and we do so for the six size categories elicited in NUREG-1829 (see Table 1).If we take the median of the Johnson distribution for a category j, which coincides with a categoryin NUREG-1829, then Frequency[LOCA > catj] is simply the median value from the elicitation(Tregoning et al., 2008, Table 7.19, page 7-55) for the current-day fleet. If j does not coincidewith a NUTREG-1829 category, and we still focus on the median frequency, then we approximateFrequency[LOCA > catj] by interpolating linearly between adjacent NUREG-1829 categories.When we consider a percentile of the Johnson distribution other than the median (or the 5thpercentile or 95th percentile) then we interpolate linearly between the corresponding percentile inJohnson distributions fit to adjacent NUREG-1829 categories. The distinction between the analysiswe conduct here and in Pan et al. (2013) concerns how we allocate these frequencies to specificweld cases.Table 1: LOCA categories from NUREG-1829.Effective Break Size(Inches) Notation75 cat,1-5 cat23 cat37 cat414 cat531 cat6We compute the probability that weld case i will experience a break of type j, usingP[catj at weldi] wjP[catj], (2)where zu = P(weldiccatj) is the conditional probability of the break occurring at weld i given acategory j break. We use set Ij to denote the set of weld cases that can experience a break of type2 NOC-AE-14003101Attachment 1Enclosure 1j, and we use' set I to index all 45 STP weld cases as described in Fleming et al. (2011) (see alsoTable 12 in the Appendix).Computation of zv is as follows. The bottom-up approach of Fleming et al. (2011) generatesthe frequency of category j breaks for weld case i, which we denote Freqb,,[LOCA > catj at weldi].Also, there are specific numbers of welds for each weld case i, and we denote the number of weldsfor weld case i by ni. Given these frequencies and the number of welds for each weld case, the wtovalues are computed as_ (Freqb,.[LOCA > catj at 'weldj] -Freqb.[LOCA > catj+l at weldi]) x niw Zicij (FreqbJ[OCA > catj at weldi] -Freqba[LOCA > catj-l at weldj]) x ?.7i (Given P[catj] from equation (1) and wju from equation (3), we form P[catj at weldi] via equation (2).As before, because the sum of all w} across i E Ij is equal to one, this approach matches theNUREG-1829 specified values for P[catj].Mathematically, this approach is identical to that described in Pan et al. (2013), but the keydistinction is in the definition of set Ij. In the continuum model, a weld can experience breaks ofsize up to its diameter and then can experience a larger (by a factor of v/2, as we discuss furtherbelow) DEGB break. However, by assuming that welds only experience DEGB breaks, the setIy is redefined to be the set of welds that have size that corresponds exactly to the category jcorresponding to the size of the DEGB.3 Comparing Continuum Break-Size Model and DEGB-Only Break ModelTable 2 provides the joint probability mass function (2) corresponding to the median for the con-tinuum break-size model with the six NUREG-1829 categories and with weld cases aggregated bypipe size. (All results we report here are for the median.) The first numerical entry in the tableindicates, given that we have a break, there is a probability of 0.586 that it is a category 1 break ona 1-inch pipe. The table's right-most column indicates the marginal probability of having a breakin each break-size interval: [1, 2), [2, 2.5), ... , [29, 31), [31, DEGB], where all values are in inchesand DEGB is the largest possible DEGB break size, all using STP pipe sizes. The table's bottomrow indicates the probability a break is in NUREG category 1, category 2, etc. Table 3 reportsanalogous results for the DEGB-only model. Note that if a pipe experiences a DEGB then becauseboth ends of the pipe are exposed it is equivalent to a "one-sided break" on a pipe with a radiuslarger by a factor of v/2. This is why, for example, a DEGB break of a 10-inch pipe is classified as3 NOC-AE-14003101Attachment 1Enclosure 1a category 5 break in Tables 2 and 3. Tables 13 and 14 in the Appendix repeat the information inTables 2 and 3 except at the resolution of STP's 45 weld cases.As the tables indicate, the DEGB-only model increases the likelihood of breaks in small pipes.In the continuum model, small breaks can occur in a 31-inch pipe. However, in the DEGB-onlymodel small breaks can only occur in correspondingly small pipes. As a result the probability of abreak in a 31-inch pipe decreases by a factor of about 16,500 (=1.75E-03/1.06E-07). As the bottomrows in Tables 2 and 3 indicate, we preserve NUREG-1829 frequencies, and hence the probabilitiesof a category 1, 2, ..., 6 break are identical in both models. Also, note that the category 6 columnsare identical in Tables 2 and 3. This is because under both models a category 6 break must be aDEGB because STP's largest pipe size is 31 inches, which matches the low end of NUREG-1829'scategory 6 bin.Table 4 repeats the right-most columns of Tables 2 and 3, and the table also includes, for eachpipe size, the probability that weld cases of that pipe size experience a DEGB under both thecontinuum and DEGB-only models. The overall probability of a DEGB in the continuum model is0.165, and the bulk of that comes from small pipes. Table 15 repeats the information of Table 4 atthe resolution of STP's 45 weld cases.Tables 5 and 6 depict the relative contributions by category for each pipe size under the con-tinuum and DEGB-only models. For example, under the continuum model, given that we have acategory 2 break, the conditional probability it is in a 2-inch pipe is 0.1787. Tables 16 and 17 inthe Appendix report the same information but at the resolution of the 45 weld cases. Figures 1-3report these same conditional probabilities for categories 4-6 for the top few weld cases under boththe continuum break-size model and the DEGB-only model.4 NOC-AE-14003101Attachment 1Enclosure 1Table 2: LOCA probabilities for STP pipe sizes for current-day estimates when welds can experiencea contimnum of break sizes. Pipe sizes are in inches.Pipe Size Cat 1 Cat 2 Cat 3 Cat 4 Cat 5 Cat 6 P(Break)1.00 5.86E-01 X X X X X 5.86E-012.00 4.06E-02 2.43E-02 5.14E-05 X X X 6.49E-022.50 7.56E-04 3.58E-04 5.03E-05 X X X 1.16E-033.00 1.53E-03 7.59E-04 3.37E-05 X X X 2.33E-034.00 4.28E-03 2.12E-03 9.13E-05 X X X 6.49E-036.00 4.07E-03 2.02E-03 7.66E-05 6.69E-06 X X 6.17E-038.00 6.56E-02 3.30E-02 1.14E-03 1.06E-04 X X 9.98E-0210.00 6.46E-04 3.25E-04 1.13E-05 7.77E-07 3.23E-08 X 9.82E-0412.00 1.17E-01 5.88E-02 2.04E-03 1.41E-04 5.85E-06 X 1.78E-0116.00 1.7SE-02 8.44E-03 8.43E-04 1.31E-04 5.50E-06 X 2.72E-0227.50 1.25E-03 3.84E-04 2.86E-05 3.71E-06 2.59E-07 1.O1E-07 1.67E-0329.00 1.85E-02 5.03E-03 5.06E-04 7.96E-05 5.23E-06 1.70E-06 2.41E-0231.00 1.31E-03 4.04E-04 3.01E-05 3.91E-06 2.73E-07 1.06E-07 1.75E-03P(Break) 8.59E-01 1.36E-01 4.90E-03 4.73E-04 1.71E-05 1.90E-06 1.OOE+00Table 3: LOCA probabilities for STP pipe sizes for current-day estimates when welds can onlyexperience a DEGB. Pipe sizes are in inches.Pipe Size Cat 1 Cat 2 Cat 3 Cat 4 Cat 5 Cat 6 P(Break)1.00 8.59E-01 X X X X X 8.59E-012.00 X 1.36E-01 1.97E-03 X X X 1.38E-012.50 X X 1.51E-03 X X X 1.51E-033.00 X X 6.11E-04 X X X 6.11E-044.00 X X 8.18E-04 X X X 8.18E-046.00 X X X 4.65E-05 X X 4.65E-058.00 X X X 4.27E-04 X X 4.27E-0410.00 X X X X 8.79E-08 X 8.79E-0812.00 X X X X 1.09E-05 X 1.09E-0516.00 X X X X 6.19E-06 X 6.19E-0627.50 X X X X X 1.01E-07 1.O1E-0729.00 X X X X X 1.70E-06 1.70E-0631.00 X X X X X 1.06E-07 1.06E-07P(Break) L8.59E-011.36E-01 4.90E-03 4.73E-04 1.71E-0.51.90E-06 1.OOE+005 NOC-AE-14003101Attachment 1Enclosure 1Table 4: The table shows the probability of a. break occurring at each pipe size, and tile conditionalprobability that a break at a given size is a DEGB, using both the continuum break-size model andthe DEGB-only model. Below the table, the sums of the probabilities for the break sizes are shownto be one, and using those probabilities as weights we show the overall probability of a break beinga DEGB under both models. Pipe sizes are in inches.Continuous DEGB OnlyPipe SizeP(Break) P(DEGB) P(Break) P(DEGB)1.00 5.86E-01 2.69E-01 8.59E-01 1.00E+002.00 6.49E-02 1.13E-01 1.38E-01 1.00E+002.50 1.16E-03 3.37E-02 1.51E-03 1.OOE+003.00 2.33E-03 6.85E-03 6.11E-04 1.00E+004.00 6.49E-03 3.29E-03 8.18E-04 1.OOE+006.00 6.17E-03 7.28E-04 4.65E-05 1.00E+008.00 9.98E-02 4.13E-04 4.27E-04 1.OOE+0010.00 9.82E-04 3.20E-05 8.79E-08 1.OOE+0012.00 1.78E-01 2.19E-05 1.09E-05 1.00E+0016.00 2.72E-02 S.13E-05 6.19E-06 1.00E+0027.50 1.67E-03 4.31E-05 1.01E-07 1.00E+0029.00 2.41E-02 4.31E-05 1.70E-06 1.OOE+0031.00 1.75E-03 3.53E-05 1.06E-07 1.OOE+001.OOE+00 1.65E-01 1.00E+00 :1.00E+006 NOC-AE-14003101Attachment 1Enclosure 1Table 5: Relative contributions by category when welds can experience a continuum of break sizes.Pipe sizes are in inches.Pipe Size Cat 1 Cat 2 Cat 3 Cat 4 Cat 5 Cat 61.00 68.19% X X X X X2.00 4.72% 17.87% 1.05% X X X2.50 0.09% 0.26% 1.03% X X X3.00 0.18% 0.56% 0.69% X X X4.00 0.50% 1.56% 1.86% X X X6.00 0.47% 1.48% 1.56% 1.41% X X8.00 7.64% 24.28% 23.32% 22.50% X X10.00 0.08% 0.24% 0.23% 0.16% 0.19% X12.00 13.60% 43.25% 41.55% 29.73% 34.13% X16.00 2.08% 6.21% 17.19% 27.76% 32.09% X27.50 0.15% 0.28% 0.58% 0.78% 1.51% 5.30%29.00 2.16% 3.70% 10.33% 16.82% 30.48% 89.13%31.00 0.15% 0.30% 0.61% 0.83% 1.59% 5.57%Total 100.00% 100.00% 100.00% 100.00% 100.00% 100.00%Table 6: Relative contributions by category when welds can experience only a DEGB. Pipe sizesare in inches.Pipe Size Cat I Cat 2 Cat 3 Cat 4 Cat 5 Cat 61.00 100.00% X X X X X2.00 X 100.00% 40.15% X X X2.50 X X 30.71% X X X3.00 X X 12.45% X X X4.00 X X 16.69% X X X6.00 X X X 9.83% X X8.00 X X X 90.17% X X10.00 X X X X 0.51% X12.00 X X X X 63.37% X16.00 X X X X 36.11% X27.50 X X X X X 5.30%29.00 X X X X X 89.13%31.00 X X X X X 5.57%Total 100.00% 100.00% 100.00% 100.00% 100.00% 100.00%7 NOC-AE-14003101Attachment 1Enclosure 1Probability of Break By Weld Case -Category 4:Continuum Break Size Model* Weld 4- 2* Weld 9 -4AWeld 25 -7AWeld 26 -7B* Weld 27 -7CTotal % = 89.99%(a) Continuum modelProbability of Break By Weld Case -Category 4:DEGB Only Model* Weld 13 -5A* Weld 26 -7BWeld 27 -7CTotal % = 94.98%(b) DEGB-only modelFigure 1: This figure depicts the probability of a break occurring in each weld case conditional onthe break being in category 4. Part (a) of the figure corresponds to the continuum break-size modeland part (b) corresponds to the DEGB-only model.8 NOC-AE-14003101Attachment 1Enclosure 1Probability of Break By Weld Case -Category 5:Continuum Break Size Model" Weld 1 -IA" Weld 4 -2Weld 9 -4A" Weld 25 -7ATotal % = 92.26%(a) Continuum modelProbability of Break By Weld Case -Category 5:DEGB Only Model" Weld 9 -4A" Weld 25 -7ATotal % = 93.16%(b) DEGB-only modelFigure 2: This figure depicts the probability of a break occurring in each weld case conditional onthe break being in category 5. Part (a) of the figure corresponds to the continuum break-size modeland part (b) corresponds to the DEGB-only model.9 NOC-AE-14003101Attachment 1Enclosure 1Probability of Break By Weld Case -Category 6:Continuum Break Size Model5.02% 5.02%* Weld 1 -1A* Weld 4- 2Weld 5 -3AWeld 6 -31Total % = 98.86%(a) Continuum modelProbability of Break By Weld Case -Category 6:Continuum Break Size Model5.02% , 5.02%* Weld 1 -1A* Weld 4 -2* Weld 5 -3AWeld 6 -31Total % = 98.86%(b) DEGB-only modelFigure 3: This figure depicts the probability of a break occurring in each weld case conditional onthe break being in category 6. Part (a) of the figure corresponds to the continuum break-size modeland part (b) corresponds to the DEGB-only model. As we indicate in the text, these conditionalprobabilities are identical under the continuum- and DEGB-only model for category 6.10 NOC-AE-14003101Attachment 1Enclosure 14 Continuum and DEGB-Only Models under a Top-Down AnalysisA top-down analysis is discussed in Pan et al. (2013) and in that model = 1/J1Ij; i.e., we ignorerelevant information regarding degradation mechanisms and simply say that each weld which canexperience a break of a particular size is equally likely to have the break. As we discuss above, thedefinitions of the sets Ij differ under the continuum break-size model and the DEGB-only model,and hence the corresponding top-down models differ. In this section, we provide tables analogousto those from the previous section using the top-down models.Table 7 provides the top-down joint probability mass function (2) analogous the one presentedin Table 2 for the continuum break-size model, and Table 8 reports analogous results for the DEGB-only model. As with the hybrid approach in which we use bottom-up values, under the top-downapproach the DEGB-only model increases the likelihood of breaks in small pipes. WVe again preserveNUREG-1829 frequencies, and hence the probabilities of a category 1, 2, ... , 6 break are identicalin both models and identical to the values we obtain in Section 3. As in the previous section, theprobabilities in the category 6 columns are identical under the continuum and DEGB-only models.As Table 9 indicates, the probability of a DEGB in the continuum break-size model is 0.0746,with the bulk of that coming from small pipe sizes. This value is less than half of the value tinderthe hybrid approach, indicating that using weights from the bottom-up approach increases theprobability of a DEGB over simply using the NUREG-1829 values and assigning equal weights topipes of equal sizes., Tables 10 and 11 specify the relative contributions by category for each pipesize under the top-down variants of the continuum and DEGB-only models.11 NOC-AE-14003101Attachment 1Enclosure 1Table 7: LOCA probabilities for STP pipe sizes for current-day estimates when welds can experiencea continuum of break sizes under the top-down model. Pipe sizes are in inches.Pipe Size Cat 1 Cat 2 Cat 3 Cat 4 Cat 5 Cat 6 P(Break)1.00 2.39E-01 X X X X X 2.39E-012.00 9.05E-02 2.01E-02 5.38E-04 X X X 1.11E-012.50 6.39E-03 1.40E-03 5.66E-05 X X X 7.84E-033.00 2.87E-02 6.29E-03 2.55E-04 X X X 3.53E-024.00 9.58E-02 2.IOE-02 8.49E-04 X X X 1.18E-016.00 9.05E-02 1.98E-02 7.29E-04 1.1OE-04 X X i.1E-018.00 5.75E-02 1.26E-02 4.63E-04 7.OOE-05 X X 7.06E-0210.00 3.19E-02 6.99E-03 2.57E-04 3.74E-05 2.25E-06 X 3.92E-0212.00 1.39E-01 3.05E-02 1.12E-03 1.63E-04 9.82E-06 X 1.71E-0116.00 1.06E-02 2.33E-03 8.57E-05 1.25E-05 7.50E-07 X 1.31E-0227.50 1.70E-02 3.73E-03 1.37E-04 1.99E-05 1.08E-06 4.76E-07 2.09E-0229.00 2.13E-02 4.66E-03 1.71E-04 2.49E-05 1.35E-06 5.95E-07 2.62E-0231.00 2.98E-02 6.52E-03 2.40E-04 3.49E-05 1.89E-06 8.33E-07 3.66E-02P(Break) 8.59E-01 1.36E-01 4.90E-03 4.73E-04 1.71E-05 1.90E-06 1.OOE+00Table 8: LOCA probabilities for STP pipe sizes for current-day estimates when welds can onlyexperience a DEGB under the top-down model. Pipe sizes are in inches.Pipe Size Cat 1 Cat 2 Cat 3 Cat 4 Cat 5 Cat 6 P(Break)1.00 8.59E-01 X X X X X 8.59E-012.00 X 1.36E-01 2.38E-03 X X X 1.38E-012.50 X X 2.20E-04 X X X 2.20E-043.00 X X 8.10E-04 X X X 8.1OE-044.00 X X 1.49E-03 X X X 1.49E-036.00 X X X 3.57E-04 X X 3.57E-048.00 X X X 1.16E-04 X X 1.16E-0410.00 X X X X 3.51E-06 X 3.51E-0612.00 X X X N 1.30E-05 X 1.30E-0516.00 X X X X 6.40E-07 X 6.40E-0727.50 X X X X X 4.76E-07 4.76E-0729.00 X X X X X 5.95E-07 5.95E-0731.00 X X X X X 8.33E-07 8.33E-07P(Break) 8.59E-01 1.36E-01 4.90E-03 4.73E-04 1.71E-05 1.90E-06 1.OOE+0012 NOC-AE-14003101Attachment 1Enclosure 1Table 9: Under the top-down model, the table shows the probability of a break occurring at eachpipe size, and the probability that a break at a given size is a DEGB, using both the continuumbreak-size model and the DEGB-only model. Below the table, the sums of the probabilities forthe break sizes are shown to be one, and using those probabilities as weights we show the overallprobability of a break being a DEGB under both models. Pipe sizes are in inches.Continuous DEGB OnlyPipe SizeP(Break) P(DEGB) P(Break) P(DEGB)1.00 2.39E-01 3.02E-01 8,59E-01 1.00E+002.00 1i.E-01 1.45E-02 1.38E-01 1.00E+002.50 7.84E-03 6.34E-03 2.20E-04 1.OOE+003.00 3.53E-02 5.18E-03 8.10E-04 1.OOE+004.00 1.18E-01 2.86E-03 1.49E-03 1.00E+006.00 1i.E-01 7.89E-04 3.57E-04 1.00E+008.00 7.06E-02 4.04E-04 1.16E-04 1.OOE+0010,00 3.92E-02 5.69E-05 3.51E-06 1.00E+0012.00 1.71E-01 4.33E-05 1.30E-05 1.00E+0016.00 1.31E-02 3.12E-05 6.40E-07 1.00E+0027.50 2.09E-02 2.28E-05 4.76E-07 1.00E+0029.00 2.61E-02 2.28E-05 5.95E-07 1.OOE+0031.00 3.66E-02 2.28E-05 8.33E-07 1.00E+001.00E+00 7.46E-02 1.00E+00 1.OOE-0013 NOC-AE-14003101Attachment 1Enclosure 1Table 10: Relative contributions by category when welds can experience a continuum of break sizesunder the top-down model. Pipe sizes are in inches.Pipe SizeCat 1 Cat 2 Cat 3 Cat 4 Cat 5Cat 61.00 27.86% X X X X X2.00 10.54% 14.77% 10.96% X X X2.50 0.74% 1.03% 1.15% X X X3.00 3.35% 4.63% 5.19% X X X4.00 11.16% 15.43% 17.31% X X X6.00 10.54% 14.58% 14.86% 23.29% X X8.00 6.69% 9.26% 9.44% 14.80% X X10.00 3.72% 5.14% 5.24% 7.90% 13.12% X12.00 16.24% 22.47% 22.90% 34.51% 57.31% X16.00 1.24% 1.71% 1.75% 2.63% 4.37% X27.50 1.98% 2.74% 2.80% 4.21% 6.30% 25.00%29.00 2.48% 3.43% 3.50% 5.27% 7.87% 31.25%31.00 3.47% 4. 80% 4.89% 7.38% 11.02% 43.75%Total 100.00% 100.00% 100.00% 100.00% 100.00% 100.00%Table 11: Relative contributions by category when welds can experience only a DEGB under thetop-down model. Pipe sizes are in inches.Pipe Size Cat I Cat 2 Cat 3 Cat 4 Cat 5 Cat 61.00 100.00% X X X X X1.50 X X X X X X2.00 X 100.00% 48.59% X X X2.50 X X 4.49% X X X3.00 X X 16.52% X X X4.00 X X 30.41% X X X6.00 X X X 75.45% X X8.00 X X X 24.55% X X10.00 X X X X 20.45% X12.00 X X X X 75.82% X16.00 X X X X 3.73% X27.50 X X X X X 25.00%29.00 X X X X X 31.25%31.00 X X X X X 43.75%Total 100.00% 100.00% 100.00% 100.00% 100.00% 100.00%14 NOC-AE-14003101Attachment 1Enclosure 15 SummaryThe analysis we present here indicates that under the implemented "continuum" model, the proba-bility a pipe experiences a DEGB, given that it has a break, is 0.165. We also present a DEGB-onlymodel in which all breaks are DEGBs, and we compare how that model allocates breaks to weldcases, relative to the continuum model. We preserve NUREG-1829 frequencies throughout, andhence the latter model dramatically decreases the probability that a large pipe experiences a break.Finally, we put aside the bottom-up frequencies and instead assume each pipe within a size categoryis equally likely to experience a break, given we have a break in that category. Under this top-downapproach for the continuum model, the probability a pipe experiences a DEGB, given that it hasa break, is 0.0746, indicating that using the bottom-up frequencies increases the probability of aDEGB by more than a factor of two.15 NOC-AE-14003101Attachment 1Enclosure 1ReferencesFleming, K. N., B. 0. Lydell, and D. Chrun (2011, July). Development of LOCA Initiating EventFrequencies for South Texas Project GSI-191. Technical Report, KnF Consulting Services, LLC,Spokane, WA.Pan, Y.-A., E. Popova, and D. P. Morton (2013, January). South Texas Project Risk-Informed GSI-191 Evaluation, Volume 3, Modeling and Sampling LOCA Frequency and Break Size. Technicalreport, STP-RIGS1191-V03.02, Revision 4, The University of Texas at Austin.Tregoning, R., P. Scott, and A. Csontos (2008, April). Estimating Loss-of-Coolant Accident(LOCA) Frequencies Through the Elicitation Process: Main Report (NUREG-1829). NUREG1829, NRC, Washington, DC.16 NOC-AE-14003101Attachment 1Enclosure 1AppendixThe appendix presents the results for each weld case as opposed to results which are aggregatedacross break sizes.Table 12: Characteristics of the 45 Weld Cases at STP1 Weld 1 Damage 1 # of T Pipe Size DEGB SizeSystem Weld # Case Mechanisms Welds (in.) (in.)Hot Leg 1 IA SC+D&C 4 29.00 41.01Hot Leg 2 1B D&C 11 29.00 41.01Hot Leg 3 1C TF+D&C' 1 29.00 41.01SC Inlet 4 2 SC+D&C 4 29.00 41.01Cold Leg 5 3A SC+D&C 4 27.50 :38.89Cold Leg 6 3B SC+D&C 4 31.00 43.84Cold Leg 7 3C D&,C 12 27.50 38.89Cold Leg 8 3D D&C 24 31.00 43.84Surge Line 9 4A SC+TF+D&C 1 16.00 22.63Surge Line 10 4B TF+D&C 7 16.00 22.63Surge Line 11 4C TF+D&C 2 16.00 22.63Surge Line 12 4D TF+D&C 6 2.50 3.54Pressurizer 13 5A TF+D&C 29 6.00 8.49Pressurizer 14 5B TF+D&C 14 3.00 4.24Pressurizer 15 5C D&C 53 4.00 5.66Pressurizer 16 5D D&,C 4 3.00 4.24Pressurizer 17 5E D&C 29 6.00 8.49Pressurizer 18 5F SC+D&C 0 6.00 8.49Pressurizer 19 5G D&C (Weld Overlay) 4 6.00 8.49Pressurizer 20 5H D&C 2 4.00 5.66Pressurizer 21 5I TF+D&C 2 2.00 2.83Pressurizer 22 53 SC+TF+D&C 0 6.00 8.49Small Bore 23 6A VF+SC+D&C 16 2.00 2.83Small Bore 24 6B VF+SC+D&C 193 1.00 1.41SIR 25 7A TF+D&C 21 12.00 16.97SIR 26 7B TF+D&C 9 8.00 11.31SIR 27 7C SC+TF+D&C 3 8.00 11.31SIR 28 7D SC+D&C 3 12.00 16.97SIR 29 7E D&C 57 12.00 16.97SIR 30 7F D&C 30 10.00 14.14SIR 31 7G D&C 42 8.00 11.31SIR 32 7H D&C 2:3 6.00 8.49SIR 33 71 D&C 5 4.00 5.66SIR 34 7J D&C, 9 3.00 4.24SIR 35 7K D&C 10 2.00 2.83SIR 36 7L D&C 0 1.50 2.12ACC 37 7M SC+D&C 0 12.00 16.97ACC 38 7N TF+D&C 35 12.00 16.97ACC 39 70 D&SC 15 12.00 16.97CVCS 40 SA TF+VF+D&C 10 2.00 2.83CVCS 41 8B TF+VF+D&C 19 4.00 5.66CVCS 42 8C VF+D&C 47 2.00 2.83CVCS 43 8D VF+D&SC 6 4.00 5.66CVCS 44 8E TF+D&C 4 4.00 5.66CVCS 45 8F D&C 1 4.00 5.6677517 NOC-AE-14003101Attachment 1Enclosure 1Table 13: LOCA probabilities for the 45 weld cases for current-day estimates when welds canexperience a continuum of break sizes.Weld Case Cat 1 Cat 2 Cat 3 Cat 4 Cat 5 Cat 6 P(Break)Weld 1 -IA 3.13E-03 8.44E-04 8.43E-05 1.27E-05 9.16E-07 2.80E-07 4.07E-03Weld 2 -lB 4.17E-05 1.13E-05 1.13E-06 1.70E-07 1.22E-08 3.73E-09 5.43E-05Weld 3 -1C 2.43E-05 6.56E-06 6.55E-07 9.88E-08 7.12E-09 2.17E-09 3.16E-05Weld 4 -2 1.53E-02 4.16E-03 4.20E-04 6.66E-05 4.29E-06 1.41E-06 2.OOE-02Weld 5 -3A 1.18E-03 3.64E-04 2.71E-05 3.52E-06 2.46E-07 9.56E-08 1.58E-03Weld 6 -3B 1.1SE-03 3.64E-04 2.71E-05 3.52E-06 2.46E-07 9.56E-08 1.58E-03Weld 7 -3C 6.56E-05 2.01E-05 1.50E-06 1.95E-07 1.36E-08 5.29E-09 8.74E-05Weld 8 -3D 1.31E-04 4.03E-05 3.OOE-06 3.90E-07 2.72E-08 1.06E-08 1.75E-04Weld 9 -4A 1.65E-02 7.83E-03 7.82E-04 1.22E-04 5.10E-06 X 2.53E-02Weld 10 -4B 8.83E-04 4.18E-04 4.18E-05 6.50E-06 2.72E-07 X 1.35E-03Weld 11 -4C 4.11E-04 1.95E-04 1.95E-05 3.03E-06 1.27E-07 X 6.29E-04Weld 12 -4D 7.56E-04 3.58E-04 5.03E-05 X X X 1.16E-03Weld 13 -5A 2.51E-03 1.24E-03 4.76E-05 4.13E-06 X X 3.80E-03Weld 14 -5B 1.21E-03 5.98E-04 2.68E-05 X X X 1.84E-03Weld 15 -5C 1.72E-03 8.47E-04 3.80E-05 X X X 2.60E-03Weld 16 -5D 1.30E-04 6.39E-05 2.87E-06 X X X 1.96E-04Weld 17 -5E 9.39E-04 4.64E-04 1.78E-05 1.55E-06 X X 1.42E-03Weld 18 -5F O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 X X 0.OOE+00Weld 19- 5G 1.31E-04 6.48E-05 2.49E-06 2.16E-07 X X 1.99E-04Weld 20 -5H 6.48E-05 3.20E-05 1.43E-06 X X X 9.82E-05Weld 21 -5I 1.73E-04 1.14E-04 X X X X 2.87E-04Weld 22 -5J O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 X X O.OOE+00Weld 23 -6A 3.77E-02 2.28E-02 X X X X 6.05E-02Weld 24 -6B 5.86E-01 X X X X X 5.86E-01Weld 25 -7A 1.10E-01 5.53E-02 1.92E-03 1.32E-04 5.51E-06 X 1.67E-01Weld 26 -7B 4.71E-02 2.37E-02 8.22E-04 7.65E-05 X X 7.1SE-02Weld 27- 7C 1.75E-02 8.82E-03 3.06E-04 2.85E-05 X X 2.67E-02Weld 28 -7D 2.OOE-03 L.O0E-03 3.50E-05 2.41E-06 l.OOE-07 X 3.05E-03Weld 29 -7E 1.23E-03 6.17E-04 2.14E-05 1.48E-06 6.14E-08 X 1.87E-03Weld 30 -7F 6.46E-04 3.25E-04 1.13E-05 7.77E-07 3.23E-08 X 9.82E-04Weld 31 -7G 9.04E-04 4.55E-04 1.58E-05 1.47E-06 X X 1.38E-03Weld 32 -7H 4.95E-04 2.49E-04 8.63E-06 8.03E-07 X X 7.53E-04Weld 33 -71 L.OSE-04 5.41E-05 2.21E-06 X X. X 1.64E-04Weld 34 -7J 1.94E-04 9.74E-05 3.98E-06 X X X 2.95E-04Weld 35 -7K 2.15E-04 1.41E-04 X X X X 3.57E-04Weld 36 -7L 0.OOE+00 0.OOE+00 X X X. X 0.00E+00Weld 37 -7M O.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 X 0.OOE+00Weld 38 -7N 3.42E-03 1.72E-03 5.99E-05 4.15E-06 1.72E-07 X 5.21E-03Weld 39 -70 1.77E-04 8.90E-05 3.10E-06 2.15E-07 8.92E-09 X 2.69E-04Weld 40 -8A 8.09E-04 4.03E-04 1.68E-05 X X X 1.23E-03Weld 41 -8B 1.54E-03 7.65E-04 3.20E-05 X X X 2.33E-03Weld 42 -8C 1.66E-03 8.28E-04 3.46E-05 X X X 2.53E-03Weld 43 -8D 2.12E-04 1.06E-04 4.41E-06 X X X 3.22E-04Weld 44 -8E 6.03E-04 3.OOE-04 1.25E-05 X X X 9.16E-04Weld 45 -8F 3.54E-05 1.76E-05 7.36E-07 X X X 5.37E-05P(Break) 8.59E-01 1.36E-01 4.90E-03 4.73E-04 1.71E-05 1.90E-06 I.00E+0018 NOC-AE-14003101Attachment 1Enclosure 1Table 14: LOCA probabilities for the 45 weld cases for current-day estimates when welds canexperience only a DEGB.Weld Case Cat 1 Cat 2 Cat 3 Cat 4 Cat 5 Cat 6 P(Break)Weld 1 -1A X X X X X 2.80E-07 2.80E-07Weld 2 -1B X X X X X 3.73E-09 3.73E-09Weld 3 -IC X X X X X 2.17E-09 2.1TE-09Weld 4 -2 X X X X X 1.41E-06 1.41E-06Weld 5 -3A X X X X X 9.56E-08 9.56E-08Weld 6 -3B X X X X X 9.56E-OS 9.56E-08Weld 7 -3C X X X X X 5.29E-09 5.29E-09Weld 8 -3D X X X X X 1.06E-08 1.06E-08Weld 9 -4A X X X X 5.74E-06 X 5.74E-06Weld 10 -4B X X X X 3.07E-07 X 3.07E-07Weld 11 -4C X X X X 1.43E-07 X 1.43E-07Weld 12 -4D X X 1.51E-03 X X X 1.51E-03Weld 13 -5A X X X 2.86E-05 X X 2.86E-05Weld 14 -5B X X 4.86E-04 X X X 4.86E-04Weld 15 -5C X X 3.31E-04 X X X 3.31E-04Weld 16 -5D X X 5.20E-05 X X X 5.20E-05Weld 17 -5E X X X 1.07E-05 X X 1.07E-05Weld 18 -5F X X X O.OOE+00 X X O.OOE+00Weld 19 -5G X X X 1.50E-06 X X 1.50E-06Weld 20 -5H X X 1.25E-05 X X X 1.25E-05Weld 21 -5I X 6.08E-04 X X X X 6.08E-04Weld 22 -5J X X X O.OOE+00 X X O.OOE+00Weld 23 -6A X 1.35E-01 X X X X 1.35E-01Weld 24 -6B 8.59E-01 X X X X X 8.59E-01Weld 25 -7A X X X X 1.02E-05 X 1.02E-05Weld 26 -7B X X X 3.07E-04 X X 3.07E-04Weld 27 -7C X X X 1.14E-04 X X 1.14E-04Weld 28 -7D X X X X 1.86E-07 X 1.86E-07Weld 29 -7E X X X X 1.14E-07 X 1.14E-07Weld 30 -7F X X X X 8.79E-08 X 8.79E-08Weld 31 -7G X X X 5.88E-06 X X 5.88E-06Weld 32 -7H X X X 5.66E-06 X X 5.66E-06Weld 33 -71 X X 2.08E-05 X X X 2.08E-05Weld 34 -7J X X 7.30E-05 X X X 7.30E-05Weld 35 -7K X 6.66E-04 X X X X 6.66E-04Weld 36 -7L X O.OOE+00 X X X X O.OOE+00Weld 37 -7M X X X X O.OOE+00 X O.OOE+00Weld 38 -7N X X X X 3.18E-07 X 3.18E-07Weld 39 -70 X X X X 1.65E-08 X 1.65E-08Weld 40 -8A X X 6.45E-04 X X X 6.45E-04Weld 41 -8B X X 2.92E-04 X X X 2.92E-04Weld 42 -8C X X 1.32E-03 X X X 1.32E-03Weld 43 -8D X X 4.04E-05 X X X 4.04E-05Weld 44 -8E X X 1.15E-04 X X X 1.15E-04Weld 45 -8F X X 6.73E-06 X X X 6.73E-06P(Break) 8.59E-01 1.36E-01 4.90E-03 4.73E-04 1.71E-05 1.90E-06 1.00E+0019 NOC-AE-14003101Attachment 1Enclosure 1Table 15: This table shows probability of a break occurring at each weld case and the probabilitythat a break at a given weld case is a DEGB using both the continuum break-size model and theDEGB-only model. Below the table, the sum of the probabilities for the weld cases are shown tobe one, and using those probabilities as weights we show the overall probability of a break being aDEGB break across all weld cases.WlCaeI Continuous I DEGB OnlyP(Break) P(DEGB) P(Break) P(DEGB)Weld 1 -1A 4.07E-03 4.21E-05 2.80E-07 1.00E+00Weld 2 -1B 5.43E-05 4.21E-05 3.73E-09 1.00E+00Weld 3 -IC 3.16E-05 4.21E-05 2.17E-09 1.00E+00Weld 4 -2 2.OOE-02 4.33E-05 1.41E-06 1.00E+00Weld 5 -3A 1.58E-03 4.31E-05 9.56E-08 1.O0E+00Weld 6 -3B 1.58E-03 3.53E-05 9.56E-08 1.00E+00Weld 7 -3C 8.74E-05 4.31E-05 5.29E-09 1.OOE+00Weld 8 -3D 1.75E-04 3.53E-05 1.06E-08 1.00E+00Weld 9 -4A 2.53E-02 8.13E-05 5.74E-06 1.OOE+00Weld 10 -4B 1.35E-03 8.13E-05 3.07E-07 1.00E+00Weld 11 -4C 6.29E-04 8.13E-05 1.43E-07 1.OOE+00Weld 12 -4D 1.16E-03 3.37E-02 1.51E-03 1.00E+00Weld 13 -5A 3.80E-03 7.28E-04 2.86E-05 1.00E+00Weld 14- 5B 1.84E-03 6.91E-03 4.86E-04 1.00E+/-00Weld 15- 5C 2.60E-03 3.32E-03 3.31E-04 1.00E+00Weld 16- 5D 1.96E-04 6.91E-03 5.20E-05 1.OOE+00Weld 17- 5E 1.42E-03 7.28E-04 1.07E-05 1.00E+00Weld 18 -5F 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00Weld 19 -5G 1.99E-04 7.28E-04 1.50E-06 1.00E+00Weld 20 -5H 9.82E-05 3.32E-03 1.25E-05 1.00E+00Weld 21 -5I 2.87E-04 1.13E-01 6.08E-04 1.00E+00Weld 22 -5J 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00Weld 23 -6A 6.05E-02 1.19E-01 1.35E-01 1.OOE+00Weld 24 -16B 5.86E-01 2.69E-01 8.59E-01 1.00E+00Weld 25 -7A 1.67E-01 2.19E-05 1.02E-05 1.00E+00Weld 26 -7B 7.18E-02 4.13E-04 3.07E-04 1.00E+00Weld 27- 7C 2.67E-02 4.13E-04 1.14E-04 1.00E+00Weld 28 -7D 3.05E-03 2.19E-05 1.86E-07 1.00E+00Weld 29 -7E 1.87E-03 2.19E-05 1.14E-07 1.00E+00Weld 30 -7F 9.82E-04 3.20E-05 8.79E-08 1.00E+00Weld 31 -7G 1.38E-03 4.13E-04 5.88E-06 1.00E+00Weld 32 -7H 7.53E-04 7.26E-04 5.66E-06 1.OOE+00Weld 33 -71 1.64E-04 3.31E-03 2.08E-05 1.00E+00Weld 34 -7J 2.95E-04 6.46E-03 7.30E-05 1.OOE+00Weld 35 -7K 3.57E-04 1.00E-01 6.66E-04 1.00E+00Weld 36 -7L 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00Weld 37 -7M O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00Weld 38 -7N 5.21E-03 2.19E-05 3.18E-07 1.OOE+00Weld 39 -70 2.69E-04 2.19E-05 1.65E-08 1.00E+00Weld 40 -8A 1.23E-03 1.37E-02 6.45E-04 1.00E+00Weld 41 -8B 2.33E-03 3.27E-03 2.92E-04 1.00E+00Weld 42 -8C 2.53E-03 1.37E-02 1.32E-03 1.OOE+00Weld 43 -8D 3.22E-04 3.27E-03 4.04E-05 1.00E+00Weld 44 -8E 9.16E-04 3.27E-03 1.15E-04 1.OOE+00Weld 45 -8F 5.37E-05 3.27E-03 6.73E-06 1.00E+001.00E+00 1.65E-01 1.OOE+00 1.OOE-0020 NOC-AE-14003101Attachment 1Enclosure 1Table 16: Relative contributions by category when welds can experience a continuum of break sizes.Weld Case Cat I Cat 2 Cat 3 Cat 4 Cat 5 Cat 6Weld 1 -IA 0.36% 0.62% 1.72% 2.69% 5.34% 14.69%Weld 2 -1B 0.00% 0.01% 0.02% 0.04% 0.07% 0.20%Weld 3 -1C 0.00% 0.00% 0.01% 0.02% 0.04% 0.11%Weld 4 -2 1.79% 3.06% 8.57% 14.08% 25.03% 74.13%Weld 5 -3A 0.14% 0.27% 0.55% 0.74% 1.43% 5.02%Weld 6 -3B 0.14% 0.27% 0.55% 0.74% 1.43% 5.02%Weld 7 -3C 0.01% 0.01% 0.03% 0.04% 0.08% 0.28%Weld 8 -3D 0.02% 0.03% 0.06% 0.08% 0.16% 0.56%Weld 9 -4A 1.92% 5.76% 15.94% 25.74% 29.76% XWeld 10 -4B 0.10% 0.31% 0.85% 1.37% 1.59% XWeld 11 -4C 0.05% 0.14% 0.40% 0.64% 0.74% XWeld 12 -4D 0.09% 0.26% 1.03% X X XWeld 13 -5A 0.29% 0.91% 0.97% 0.87% X XWeld 14 -5B 0.14% 0.44% 0.55% X X XWeld 15 -5C 0.20% 0.62% 0.78% X X XWeld 16 -5D 0.02% 0.05% 0.06% X X XWeld 17 -5E 0.11% 0.34% 0.36% 0.33% X XWeld 18 -5F 0.00% 0.00% 0.00% 0.00% X XWeld 19 -5G 0.02% 0.05% 0.05% 0.05% X XWeld 20 -5H 0.01% 0.02% 0.03% X X XWeld 21 -5I 0.02% 0.08% X X X XWeld 22 -5J 0.00% 0.00% 0.00% 0.00% X XWeld 23 -6A 4.39% 16.78% X X X XWeld 24 -6B 68.19% X X X X XWeld 25 -7A 12.81% 40.72% 39.12% 27.98% 32.13% XWeld 26 -7B 5.49% 17.45% 16.76% 16.18% X XWeld 27 -7C 2.04% 6.49% 6.23% 6.02% X XWeld 28 -7D 0.23% 0.74% 0.71% 0.51% 0.59% XWeld 29 -7E 0.14% 0.45% 0.44% 0.31% 0.36% XWeld 30 -7F 0.08% 0.24% 0.23% 0.16% 0.19% XWeld 31 -7G 0.11% 0.33% 0.32% 0.31% X XWeld 32 -7H 0.06% 0.18% 0.18% 0.17% X XWeld 33 -71 0.01% 0.04% 0.05% X X XWeld 34 -7J 0.02% 0.07% 0.08% X X XWeld 35 -7K 0.03% 0.10% X X X XWeld 36 -7L 0.00% 0.00% X X X XWeld 37 -7M 0.00% 0.00% 0.00% 0.00% 0.00% XWeld 38 -7N 0.40% 1.27% 1.22% 0.88% 1.01% XWeld 39 -70 0.02% 0.07% 0.06% 0.05% 0.05% XWeld 40 -8A 0.09% 0.30% 0.34% X X XWeld 41 -8B 0.18% 0.56% 0.65% X X XWeld 42 -8C 0.19% 0.61% 0.70% X X XWeld 43 -8D 0.02% 0.08% 0.09% X X XWeld 44 -8E 0.07% 0.22% 0.26% X X XWeld 45 -8F 0.00% 0.01% 0.01% X X XSum 100.00% 100.00% 100.00% 100.00% 100.00% 100.00%21 NOC-AE-14003101Attachment 1Enclosure 1Table 17: Relative contributions by category when welds can experience o0ly a DEGB.22 NOC-AE-1 4003101Attachment 1Enclosure 2Enclosure 2 to Attachment 1Supporting resolution of APLAB, STP PRA Model -General: RAI 2"Determination of Adequacy of Plant-Specific PRA to Support Risk-InformedResolution of GSI-191, Rev. 1" FinalRevision 1 June 3, 2014NOC-AE-14003101Attachment 1Enclosure 2Determination of Adequacy of Plant-SpecificPRA to Support Risk-Informed Resolution ofGSI-191IntroductionThe Probabilistic Risk Assessment (PRA) supporting a risk-informed submittal is expected to conform tospecific technical adequacy requirements. Regulatory Guide 1.200' (RG 1.200) provides detailedguidance regarding these technical adequacy requirements by addressing the technical adequacy of thePRA.RG 1.200 addresses initiating events from two distinct hazard categories: internal hazards and externalhazards. Internal hazards involve internal events (e.g., turbine trip, loss of offsite power, etc.) andinternal floods. External hazards include seismic events, high winds, external floods and 'other' externalhazards. In RG 1.200, internal fires are addressed as an "external" hazard.RG 1.200 addresses requirements for both Level 1 (i.e., that portion of a risk scenario that starts with aninitiating event, addresses the response of the operators and equipment, and proceeds to thedetermination of either "successful termination" of the challenge or fuel damage) and Level 2 (i.e., thatportion of a risk scenario that addresses the progression of events following fuel damage leading to apotential release of radioactive material from the plant).Technical elements for the Level 1 internal events portion of the risk model are: initiating eventanalysis, success criteria analysis, accident sequence analysis, systems analysis, parameter estimationanalysis, human reliability analysis, and quantification. Technical elements for the Level 2 portion of therisk model are: plant damage state analysis, accident progression analysis, source term analysis andquantification. Documentation and uncertainty characterization are important components of thetechnical element requirements.RG 1.200 identifies the ASME/ANS PRA Standard' (the "Standard") as one acceptable approach todemonstrate technical adequacy. Appendix A3 of RG 1.200 provides a summary of the NRC position oneach high level and detailed supporting requirement found in the Standard. The Standard covers the1 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.200, "An Approach for Determining theTechnical Adequacy of Probabilistic Risk Analysis Results for Risk-Informed Activities," Revision 2, March 2009.2 ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessmentfor Nuclear Power Plant Applications," Addendum A to RA-S-2008, ASME, New York, NY, American Nuclear Society,La Grange Park, IL, February 2009.3 Appendix A is entitled "NRC Regulatory Position on ASME/ANS Standard."1 FinalRevision 1 June 3, 2014NOC-AE-14003101Attachment 1Enclosure 2same scope as does RG 1.200 (i.e., at-power, internal events and external events). Note that per RG1.200 and the Standard, a Peer Review is a required element of an acceptable PRA.The level of compliance with the detailed supporting requirements articulated in the Standard isdescribed in terms of three "capability categories." For a PRA to be considered technically adequate forrisk informed applications specific supporting requirements must meet capability category II or higher.The development of new editions to PRA standards continues as part of the processes for nationalconsensus standards. Examples of draft standards currently being balloted or under acceptance reviewinclude Low Power, Shutdown PRA, Non-LWR PRA, Advanced LWR PRA (new plants), Level 2 PRA andLevel 3 PRA. Additional standards are being considered (e.g., spent fuel pool PRA). If additional PRAstandards are balloted and accepted, it is likely that NRC will review them and potentially incorporatetheir underlying requirements -with possible refinements -into future revisions of RG 1.200. Thesenew standards are not part of the current revision of RG 1.200.RG 1.200 endorses the framework articulated in Regulatory Guide 1.1744 (RG 1.174). RG 1.174 providesguidance on the use of PRA in the support of risk-informed plant-specific changes in regulation. RG1.174 references the role of RG 1.200 in providing guidance on the adequacy of a plant-specific PRA tosupport regulatory decision making, including:"...demonstration that the baseline PRA (in total or specific parts) used in regulatoryapplications is of sufficient technical adequacy..." sRG 1.174 also included guidance on addressing issues potentially important to safety that are notspecifically addressed in the plant-specific PRA. These issues might include issues arising fromlimitations in the scope of the plant-specific PRA. For example, if the PRA does not include explicitconsideration of scenarios originating during low power or shutdown conditions, then a submittalfollowing the RG 1.174 guidance framework will have to provide either a justification of why suchscenarios are not applicable to the specific submittal request or a bounding impact on risk from suchscenarios. The relevant point is that RG 1.174 provides a path forward to address potential risk-informed plant-specific regulatory decisions even if the plant-specific PRA does not address all relevantrisk scenarios explicitly, when such scenarios are not within the scope of the current revision of RG1.200.In summary: the technical adequacy of a PRA is to be measured against the current revision RG 1.200,for those parts of the PRA that are relevant to the specific regulatbry decision. Relevant and applicable4 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, "An Approach for Using Probabilistic RiskAssessment in Risk-informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, May 2011.s RG 1.174, page 15.2 FinalRevision 1 June 3, 2014NOC-AE-14003101Attachment 1Enclosure 2technical supporting requirements identified in the Standard are expected to meet (at least) capabilitycategory II.Approach to Determine Technical Adequacy to Address GSI-191A five step process is outlined to determine and document the adequacy of the plant-specific PRA toprovide risk-informed information to support resolution of GSI-191. These steps are:1. Determination of the level of compliance of the PRA to the current version of RG 1.200.2. Identification of the PRA elements relevant to GSI-1913. Identification of relevant high level and supporting requirements from RG 1.2004. Comparison of relevant PRA elements to relevant RG 1.200 requirements5. Documentation of adequacy assessmentDetermination of the Status of Compliance of the Plant-Specific PRA(Step 1)The first step in the process of assessing the technical adequacy of the plant-specific PRA is to documentthe status of the PRA. Key questions include1. Has the PRA been peer reviewed?2. To the current revision of the Standard?3. Did the peer review process consider positions documented in Appendix A of RG 1.200?4. What was the conclusion of the peer review?5. Did the peer review result in any findings or observations? If so, what are they?If the PRA has been successfully reviewed against the Standard and RG 1.200, with no significantrelevant findings and observations from a peer review, then the technical adequacy evaluation processis complete.However, many plants have not yet completed upgrades to their PRAs to meet all the technicalrequirements for fire and seismic modeling. These plant-specific PRAs would most likely have been peerreviewed to an earlier version of the Standard and RG 1.200. (The primary difference between revision16 and revision 2 of RG 1.200 (as well as the primary difference between the 2002 version of the ASME6 Revision 1 of RG 1.200 references the 2002 version of the ASME Standard, as augmented by Addenda A and B(2003 and 2005, respectively). These versions address at-power internal events Level 1 and limited Level 2 PRA.They do not address fire or seismic scenarios.3 FinalRevision 1 June 3, 2014NOC-AE-14003101Attachment 1Enclosure 2Standard -as augmented in 2003 and 2005 -and the 2009 ASME/ANS Standard) is the inclusion of fireand seismic elements in revision 2.A plant-specific PRA that has been successfully peer reviewed with respect to an earlier version of RG1.200 can be used to support risk-informed regulatory decision making, as long as there are nosignificant differences in the technical requirements relevant to the submittal.STP Status: The STP plant-specific PRA was successfully peer reviewed against the requirementsof revision 1 of RG 1.200. This includes internal events and subsequent plant response forevents such as LOCAs, which is the initiating event of concern relative to GSI-191. While the STPPRA does include fire and seismic scenarios, those portions of the PRA are not compliant withrevision 2 of RG 1.200. As long as no fire or seismic induced scenarios contribute to GSI-191 fueldamage phenomena, the STP PRA is technically adequate to provide risk-informed informationon this issue. Allfindings and observations from the STP peer review against revision 1 of RG1.200 were addressed during the process of implementing (and receiving approvalfor) the STPRisk Managed Technical Specification program7.Identification of the PRA Elements Relevant to GSI-191 (Step 2)The PRA can be thought of as an organized set of scenarios. Each scenario begins with an initiatingevent that includes a representation of the response of the plant and operators to that initiating event.The identification of relevant PRA elements therefore begins with a consideration of the initiatorsincluded in the PRA.Selection criteria are established to characterize the PRA scenarios. The scenarios of interest in anevaluation of GSI-191 safety concerns must meet three criteria:1. The scenario response model for the initiator (i.e., LOCA) includes taking credit for emergencycore cooling system (ECCS) recirculation mode of operation to provide core cooling,2. The scenario involves the potential to liberate insulation installed around associated reactorcoolant system piping inside primary containment,3. And, the scenario includes a mechanism that transports the liberated piping insulation and otherpostulated debris to the emergency containment sump(s).A plant-specific evaluation is necessary in the evaluation of the PRA against these criteria. Theidentification of initiating events of interest is best illustrated by considering a specific example.7 STP Nuclear Operating Company, "STP Project Units 1 and 2, Docket Nos. STN 50-498, STN 50-499, Response toNRC Requests for Additional Information on STP Proposed Risk Managed Technical Specifications (TAC Nos. MD2341 & MD 2342)," NOC-AE-07002112, February 28, 2007.4 FinalRevision 1 June 3, 2014NOC-AE-14003101Attachment 1Enclosure 2STP Example:Many scenarios in the STP PRA take credit for ECCS sump recirculation. Specific initiators whoseresponse model includes recirculation are:1. RCP Seal LOCA2. Very Small LOCA3. Non-isolable Small LOCA4. Isolable Small LOCA5. Open safety relief valve (SRV) (one)6. Open SRV (two or more)7. Medium LOCA8. Large LOCA9. Steam Line Break Inside Containment10. Steam Line Break Outside Containment11. Other transient initiators including support system failure initiators12. Internal Plant Fires13. Seismic Events14. Internal Floods15. Other LOCAs Inside Containment (e.g., breaks in the RHR system)Initiator Groups 1 through 5 involve modest openings in the primary system. These events donot meet the necessary Criteria 2 and 3. Groups 1 through 4 result in only a modest amount ofinsulating material being liberated (with the amount associated with Group 1 being small).Small and very small LOCAs, by definition, do not result in containment spray initiation, so theylack a mechanism to transport material to the sump.Groups 5 and 6 involve the opening of SRVs. One SRV opening is equivalent to a small LOCA, sothe above argumentfor Groups 1 through 4 holds for this group also. In other words, thenecessary Criteria 2 and 3 are not satisfiedfor Group 5. In addition, the location of the SRVs issuch that a relatively small amount of target insulation is found near the SRVs. This wouldmean that criterion 2 is not met for either Group 5 or 6. Thus, initiator Groups 1 through 6 donot satisfy the criteria necessary to result in GSI-191 safety concern phenomena.5 FinalRevision 1 June 3, 2014NOC-AE-14003101Attachment 1Enclosure 2Initiator Groups 7 and 8 involve plant responses that potentially meet all three necessaryconditions. Thus, Groups 7 and 8 are therefore retained for further evaluation.Initiator Groups 9 through 13 include consideration of sump recirculation for those sequencesinvolving feed-and-bleed or a stuck open PORV whose flow is directed to the pressurizer relieftank (PRT), which eventually overpressurizes to the pressure limit of the engineered rupture disk.Engineering assessments indicate that little insulation material is found in the vicinity of the PRTrupture disk, so that a small amount of material would be made available to potentially betransported to the containment sumps (necessary condition 2). In addition, for Initiator Groups10 through 13 the safety injection pumps will continue to operate in the injection phase untilclosed loop RHR cooling is established and the containment sprays will not actuate so that notransport mechanism will be available to transport any liberated material to the sumps(necessary criterion 3). Initiator Groups 10 and 11 are screened from further evaluation at STP,as they also do not meet the necessary Criteria 2 and 3. Initiator Groups 10 and 11 do not resultin conditions necessary to result in GSI-191 safety concern phenomena8.For Initiator Group 9,containment spray is anticipated to actuate in response to the break. The plant response couldinclude sump recirculation, but only if heat removal via the steam generators is lost.Initiator Groups 12, 13, and 14 are associated with external events and generally have not beenshown to be significant contributors to initiating events that meet the required criteria.However, Internal fires and seismic events require additional consideration. As in the case ofGroups 9 through 11, feed-and-bleed and stuck open SRV scenarios do not meet the conditionsnecessary to result in GSI-191 phenomena. The necessary additional consideration for fires andseismic scenarios focuses on the question as to whether breaches in the primary system can becaused directly by the initiator.For seismic events, it is a common and generally considered conservative assumption that evenfor modest accelerations, one or more primary side pressure boundary instrument tubes mayfail resulting in the equivalent of a very small LOCA. This family of seismic induced LOCAscenarios is screened based on failure to meet the necessary Criteria 2 and 3. The robust natureof the primary system makes other seismically induced LOCAs requiring sump recirculation (i. e.,groups 7 and 8) very unlikely. In addition, while small, medium or large LOCAs are possible atsufficiently high seismic accelerations, the common assumption is that redundant componentsare fully correlated. Such LOCAs are of very smallfrequency. Under this assumption, forexample, a medium LOCA on one primary loop would be assumed to be accompanied bymedium LOCA on all other loops. The result is that seismically induced medium and large LOCAs8 In addition, at STP, the high head injection pump shutoff head is below the pressure necessary to open the PORV,reducing the likelihood of inducing a stuck open PORV.6 FinalRevision 1 June 3, 2014NOC-AE-14003101Attachment 1Enclosure 2are modeled as being excessive LOCAs-which have no success sequences by definition (i.e., theyare mapped directly to core damage).Fire induced LOCAs leading to opening of a pressurizer PORV or reactor vessel head vents are, inprinciple, possible. These scenarios are screened from further consideration because they do notmeet necessary Criteria 2 and 3.With these additional considerations, Initiator Groups 12 and 13 are screened from furtherconsideration. Initiator Groups 12 and 13 do not result in conditions necessary to result in GSI-191 safety concern phenomena at STP.Internal flooding (Group 14) represents a hazard group that, as far as GSI-191 phenomena areconcerned, is identical to other transients or support system failures that may degrade to feed-and-bleed (group 11). Internal flooding scenarios are screened from further considerationbecause they do not meet the necessary Criteria 2 and 3.Group 15 considers other LOCAs inside containment. At STP, the RHR system is wholly withincontainment, so that under the very unlikely conditions of an interfacing system pressurization,the RHR piping could become overpressurized. Although the RHR pumps and heat exchangersare inside containment these components are within their own respective cubicles and thereforethe amount of debris that could be liberated is highly likely to be confined within the cubiclearea itself and would not be readily transported to the emergency sumps. Additionally, the RHRsystem is designed with relief valve capacity and associated setpoints (600 psig) are sufficientthat limit pressure transients from events such as interfacing system LOCAs. Theseconsiderations result in LOCA likelihoods that are bounded by Initiator Groups already within thescope of the analysis.Conclusion: From a plant response analysis point of view, the initiators considered in the STPPRA were reviewed with respect to the conditions and criteria necessary to potentially result inGSI-191 safety concern phenomena. Only Medium and Large LOCAs survived this screeningprocess and are retained for further evaluation.Identifying Relevant Supporting Requirements from Regulatory Guide1.200 (Step 3)As indicated above, RG 1.200 and the Standard provide guidance to measure the adequacy of a PRA tobe used in regulatory decisions. The PRA scope addressed in RG 1.200 and the Standard covers Level 1and limited Level 2, at-power conditions, internal events and external events.7 FinalRevision 1 June 3, 2014NOC-AE-14003101Attachment 1Enclosure 2Step 3 identifies those portions of RG 1.200 and the Standard that are applicable for the applicationunder consideration. In step 3, the portions of RG 1.200 and the Standard pertaining to the relevantPRA elements, as identified in step 2, are identified.The process is necessarily plant-specific and is illustrated by an example.STP Example:In step 2, the initiating event categories relevant to consideration of GSI-191 phenomena havebeen shown to be limited to two specific internal events: medium and large LOCA.The step 3 process therefore begins with consideration of the high level and supportingrequirements for the accident sequence analysis of medium and large LOCAs. The assessmentthen turns to the assessment of high level and supporting requirements for the success criteriaanalysis, accident sequence analysis, systems analysis, parameter estimation analysis, humanreliability analysis, and quantification, as these elements support the PRA modeling of mediumand large LOCA. The evaluation continues for the limited Level 2 analysis, which containstechnical elements addressing the plant damage state analysis, accident progression analysis,source term analysis and quantification. Documentation and uncertainty characterization areimportant components of the technical element requirements. These are covered by the highlevel and supporting requirements of Parts 2 and 3 of the Standard. All of these high level andassociated supporting requirements have been met in the STP PRA.Comparison of Relevant PRA Elements to Relevant RG 1.200 Requirements(Step 4)In step 4, the relevant PRA elements are assessed for adequacy. This step essentially brings together theresults of steps 2 and 3, using those relevant elements of RG 1.200 identified in step 3 as a basis tomeasure the PRA elements for their technical adequacy to support regulatory decisions.To support regulatory decision making, each relevant technical supporting requirement must meet atleast capability category II. Failure to meet at least capability category II for any relevant element shouldbe documented as a finding. Any findings or observations should be addressed to determine theirpotential impact on supporting resolution of GSI-191.STP Example:The STP PRA has undergone a peer review under revision 1 of RG1.200. The relevant technicalelements of revision 1 (in Parts 2 and 3) as related and applicable to GSI-191 are substantiallythe same as in revision 2 of RG 1.200, so the peer review can be considered "current" withrespect to the PRA elements relevant to GSI-191 phenomena. This peer review thereforecompletes the requirements of step 4.Findings and Observations from that peer review have also been resolved.8 FinalRevision 1 June 3, 2014NOC-AE-14003101Attachment 1Enclosure 2Documentation of Adequacy Assessment (step 5)The process followed to determine the technical adequacy of the PRA to support GSI-191 adequacy,including resolution or treatment of any findings or observations, is to be documented.ConclusionA process to systematically determine the technical adequacy of a PRA to support resolution of GSI-191has been developed. The process is based on the framework of RG 1.200. The framework first identifiesthose elements of the PRA that are relevant to the issue. These elements are then compared to thecorresponding requirements of RG 1.200, revision 2. A PRA found to be technically adequate can thensupport a risk-informed submittal as described in RG 1.174.The process has been demonstrated using the STP PRA as an exampleSTP Example:The STP PRA has been shown to be technically adequate to support the resolution of GSI-191.RG 1.174 does impose additional risk information requirements. For example, we note that RG1.174 requires risk informed analyses to address all relevant operational modes. All modes withthe primary system pressurized potentially can result in the liberation of insulating material andthe need for sump recirculation. The STP PRA only addresses at power conditions, however, thePRA bounds the risk associated with pressurized conditions when not at power (i.e., not in Mode1 operations).9 NOC-AE-1 4003101Attachment 2Attachment 2Response to ESGB Request for Additional Informationa. Chemical Effects: RAI 3, 7,11, 17, 20, 22 NOC-AE-1 4003101Attachment 2Page 1 of 16ESGB, Steam Generator Tube Integrity and Chemical Engineering -Chemical Effects:RAI 3Please provide the technical basis for the 1 E-05 probability for the maximum chemicaleffect for each break size. The engineering judgment used to determine that valueappears to be arbitrary and other expert assessors could easily reach differentconclusions concerning a tail probability.STP Response:The exponential Probability Density Function (PDF) is a shifted, truncated, and single-parameter function that requires only the mean value to specify the entire continuousdistribution for all x > 1 , where x is the chemical head-loss factor. The maximumchemical effect factors for small, medium and large breaks -15.3 for small break LOCA(SBLOCA), 18.2 for medium break LOCA (MBLOCA), and 24 for large break LOCA(LBLOCA) were calculated as percentiles of their respective distributions that preserve atail probability of 1E-05. This means that only 1 in 100,000 random samples from thedistribution would be greater than the reported maximum. The maxima were included inevery Latin hypercube sample (LHS) replicate, and they were assigned a weight of 1E-05 to represent all chemical factors that might be higher.Assignment of a maximum chemical factor at 1E-05 is only arbitrary in the sense thatthis choice controls the probability weight carried by any failures induced by chemicalfactors sampled from the higher range of the distribution. The weight of 1E-05 waschosen to correspond to percentiles that ensure a quantifiable number of chemicallyinduced Emergency Core Cooling System (ECCS) failures. If no induced failures wereobserved, then the maxima would indeed be suspect, and the tail probability would needfurther reduction. With conventional debris head loss in the range of a few feet and astructural limit of only 9.35 ft., chemical factors exceeding 10 lead to failure. Therefore,the selected maxima would be considered conservative for the Loss of Coolant Accident(LOCA) spectra. Beyond the stated maxima, probability weights become vanishinglysmall, and sampling beyond the stated maxima would not induce any additional failures.
NOC-AE-1 4003101Attachment 2Page 2 of 16ESGB, Steam Generator Tube Integrity and Chemical Engineering -Chemical Effects:RAI 7CHLE Tank Tests 3 and 4 were performed with excessive quantities of aluminum relativeto the plant and with a temperature profile intended to induce chemical precipitation.These tests resulted in chemical precipitation and provided useful information related tohead loss loop response to chemical precipitates. The existing tests do not appear toaddress the extent of deviation from the best estimate plant conditions that could resultin chemical precipitation. One potential method to inform engineering judgment withrespect to chemical effects probabilities could be a series of smaller scale tests designedto evaluate the threshold concentrations of species that could result in precipitation. Forexample, tests could be designed to evaluate how much aluminum or calcium in solutionwould cause precipitation that may result in significant increases in head loss. Thesetypes of tests were included in the original chemical effects test plans but wereapparently cancelled. Please discuss any plans for smaller scale testing to investigatethreshold values for precipitation and whether that information would provide greaterconfidence in determining the probability that a post-LOCA plant condition would resultin chemical precipitate formation. If there are no plans for additional tests, pleaseprovide justification for this engineering judgment.STP Response:Chemical Head Loss Experiment (CHLE) Tank Tests 1 and 2 (the MBLOCA andLBLOCA tests) did not result in the formation of chemical precipitates. Based on thealuminum concentrations measured in solution, those results were consistent with theexisting equilibrium-based model for the prediction of the threshold concentrations ofspecies that could result in precipitation. Experiments conducted by Argonne NationalLaboratory (ANL) were identified that support the equilibrium-based model. Figure 1 is areprint from "Aluminum Solubility in Boron Containing Solutions as a Function of pH andTemperature" (ADAMS Accession # ML091610696) (1). This figure shows the results ofa large number of bench-top and vertical loop experiments and other literature data atvarious pH and temperature values, with separation of the data into a region whereprecipitation occurred and a region where precipitation did not occur. The authorspresented equations for empirical lines separating the two regions with the exception ofa few outliers, and a second set of equations for lines (shifted upward) that encompassall instances of precipitation.The authors of Ref. 1 subsequently published the data in Nuclear Engineering andDesign(2). The second publication included an additional line on the graph showing theprediction of the solubility of amorphous aluminum hydroxide based on an equilibrium-based model (Visual MINTEQ). Figure 2 is a reprint of the figure and demonstratesexcellent agreement between ANL's empirical boundary lines and the predictions ofVisual MINTEQ. Using the same approach as the original figure, an upward shift of theVisual MINTEQ line would encompass all instances of precipitation. The upward shift ofthe Visual MINTEQ line necessary to encompass the precipitation data is a 0.45-unitincrease in 'pH + p[AI]T'; in addition, this upward shift encompasses all precipitation datathroughout the entire temperature range with a single equation, whereas the empiricalboundaries in Ref. 1 included a separate equation for the data above 72 0C (175 'F).
NOC-AE-14003101Attachment 2Page 3 of 1612.51211.5-. 0 6061 Al TestAA11K10.5109.59No PPT, WCAP-16785 -* PPT. non-ftlocculated ANL Loop, No PPT.* PPT flocculated
- ANL Loop, PPT.ICET-1&5 £ ANL, STB Benchtop,60 80 100 120 140 160 180 200Temperature (F)220Figure 1 (Reprinted from Ref. 1): Al stability map in the 'pH+p[AI]T' vs. temperaturedomain for solutions containing boron. Filled and open symbols mean the occurrence ofAl hydroxide precipitation and no precipitation, respectively. 'pH' and 'p[AI]T' mean thesolution pH at temperature and the negative log to the base 10 of the total aluminumcontent as dissolved or precipitate in units of mollkg.
NOC-AE-14003101Attachment 2Page 4 of 16Temperature (OF)60 s0 100 120 140 160 180 200 22012512 * , 6061 AITest1100 AI Test11,5-- -" 0. 1051095 No PPT. WCAP-16785* PPT non-flocculated ANL Loop. No PPT.* PPT flocculated ANL Loop. PPT UICET.1&5 A ANL. STB Benclitop20 30 40 50 60 70 80 90 100Temperature (°C)Figure 2 (Reprinted from Ref. 2): Al hydroxide precipitation map in the 'pH+p[AIIT' vs. temperaturedomain based on ANL's bench top and loop test data and literature data.CHLE Tank Tests 3 and 4 were developed (as a replacement for the tests described inthe original test plan) to confirm the validity of the literature data and the existing VisualMINTEQ model for amorphous aluminum hydroxide solubility with the plant-specificchemistry at STP. Figure 3 presents the results of all 5 CHLE tank tests in the sameformat as the ANL data. The data demonstrate that CHLE Tank Tests 1, 2, and 5occurred in the non-precipitation region, and that the precipitation that resulted fromexcessive quantities of aluminum in CHLE Tank Tests 3 and 4 were consistent with theexisting data and model. An upward shift of 0.45 units of "pH + p[AI]T" encompasses theprecipitation data of Tests 3 and 4; this upward shift is similar to that necessary toencompass the precipitation region in the ANL data.Since CHLE Tank Tests 3 and 4 adequately confirmed the threshold concentrations atwhich precipitation would occur based on existing models and data, no additional testsare planned. The consistency between the results of CHLE Tank Tests and existingdata and model provides justification for the engineering judgment approach used foraluminum solubility in the license submittal. However, the engineering judgment wasbased on a direct application of the Visual MINTEQ amorphous aluminum hydroxidesolubility prediction without the shift of 0.45 units, since the license submittal occurredbefore CHLE Tank Tests 3 and 4 had been conducted. The shift of 0.45 units results ina small reduction of the concentration at which aluminum precipitates in the range of pH7.0 to 7.3 at 60 °C (140 OF).
NOC-AE-1 4003101Attachment 2Page 5 of 1613.012.512.011.50.<.11.0÷0.10.510.09.59.0TestT5 O0 <- Test T200 < -TestT1S0 , Test T30 o 0 aCT-est T4Prediction of Visual MINTEQPrediction of Visual MINTEQ shifted up by 0.45 units20 30 40 50 60Temperature (°C)70 80 90 100Figure 3: Al hydroxide precipitation map of the CHLE Tank Tests in the 'pH+p[AI]T' vs.temperature domain. Open symbols (gray) indicate results where precipitation was not observedand closed symbols (black) indicate where precipitation was observed. Each test is representedby an approximately horizontal series of data points (near-constant values of 'pH+p[AI]T' as thetemperature declined over the duration of the test).References:1. Bahn, C.B., Kasza, K.E., Shack, W.J., and Natesan, K. "Aluminum Solubility inBoron Containing Solutions and a Function of pH and Temperature. ADAMSAccession No. ML091610696, Argonne National Laboratory, September, 2008.2. Bahn, C.B., Kasza, K.E., Shack, W.J., Natesan, K, and Klein, P. "Evaluation ofprecipitates used in strainer head loss testing: Part Ill. Long-term aluminumhydroxide precipitation tests in borated water." Nuclear Engineering and Design,vol. 241, no. 5, pp. 1914-1925, 2011.
NOC-AE-1 4003101Attachment 2Page 6 of 16ESGB, Steam Generator Tube Integrity and Chemical Engineering -Chemical Effects:RAI 11aThe conclusions contained in document CHLE-014, "T2 LBLOCA Test Report," (letterdated October 13, 2013, available in ADAMS Accession No. ML13323A673) state, in part,"Chemical products did form under the simulated STP LBLOCA conditions but primarilywere adhered to the galvanized coupons." In addition, CHLE-020, "Test Results for a 10-day chemical effects test simulating LBLOCA conditions (T5)," states on page 10, "Thehigh turbidity at the beginning of Tests T5 and T2 shown in Figure 3b might be caused bydetachment of zinc particles from the zinc coupons and galvanized steel coupons due tothe high temperature during the first 80 minutes of the test." Page 75 of Volume 6.2states, "Although a zinc (Zn) product was observed to form under STP LOCA testconditions, it was not included in this analysis since the product was determined to becrystalline and mainly adhere to structures within containment as opposed to readilytravel with solution." Based on international experience, Framatome ANP, Inc. reporttitled, "Influence of Corrosion Processes on the Protected Sump Intake after CoolantLoss Accidents," December 2006 (ADAMS Accession No. ML083510156), zinc corrosionproduct dislodged by falling water caused a significant increase in head loss, pleasediscuss:(a) If following a LOCA, water either falling from a pipe break or from other locations inthe containment building could dislodge zinc corrosion product from galvanized steelsurfaces that could transport to the strainer.STP Response:It is possible for falling water from a pipe break or other locations to impinge on thegratings or other galvanized surfaces within the containment building. Direct jetimpingement from a break could cause damage to galvanized surfaces but would be fora relatively short duration and over a limited area. Continued impingement from theContainment Spray System (CSS) could occur for several additional hours until thesystem is secured. If galvanized surfaces are located below the break, falling watercould impinge on the surfaces even after the system is secured, although the surfacearea impinged in this manner would be limited in extent.Chemical Head Loss Experiment (CHLE) tank tests that contained galvanized steel orzinc surfaces experienced an initial peak in turbidity and zinc concentrations, asdescribed in the CHLE-020 document (1). Testing conducted for another licensee afterthe STP license submittal demonstrated that the zinc release was associated with theinitial period of low pH when Tri-sodium Phosphate (TSP) buffer was not present (2)When TSP was present and the pH was circumneutral, the initial release of zinc did notoccur. Thus, it is expected based on the current data that the period of time that zincwould be released from galvanized surfaces would be limited to the initial portion of theaccident sequence before the TSP has fully dissolved into the containment solution.
NOC-AE-1 4003101Attachment 2Page 7 of 16References:1. UNM, CHLE-020: Test Results for a 10-day chemical effects test simulatingLBLOCA conditions (T5), Rev. 3. University of New Mexico, Albuquerque, NM.Feb. 22, 2014, ML14072A079.2. UNM, CHLE-SNC-006: Bench Test Results for Series 2000 Tests for VogtleElectric Generating Plant, Rev. 1. University of New Mexico, Albuquerque, NM.Nov. 29 2013.
NOC-AE-1 4003101Attachment 2Page 8 of 16ESGB, Steam Generator Tube Integrity and Chemical Engineering -Chemical Effects:RAI 11b(b) Whether chemical effects contributions from zinc should be considered as part of theSTP chemical effects analysis.STP Response:Zinc from galvanized surfaces might be considered to contribute to head loss in twoways. First, zinc products dislodged from galvanized surfaces during the initial phasesof the Loss of Coolant Accident (LOCA) (before the TSP dissolves), as observed insome of the Chemical Head Loss Experiment (CHLE) tank tests, may contribute anadditional particulate source during the initial development of the debris bed. Thiscontribution was considered to be small (less than 10 percent, based on concentrationsmeasured in solution) compared to other sources of latent debris in the containmentbuilding and was not explicitly considered as a separate source of particulate. Thesecond source of zinc from galvanized surfaces is the slow formation of zinc phosphateon the galvanized surface due to reactions between the zinc in the galvanized coatingand the phosphate in the solution. Visual observations of the coupons in the CHLE tanktests indicated that this product formed slowly over a period of many days and remainedlargely adhered to the coupons. While quantitative rates of zinc phosphate formationwere not obtained from the CHLE tank tests, qualitative observations indicated that theproduct would not be present until later in the accident sequence when temperatureswere lower and strainer flow rates were lower, allowing additional margin for head lossthrough the strainer. Analysis indicated that the product was crystalline, which would beexpected to contribute to less head loss than amorphous corrosion products. Based onthe late formation of this material, its adherence to surfaces, and crystalline nature, thismaterial was considered less significant in the STP chemical effects analysis and itsformation was not explicitly considered in the analysis described in Volume 6.2. Thebump-up factor used to apply chemical effects head loss was not calculated on the basisof individual chemical products, and the potential for zinc phosphate to be a contributorto chemical head loss was implicitly considered during the development of the bump-upfactors.
NOC-AE-1 4003101Attachment 2Page 9 of 16ESGB, Steam Generator Tube Integrity and Chemical Engineering -Chemical Effects:RAI 17Page 187 of Volume 3 states "the chemical effects bump-up factor should never be lessthan one, and there is a practical maximum above which all events will lead to sumpfailure." Please discuss in more detail the approximate value of a bump up factor thatwill lead to sump failure. Please provide the values for conventional head loss that areassumed.STP Response:Every simulated break has its own time-dependent conventional head loss that iscalculated based on debris accumulation and flow rate, and added to a baseline clean-strainer head loss of 0.22 feet of water. Chemical factors are applied to the conventionalhead loss when the temperature is less than 140+/- 5 OF and the fiber load exceeds 1/16in. equivalent thickness. Total head loss is compared at every time step to theperformance metrics of (1) NPSHAvaii, (2) void fraction, and (3) mechanical buckling.For every break with a conventional head loss in the range of 1 ft. of water and amechanical loading limit of only 9.35 ft., a chemical head-loss factor exceeding 10 willinduce failures. A chemical head-loss factor of 43 would lead to buckling failure of thestrainer for all simulated breaks in Case 01, full train operation. A chemical head-lossfactor of 209 would lead to the violation of the NPSH margin criterion and failure for allsimulated breaks in Case 01, full train operation.These solutions were obtained by extracting the necessary data from the CASA GrandeCase 01 simulation. An example for the large-break population of chemical factorsneeded to induce mechanical buckling failure is shown in Figure A. The cumulativedistribution function (CDF) illustrates the percentage of LBLOCA cases that would fail forchemical factors < x .1.000.900.800.70CL 0.60E 0.50.0.400.30 .0.200.100.0.0. 1. 0 2 0 0 0 .0.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0Minium Bump-U~p Factor Required to Exceed Buckling limitFigure A. CDF for minimum chemical factor required to exceed strainer buckling limit forlarge breaks.
NOC-AE-14003101Attachment 2Page 10 of 16ESGB, Steam Generator Tube Integrity and Chemical Engineering -Chemical Effects:RAI 20Please discuss what benchmarking was performed with a) STP specific strainer tests andb) industry test data with similar conditions for the baseline head loss and chemicaleffects bump up factor.STP Response:Prototypical STP-specific strainer head loss tests were conducted at Alden ResearchLaboratory (ARL) in February (1 2) and July, 2008(3, 4),. The February head loss tests weresuperseded by the July head loss tests, because the February tests used walnut flour asa particulate surrogate. The reduced amount reflects most closely the amount of debrisfrom the majority of the large breaks. The CASA Grande head loss population for Case01 (all equipment starts and runs) was compared to all of the prototypical strainer headloss tests conducted at ARL. The maximum conventional CASA Grande head loss was8.2 ft, which bounds the maximum tested head losses for all of the ARL tests, exceptTest 3 in February. Test 3 was terminated after large head losses, greater than 15 ft,were observed following the addition of fine fibrous debris(2); as stated above, this testused walnut flour as a particulate surrogate and was superseded. The maximumpredicted chemical effects CASA Grande head loss was 154.9 ft, which bounds all theARL tests. The maximum predicted total CASA Grande head loss was 161.9 ft, whichbounds all the ARL tests.Also, expected values of exponential distributions applied for chemical head-loss factorswere chosen to be consistent with strainer test data showing chemical induced head-loss increases of approximately a factor of 2Vogtle Electric Generating Plant (Vogtle) conducted prototypical strainer head loss testsat the Alion hydraulics laboratory (5.Figure 1 displays the prototypical Vogtle and STPstrainer modules. Both modules are PCI Sure-Flow@ designs.rP (right) Prototypical NOC-AE-1 4003101Attachment 2Page 11 of 16All the STP strainer tests were conducted at an approach velocity of 0.0086 ft/s(2',4); theVogtle strainer tests were conducted at an approach velocity of 0.0150 ft/s(5). Debrisweights per prototypical strainer area for general debris types are displayed in Table 1.However, the specific insulation products tested under each debris-type category differbetween the two plants. For example, STP tested NUKON and Thermal Wrap underthe low-density fiberglass (LDFG) category, whereas Vogtle only tested NUKON underthe LDFG category.Table 1: Debris Comparison between STP and Vogtle(2'4, 5)LDFG LDFGFines Smalls Particulateper per per SodiumStrainer Strainer Strainer Calcium Aluminum AluminumArea, Area, Area, Phosphate, Oxyhydroxide, Silicate,Utility Test Ibm/ft2 Ibm/ft2 Ibm/ft2 Ibm/ft2 Ibm/ft2 Ibm/ft2STP Feb.2008 0.22 0.30 1.16 0.10 0.43 0Test 4STP Feb.2008 0.14 0.19 1.16 0.10 0.43 0Test 5STP July.2008 0.06 0.11 0.64 0.10 0.45 0Test 2Vogtle All Tests 0.31 0.14 6.60 0.09 0 0.14The maximum conventional, chemical effects, and total head losses predicted by CASAGrande for Case 01 also bound Vogtle's prototypical strainer head loss testing, whichmeasured 5.5 ft, 6.3 ft, and 11.8 ft, respectively(5).For the head loss comparisons cited above, the STP tests at ARL and Vogtle tests atAlion were not corrected to a common flow rate and temperature, which is conservative.The temperature range for these tests was 51'F to 11 7OF(1-5) and the temperature rangefor CASA Grande is 117'F to 255°F (LAR Enclosure 4-3, Table 2.2.13) correcting thetests to a higher temperature would reduce the head loss. The ARL tests modeled themaximum STP flow condition. Case 01 of CASA Grande was simulated at the maximumSTP flow condition, but containment spray pumps were secured during the LOCA.Therefore, the CASA Grande head loss population may inherently include head lossesat flow rates lower than the ARL test condition. The Vogtle tests were conducted at ahigher flow rate than the STP tests. Therefore, correcting the Vogtle tests to a lower flowrate would reduce the head loss. All factors considered, the benchmark comparisons ofmaximal computed head loss meet or exceed all applicable test data for STP andVogtle.
NOC-AE-1 4003101Attachment 2Page 12 of 16References:1. 0415-0100067WN / 0415-0200067WN. "South Texas Project Test Plan Feb2008". Revision A. 11/24/2008.2. 0415-0100069WN / 0415-0200069WN. "South Texas Project Test Report forECCS Strainer Performance Testing Feb 2008". Revision A. 11/24/2008.3. 0415-010007OWN / 0415-0200070WN. "South Texas Project Test Plan."Revision A. 8/14/2008.4. 0415-0100071WN / 0415-0200071WN. "South Texas Project Test Report forECCS Strainer Testing July 2008". Revision A. 11/24/2008.5. ALION-CAL-SNC-7410-005. "Head Loss Testing of a Prototypical Vogtle 1 and 2Strainer Assembly". Revision 0. 12/31/2009.
NOC-AE-1 4003101Attachment 2Page 13 of 16ESGB, Steam Generator Tube Integrity and Chemical Engineering -Chemical Effects:RAI 22aA total of five CHLE tank tests were performed to evaluate STP plant-specific chemicaleffects tests. CHLE Tests 1 and 2 were intended to evaluate an MBLOCA and anLBLOCA, respectively. Please address the following questions related to Tests 1 and 2:(a) Please discuss why the test screen debris bed is an acceptable method for detectionof chemical precipitates given the earlier test "CHLE-010, CHLE Tank Test Results forBlended and NEI [Nuclear Energy Institute] Fiber Beds with Aluminum Addition," thatshowed no head loss response even in the presence of large quantities of aluminumoxyhydroxide precipitate generated according to the WCAP-16530-NP-A protocol.Note: Additional details are available a September 6, 2012, meeting summary datedOctober 4, 2012 (ADAMS Accession No. ML12270A055).STP Response:The tests described in "CHLE-010, CHLE Tank Test Results for Blended and NEI FiberBeds with Aluminum Addition"1 did not contain aluminum oxyhydroxide precipitategenerated according to the WCAP-16530-NP-A protocol. In the CHLE-010 test series,aluminum oxyhydroxide precipitate was generated by injecting an aluminum nitratesolution directly into the tank recirculation line. The aluminum nitrate solution wasinjected in periodic batches at a slow rate corresponding to an increase in aluminumconcentration in the tank of 0.02 mg/L per minute. Precipitation occurred when thealuminum nitrate came in contact with the solution in the tank, which contained plant-specific concentrations of boric acid, lithium hydroxide, and tri-sodium phosphate (TSP)and was heated to about 45 0C (113 OF) at the time of aluminum nitrate injection. Incontrast, the WCAP-16530-NP-A protocol involves a more rapid addition of solidaluminum nitrate and sodium hydroxide into normal potable water in a mixing tank atambient temperature with a target aluminum oxyhydroxide concentration between 2,100and 11,000 mg/L. The CHLE-010 tests were designed to simulate the slow release ofaluminum during corrosion and the conditions for precipitate formation were substantiallydifferent from the WCAP-16530-NP-A protocol.Earlier tests, described in "CHLE-008, Debris Bed Preparation and Formation TestResults"'2 and provided to the NRC Staff in STP Letter NOC-AE-14003075, datedFebruary 27, 2014 (ML14072A076), did include aluminum oxyhydroxide precipitategenerated according to the WCAP-16530-NP-A protocol. In those tests, addition of theWCAP precipitates to columns with the same type of debris bed as the MBLOCA andLBLOCA tests (the NEI-prepared debris bed) did result in substantial head loss. InCHLE-008 Test 5, the addition of WCAP precipitates to an NEI-prepared debris bed atan approach velocity of 0.093 ft/s caused such a rapid increase of head loss that thestainless steel support screen collapsed before the entire batch of precipitates wasadded. In CHLE-008 Tests 6 and 8, addition of WCAP precipitate to the column resultedin substantial head loss through the NEI-prepared debris bed at an approach velocity of0.01 ft/s, which is comparable to that of the STP strainers. The quantity of aluminumoxyhydroxide precipitate that caused significant head loss corresponded to a screenloading of 246 g/m2.For comparison, the strainer testing done for STP at AldenResearch Laboratory by AREVA 3 had a final aluminum oxyhyroxide precipitate screenloading of 2,200 g/m2.Test 13 in CHLE-008 involved the addition of WCAP precipitatesto the CHLE tank and excessive head loss was detected in all three columns. The NOC-AE-1 4003101Attachment 2Page 14 of 16CHLE-008 test results demonstrate that the debris beds used in the MBLOCA andLBLOCA tests were capable of detecting aluminum oxyhydroxide precipitates generatedaccording to the WCAP-16530-NP-A protocol via a head loss measurement.The discussion between the NRC and STP during the September 2012 conferencemeeting mentioned in the RAI focused on the relative degree of sensitivity between theNEI-prepared debris beds and the blender-prepared debris beds. The CHLE-008 andCHLE-010 test results demonstrated that the threshold loading rate for the detection ofhead loss in blender-prepared debris beds was lower than in the NEI-prepared debrisbeds. Unfortunately, the blender-prepared debris beds experienced significant headloss when chemical precipitates were not present and exhibited other forms of instabilitysuch as a non-linear response to fluid velocity or quantity of fiber, making themunsuitable for detection of chemical precipitates in the CHLE tests. In addition, acomparison of the loading rates at which head loss first occurred in the NEI-prepareddebris beds in the vertical column and the mixed-debris bed in the AREVA strainertesting indicates that the threshold for detecting head loss is not as low in the NEI-prepared debris bed as it is in the mixed-debris bed. Thus, neither the NEI-prepared northe blender-prepared debris beds provided a sufficient method of detecting chemicalprecipitates via a head loss measurement.It is also important to note that the MBLOCA and LBLOCA tests employed multipleparameters to detect the presence of precipitates in addition to head loss through thedebris beds. Samples were periodically analyzed for total and dissolved aluminumperiodically during both tests. No significant difference between total and dissolvedaluminum was detected, indicating that all aluminum was in a dissolved form. Themeasured total concentrations were below the saturation concentration predicted foramorphous aluminum hydroxide by Visual MINTEQ, corroborating the evidence from thetotal and dissolved measurements. Turbidity remained low throughout the MBLOCA andLBLOCA tests. Results reported in CHLE-010 demonstrated a linear response betweenthe addition of aluminum and the turbidity of the solution, indicating that turbidity iscapable of detecting precipitates that form in this experimental system. Thus, even inthe absence of the column head loss data, the results from the MBLOCA and LBLOCAtests can be used to demonstrate that aluminum chemical precipitates do not form in thesimulated MBLOCA and LBLOCA environments.To account for the uncertainty associated with the head loss characteristics of variousdebris beds and the various ways of generating chemical precipitates, safety margin wasadded to the chemical effects contribution to head loss by applying a bump-up factor tothe calculated value of conventional head loss as described in the LAR Enclosure 4-1,page 9 and in more detail in LAR Enclosure 4-3, Section 5.6.3. The chemical head lossbump-up factor did not directly use head loss data from the CHLE tests. As a result, theability of the test screen debris bed to detect chemical precipitates has shown not toinfluence the results described in the STP license submittal.
NOC-AE-1 4003101Attachment 2Page 15 of 16References:1. University of New Mexico, CHLE-010: CHLE Tank Test Results for Blended andNEI Fiber Beds With Aluminum Addition, Rev. 3. Feb. 10, 2014. (ML14072A083)2. University of New Mexico, CHLE-008: Debris Bed Preparation and FormationTest Results, Rev. 4. Feb. 3, 2014. (ML14072A082)3. AREVA, South Texas-Project Test Report for ECCS Strainer Testing, Doc. 66-9088089-000.
NOC-AE-1 4003101Attachment 2Page 16 of 16ESGB, Steam Generator Tube Integrity and Chemical Engineering -Chemical Effects:RAI 22b(b) Please describe why the use of only aluminum and fiberglass in the MBLOCA testadequately represents the plant specific environment.STP Response:The objective of the CHLE testing program was to generate experimental data to supportan overall risk-informed approach to the resolution of GSI-191, while also conducting amanageable number of tests. Each test within the program had multiple objectives, withthe intent that the testing program as a whole provided data to support the resolution.Inclusion of all materials in all tests would not necessarily have provided the mostcomprehensive data, since in some cases the presence of one material might reduce thecontribution of chemical effects from another material. In the case of the MBLOCA andLBLOCA tests, the inclusion of zinc in the LBLOCA test and not in the MBLOCA test,coupled with comparisons of predicted release rates from the WCAP equations,demonstrated that the release of aluminum was greater when zinc was not present.That important outcome would not have been recognized if zinc had been included in alltests.While the tests included different materials and aspects of the LOCA to satisfy differentobjectives, the conditions within each test were representative of the plant-specificenvironment for the included materials. Many factors, including the quantities of boricacid, tri-sodium phosphate (TSP), and lithium hydroxide; the timing of TSP dissolution,acid generation, and spray duration; the temperature profile; and approach velocitythrough the screens were all representative of the plant-specific environment. For theMBLOCA test, the quantities of aluminum and fiberglass were also representative of theplant-specific environment during a MBLOCA.
NOC-AE-1 4003101Attachment 3Attachment 3Response to SCVB Request for Additional Information:RAI 1,2,3,4,5,6,7,8,9 NOC-AE-1 4003101Attachment 3Page 1 of 21SCVB, Containment and Ventilation Branch: RAI laIn support of Enclosure 2-3, "Request for Exemption from Certain Requirements ofGeneral Design Criterion [GDC] 38," please provide the following:(a) Please list the specific STP plant systems that will not meet the requirements of GDC-38.STP Response:Section 1 of Enclosure 2-3 of the license amendment request (LAR) identifies theContainment Spray System (CSS) as the only system for which the proposed exemptionto GDC-38 would apply. The CSS is the only system credited for meeting GDC-38 thattakes suction from the containment sumps during the recirculation mode of accidentmitigation and is consequently subject to debris effects, which are the focus of the risk-assessment provided in the STP licensing application.
NOC-AE-1 4003101Attachment 3Page 2 of 21SCVB, Containment and Ventilation Branch: RAI lb(b) Please describe the specific requirements of GDC-38 that will not be met by each ofthe plant systems listed in response to item (a) above.STP Response:The second paragraph of GDC-38 prescribes the deterministic approach to the analysisthat assures compliance with the criterion. The proposed exemption to GDC-38 utilizesthe application of a risk-informed approach per Regulatory Guide 1.174 for theassessment of the effects of debris. Debris effects could affect all three emergencysumps and consequently all three trains of the Containment Spray System (CSS). Inaccordance with the definition of single failure in 10CFR50, Appendix A, this would beconsidered a single occurrence that could result in the failure of multiple components.The risk-informed assessment demonstrates the change in core damage or large earlyrelease frequency from the debris effects is very low in accordance with the guidance ofRG 1.174 NOC-AE-1 4003101Attachment 3Page 3 of 21SCVB, Containment and Ventilation Branch: RAI 2aIn support of Enclosure 2-4, "Request for Exemption from Certain Requirements ofGeneral Design Criterion 41," please provide the following:(a) Please list the specific STP plant systems that will not meet the requirements of GDC-41.STP Response:Section 1 of Enclosure 2-4 of the license amendment request (LAR) identifies theContainment Spray System (CSS) as the only system for which the proposed exemptionto GDC-41 would apply. The CSS is the only system credited for meeting GDC-41 thattakes suction from the containment sumps during the recirculation mode of accidentmitigation and is consequently subject to debris effects, which are the focus of the risk-assessment provided in the STP licensing application.
NOC-AE-1 4003101Attachment 3Page 4 of 21SCVB, Containment and Ventilation Branch: RAI 2b(b) Please describe the specific requirements of GDC-41 that will not be met by each ofthe plant systems listed in response to item (a) above.STP Response:The second paragraph of GDC-41 prescribes the deterministic approach to the analysisthat assures compliance with the criterion. The proposed exemption to GDC-41 utilizesthe application of a risk-informed approach per Regulatory Guide 1.174 for theassessment of the effects of debris. Debris effects could affect all three emergencysumps and consequently all three trains of the Containment Spray System (CSS). Inaccordance with the definition of single failure in 10CFR50, Appendix A, this would beconsidered a single occurrence that could result in the failure of multiple components.The risk-informed assessment demonstrates the change in core damage or large earlyrelease frequency from the debris effects is very low in accordance with the guidance ofRG 1.174.
NOC-AE-1 4003101Attachment 3Page 5 of 21SCVB, Containment and Ventilation Branch: RAI 3aEnclosure 3, page 4, paragraph "Use of a Risk-Informed Approach to Resolving GSI-191",states:The design and licensing basis descriptions of accidents requiringECCS operation, Including analysis methods, assumptions, andresults provided in UFSAR [Updated Final Safety Analysis Report]Chapters 6 and 15 remain unchanged. This is based on thefunctionality of the ECCS and CSS during design basis accidentsbeing confirmed by demonstrating that the calculated riskassociated with GSI-191 for STP Units 1 and 2 is "Very Small" andless than the Region III acceptance guidelines defined by RG 1.174.The current licensing basis containment analysis methodology used to confirm theadequacy of the containment heat removal system (which complies with 10 CFR 50Appendix A GDC-38) described in the UFSAR is different from the proposedmethodology which resolves GSI-191 on a risk-informed basis and proposes anexemption request from GDC-38. For example: (a) difference in the single failureassumption in the proposed and current analysis; (b) computer codes RELAP for LOCAmass and energy (M&E) release, and MELCOR for the LOCA sump temperature responseare used in the proposed analysis, and SATAN-VI, WREFLOOD, FROTH are used for M&Erelease and CONTEMPT4/MOD5 is used for sump temperature response in the currentanalysis; and (c) the proposed analysis inputs and assumptions are required to beconservative from GSI-191 perspective and also required to be conservative for sumptemperature response whereas the current analysis inputs and assumptions areconservative for sump temperature response,(a) Please justify why the UFSAR licensing basis description of the methodology used forconfirming the adequacy of containment heat removal system which complies withGDC-38 should not be replaced with the proposed licensing basis methodologywhich takes an exemption from GDC-38.STP Response:The licensing basis for the assessment of the effects of debris is being revised and thedescription of the risk assessment will be described in the STP UFSAR as discussed inthe license amendment request, Enclosure 3, Attachment 2. The results of the risk-informed assessment demonstrate that the containment sumps are sufficiently reliable insupport of the Containment Spray System (CSS) such that the function of the CSS withrespect to containment analysis remains as currently described in the UFSAR.
NOC-AE-1 4003101Attachment 3Page 6 of 21SCVB, Containment and Ventilation Branch: RAI 3b(b) Tabulate the differences between the inputs and assumptions between the currentlicensing basis containment analysis that calculates the most limiting sump fluidtemperature profile for available NPSH calculation and the proposed containmentanalysis performed for risk-informed GSI-191. Please justify that the inputs andassumptions in the proposed methodology are conservative from both GSI-191 andsump temperature response perspectives.STP ResponseBackground and Reference to Submittal DocumentationThe proposed methodology is intended to be realistic, so it is not necessarilyconservative with respect to inputs and assumptions. The proposed methodology is notproposed to replace the conservatively-based methodology described in the UFSAR.More details are provided in the following paragraphs and tables.As described in the LAR, the proposed exemptions from General Design Criteria (GDC)-35, the "Emergency Core Cooling", GDC-38, "Containment Heat Removal", and GDC-41, "Containment Atmosphere Cleanup" are for approval of a risk-informed approach foraddressing GSI-191 and responding to Generic Letter (GL) 2004-02 for STP Units 1 and2 as the pilot plants for other licensees pursuing a similar approach. As furtherdescribed, STPNOC seeks NRC approval based on a determination that the risk-informed approach and the risk associated with the postulated failure mechanisms dueto GSI-191 concerns meets the guidance, key principles for risk-informed decision-making, and the acceptance guidelines in RG 1.174.STP is not proposing to apply the risk-informed approach to revise the licensing basis forcontainment design described in the UFSAR. The proposed risk assessment evaluates aspectrum of Loss of Coolant Accident (LOCA) scenarios to quantify the amount of debrisof various types that might be generated and transported to the emergency sumps, andhow that debris might affect available NPSH for Emergency Core Cooling System(ECCS) and Containment Spray System (CSS) pumps taking suction from the sumps inthe recirculation mode. It also evaluates potential transport of debris to the reactor core.It calculates failure probabilities that are fed to the STP PRA.Because the LAR uses a risk analysis, the engineering inputs in several areas aredifferent from the methods used in the existing deterministically-based Licensing Basis(LB) analyses. The containment analysis is an example of such an area of difference.The risk-informed approach to resolving GSI-191 applies the Probabilistic RiskAssessment (PRA) model to quantify the risk associated with GSI-191 concerns bycalculating the difference in risk for two cases:* The actual plant configuration for STP Units 1 and 2, with failures due to GSI-191concerns, and" The same plant configuration for STP Units 1 and 2, except for the assumptionthat there are no failures due to GSI-1 91 concerns.
NOC-AE-1 4003101Attachment 3Page 7 of 21Enclosure 1 to the LAR provides the generic methodology for the proposed risk-informedapproach to resolving GSI-191, consistent with RG 1.174 guidance. This enclosuredescribes the required inputs to the PRA model, the basic structure for appropriatelymodeling the inputs, and performance criteria used to calculate the risk. As described inEnclosure 1, the risk-informed approach to resolving GSI-191 uses the plant-specificPRA with realistic modeling to quantify the residual risk associated with GSI-191 and toevaluate for acceptable sump design in support of successful ECCS and CSS operationin recirculation mode following postulated LOCAs with the debris effects discussed inGSI-1 91.The Current LB Modeling ApproachThe licensing basis (LB) containment analysis is based on the CONTEMPT computercode and is documented in STP calculation NC07032. The CONTEMPT code isdocumented in two separate documents, NUREG/CR 3716 and NUREG/CR 4001.The current LB containment analysis uses 260 OF for the sump fluid temperature anddoes not address strainer failure due to the concerns raised in GSI-191. The analysis isdesigned to maximize containment pressure (and temperature), which would actuallyimprove net positive suction head available. The condition assumed in the LB analysis isvery unlikely to be realized in operation. Because the LB CONTEMPT methodology isnot intended to reflect realistic containment response behavior, and is based on anextremely unlikely scenario, the LB modeling approach differs from the approach usedfor compliance with RG 1.174 requirements.In particular, CONTEMPT4 has been verified to perform two major analyses [4, PageLOCA U-1 1]:* containment peak pressure and temperature analysis, and* containment environmental thermodynamic conditions for equipment qualificationand isolated pipe pressurization purposes.The Risk Informed Modeling ApproachThe containment analysis is based on the STP MELCOR model that runs simultaneouslywith the STP RELAP5 model designed for use in the STP PRA for the evaluationrequired in the risk-informed approach for addressing the GSI-1 91 issue. As described inLAR Enclosure 5, Item 5.a.13: In-Vessel Fiber Limits, several parameters related togeometry, thermal hydraulics/heat transfer, and engineered safety features used in theMELCOR input were taken from a previously certified Modular Accident AnalysisProgram (MAAP) STP containment model. Hence, the certification document for theMAAP model is appropriately referenced throughout the text.The following table lists the major qualitative differences by modeling subject areabetween the risk assessment and LB containment models. In the next section, numericalvalues are compared.
NOC-AE-1 4003101Attachment 3Page 8 of 21Subject RELAP/MELCOR CONTEMPT DifferenceSubcompartment Yes No STP CONTEMPT Model has 1 largeanalysis volume (1 pool, 1 atmosphere) and isnot validated for subcompartmentanalysis. MELCOR has several sub-compartments.Modeling goals Containment pressure Peak Pressure analysis Fundamentally different modelingresponse Sump Pool (structural design goals drove the modeling decisionstemperature testing, leak rate testing) for each code and explain some ofresponse and containment the differences in assumptions forthermodynamic each model.conditionsModel variations Single containment 2 separate models: one The single MELCOR model works inmodel regardless of for each transient stage concert with RELAPS-3D for allprimary side (injection, recirculation). transient stages. Primary sidecharacteristics, break Also, different characteristics are the exclusivesize, or stage of the assumptions for domain of RELAP5-3D during atransient (e.g. different steam coupled run. The CONTEMPT modelbefore/after sump generator types, primary may assume several different formsrecirculation) side characteristics, depending on transient stage,modeling goals, etc. primary side characteristics, etc.Modeling Best-estimate Conservative ESF delays, free volume calculations,philosophy other model characteristics are best-estimate for the MELCOR model butare generally conservative for theCONTEMPT modelCode execution Once-through, Can be an iterative There is no need for a collection orcoupled run from process consisting of succession of runs forstart of the transient initial runs, sensitivity MELCOR/RELAP. Under certainto conclusion runs, confirmatory runs, circumstances with CONTEMPT, theetc. user must perform several runs toascertain set-points, switchovertimes, etc.Code elements Control volumes, flow Control volumes, heat STP CONTEMPT model has no flowpaths, heat sinks, engineered safety paths since it is a single-volumestructures, features, control logic model. Both models have controlengineered safety volumes, heat sinks/structures, andfeatures, control logic engineered safety features. Details ofthe code elements differ.Engineered Safety Fan coolers, sprays Fan coolers, sprays Both codes model fan coolers andFeatures sprays with some correlation orphysics model (not identical ones).Actuation set-points are based ondiffering sets of assumptions. Again,best-estimate scenarios are used forMELCOR and conservative scenariosare used for CONTEMPT.
NOC-AE-1 4003101Attachment 3Page 9 of 21Subject RELAP/MELCOR CONTEMPT DifferenceContainment heat Neglect heat loss Include heat loss to Because CONTEMPT is not coupled inremoval through containment environment, account real problem time to another codewalls, no CSS or RHR for RHR heat exchangers that models the RHR heat exchangersheat exchanger and pumps in and details of LHSI/HHSI, thesemodeling (handled in CONTEMPT elements must be modeled. RELAP5-RELAP5-3D) 3D handles these aspects of thecalculation in the MELCOR/RELAP5-3Dcoupled run.Heat sinks/ MELCOR built-in Uchida and/or Tagami Condensation heat transfer is treatedcondensation correlations for both correlations used. Steel differently and heat sinks haveatmosphere and pool liners included on different characteristics. Liquid poolheat transfer containment walls with heat transfer is calculated internally bycoefficient air gap between liner MELCOR but assumed in thecalculations. Concrete and concrete. Constant CONTEMPT model.containment walls are heat transfer coefficientmodeled without the with pool.steel liner.Sump pool No decay heat added. Only 1 large pool for The large, lumped pool of CONTEMPTtreatment Mass and energy whole containment, vs. the smaller, annular sub-subtracted from the perhaps not intended to compartment'pool of MELCORpool based on capture the trueRELAP5-3D behavior of the sumpinstructions pool. Decay energyadded directly to pool in1 hour.Pipe break Communicated from Mass and energy STP CONTEMPT model uses mass andmass/energy source RELAP5-3D via release methodology energy release directly into thecoupling interface as described in WCAP Containment. RELAP/MELCOR usesproblem time 10325-P-A up to 3600 coupled RCS and Containmentprogresses. The response model.source is split byMELCOR into part recirculationliquid water, part methodology decaysteam, and part "fog" heat model from 3600seconds to 106 seconds.
NOC-AE-1 4003101Attachment 3Page 10 of 21Summary Comparison of Main Parameter ValuesThe main numerical input parameters controlling the initial conditions and boundaryconditions for timing and actuation, etc., between RELAP5/MELCOR and CONTEMPTanalyses are summarized in the following table.CONTEMPT Value [ ELAP5= ValueI 13D/MELCORSingle compartmentfree volume3.3E+6 ft3Sum over allcompartments3329332.0 ft3004.E4-U-.4-03'a-Initial containmenttemperatureInitial atmospheretemperature114 F119.93 FInitial containment 14.5 psia (max T) or 15.1pressure psia (max P)Initial relative humidity 20 %Initial RWSTtemperature 130 F12.0 psig (9.5 psig setpointSpray setpoint (HI-3) + 2.5 psig uncertainty)Spray actuation times Depends on time of HI-3.85.0 sFan cooler setpoint 5.5 psig (3.0 psig setpoint +(HI-i) 2.5 psig uncertainty)Fan cooler actuation45 stimesConcrete -0.8 BTU/hr-ft-FThermal conductivity Stainless Steel -9.4 BTU/hr-ft-FConcrete -0.208 BTU/Ibm-FSpecific heat capacity Stainless Steel -0.111BTU/Ibm-FInitial containmentpressure 14.94 psiaInitial relativehumidity, partialpressure of watervaporInitial RWSTtemperature 85 FSpray pressuresetpoint 9.5 psig15 s delay after setpoint,linear ramp to full flowFan cooler pressure 3.0 psigsetpoint (HI-i) 3.0_psigFan cooler actuationti oes a15 s delay after setpointtimesConcrete -0.54 BTU/ft-hr-FStainless Steel -f(T), variesConcrete -0.20 BTU/Ibm-FSpecific heat capacity Stainless Steel -f(T), variesDensityConcrete -144 Ibm/ft3Stainless Steel -488 Ibm/ft3DensityConcrete -144 Ibm/ft3Stainless Steel -495 Ibm/ft3 NOC-AE-1 4003101Attachment 3Page 11 of 21References:1. STP Calculation NC07032, "Containment LOCA Pressure/Temperature Analysis",STI 33686837.2. NUREG/CR 3716, "CONTEMPT4/MOD4 -A Multicompartment Containment SystemAnalysis Program"3. NUREG/CR 4001, "An Improvement to CONTEMPT/MOD4 MulticompartmentContainment System Analysis Program for Ice Containment Analysis4. STP Guide "LOCA P/T ANALYSIS USER GUIDE", Revision 4, ST1318704315. TAMU-GSI-002, "MELCOR Input Deck Certification: South Texas Project Large DryContainment, STI 336470846. WCAP 10325-P-A "Westinghouse LOCA Mass and Energy Release Data forContainment Design March 1979 Version" dated May 1983.
NOC-AE-1 4003101Attachment 3Page 12 of 21SCVB, Containment and Ventilation Branch: RAI 3c(c) In case the UFSAR licensing basis description of the containment heat removalsystem, including its related mass and energy release analysis methodology, isrequired to be replaced, please provide the revised UFSAR input for NRC staff reviewand approval.STP Response:STP proposes to supplement the existing UFSAR description with a description of therisk assessment of debris effects described in Enclosure 3, Attachment 2 of Reference 1to the cover letter.
NOC-AE-1 4003101Attachment 3Page 13 of 21SCVB, Containment and Ventilation Branch: RAI 4aThe current licensing basis methodology for the iodine removal is documented in UFSARSection 6.5.2, "Containment Spray System -Iodine Removal." The iodine removal isaccomplished by the CSS which meets the requirements of 10 CFR 50 Appendix A GDC-41. The proposed risk-informed GSI-191 methodology takes exemption from compliancewith GDC-41 requirements.(a) Please justify why the UFSAR licensing basis description of the iodine removalshould not be revised with the proposed methodology which takes exemption fromGDC-41.STP Response:The current licensing basis methodology for the iodine removal, as documented inUFSAR Section 6.5.2, is not being modified or replaced.The licensing basis for the assessment of the effects of debris is being revised and thedescription of the risk assessment will be described in the STP UFSAR as discussed inthe license amendment request, Enclosure 3, Attachment 2. The results of the risk-informed assessment demonstrate that the containment sumps are sufficiently reliable insupport of the Containment Spray System (CSS) such that the function of the CSSremains as currently described in the UFSAR with respect to dose assessment.
NOC-AE-1 4003101Attachment 3Page 14 of 21SCVB, Containment and Ventilation Branch: RAI 4b(b) Please tabulate the differences between the inputs and assumptions between thecurrent licensing basis containment atmosphere cleanup method and the proposedcontainment atmosphere cleanup which takes exemption from the GDC-41requirements.STP ResponseThe current licensing basis containment atmosphere cleanup method does notspecifically address the effects of debris on the Containment Spray System (CSS). Fromthe standpoint of GDC-41 and CSS, the parameter of interest is available NPSH in therecirculation mode. Other than the evaluation of the debris effects on CSS, the riskassessment does not evaluate containment atmosphere cleanup. The risk assessmentshows that the probability of debris affecting available NPSH for CSS such that the CSSwill not perform its function is very small in accordance with the RG 1.174 acceptancecriteria.
NOC-AE-1 4003101Attachment 3Page 15 of 21SCVB, Containment and Ventilation Branch: RAI 4c(c) In case the UFSAR licensing basis description of the iodine removal system isrequired to be replaced, please provide the revised UFSAR input for NRC staff reviewand approval.STP Response:STP proposes to supplement the existing UFSAR description with the risk assessmentof debris effects described in the license amendment request, Enclosure 3, Attachment2.
NOC-AE-1 4003101Attachment 3Page 16 of 21SCVB, Containment and Ventilation Branch: RAI 5In support of Volume 6.2, please list the differences between the heat sinks in the currentlicensing basis containment analysis documented in the UFSAR Tables 6.2.1.1-7 and6.2.1.1-8 and in the proposed containment analysis for risk-informed GSI-191. Pleaseprovide justification in cases where the conservatism is reduced in the proposedanalysis.STP ResponseAs discussed in the response to SCVB RAI 3.b, above, STP is not proposing to applythe RG 1.74 risk-informed approach to revise the licensing basis for containment designdescribed in the UFSAR. The containment pressures and temperatures calculated in therisk-informed analysis depend on the specific cases evaluated and are time-dependent;however, the values that correspond to the current UFSAR design basis conditions arecomparable to the current design and licensing basis results. The results of the analysisshow that the probability that debris will prevent the Emergency Core Cooling System(ECCS) and Containment Spray System (CSS) from performing their required function isvery small in accordance with the criteria of RG 1.174 and those systems are consideredable to perform their functions as described in the UFSAR. There is no change in theirdesign basis with respect to containment design.
NOC-AE-1 4003101Attachment 3Page 17 of 21SCVB, Containment and Ventilation Branch: RAI 6In support of Volume 6.2, please list the differences between the LOCA surface heattransfer model for heat sinks in the current licensing basis analysis documented inUFSAR Table 6.2.1.1-9 and the model in the proposed containment analysis for risk-informed GSI-191. Provide justification for the differences in case the conservatism isreduced in the proposed analysis.STP ResponseSee the response to SCVB-RAI 5, above.
NOC-AE-1 4003101Attachment 3Page 18 of 21SCVB, Containment and Ventilation Branch: RAI 7NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports forNuclear Power Plants: LWR Edition" (SRP), Section 6.2.1.5, "Minimum ContainmentPressure Analysis for Emergency Core Cooling System Performance Capability Studies,"describes the minimum containment pressure analysis for ECCS performance capability.RG 1.157, "Best Estimate Calculation for Emergency Core Cooling System Performance,"May 1989 (ADAMS Accession No. ML003739584), Section 3.12.1, "ContainmentPressure," provides guidance for calculating the containment pressure response usedfor evaluating cooling effectiveness during the post-blow-down phase of a LOCA.UFSAR Section 6.2.1.5 documents the current minimum containment pressure analysisfor performance capability studies of the ECCS. Please describe the proposedcontainment analysis, including assumptions and inputs, performed for the calculationof minimum containment pressure input for the ECCS analysis that calculates the peakcladding temperature for risk-informed GSI-191. Please justify that the inputs andassumptions are conservative for the purpose.STP ResponseThe risk assessment is limited to evaluating the effect of debris on Emergency CoreCooling System (ECCS) and Containment Spray System (CSS) in the recirculationmode. As discussed in the responses above, it is not proposed to replace the currentdesign basis containment analyses. It is not proposed as a change to the ECCSevaluation model and STP does not propose to apply it to show that 10CFR50.46(b)(1)limits are met for peak cladding temperature. The results of the analysis show that theprobability that debris will prevent the ECCS from performing its required function is verysmall in accordance with the criteria of RG 1.174 and the system is considered able toperform its function as described in the UFSAR. There is no change in the ECCS designbasis with respect to containment pressure.
NOC-AE-1 4003101Attachment 3Page 19 of 21SCVB, Containment and Ventilation Branch: RAI 8Volume 6.2, page 117, Item 5.a.14, "In-Vessel Thermal Hydraulic Analysis," lists sixscenarios simulated using the 3D Vessel-ID Core Model. Please describe and justify thebasis for selection of these LOCA breaks scenarios.STP ResponseEarly in development of the risk-informed approach for GSI-191 investigation, extremecore blockage scenarios investigations for core and Reactor Coolant System (RCS)response were unavailable in the academic literature. STP therefore undertook basic re-search to understand such core and RCS responses in theoretically extreme scenarios.The main idea behind these simulations was to investigate and understand, assumingthat flow blockage could occur, which extreme theoretical scenarios would go to successand which would lead to failure. Results of the studies performed have since beenpublished in peer-reviewed literature.Scenarios were developed assuming instantaneous blockage at the time of recirculationswitchover for hot and cold leg break locations and size (small, medium, and large) toaccount for different flow patterns and RCS response as described in the LAR. Thebreak sizes were chosen at high values for small, medium, and large STP LOCAcategories. All the cases assumed that one of the Emergency Core Cooling System(ECCS) trains is in the broken leg (STP Loop B) thereby minimizing effective injectionflow to the core. This matrix results in six scenarios.The primary objective was to study the core and RCS response for the main breaklocations and sizes to gain understanding of the severity of such responses underextreme conditions of blockage. Because flow from the sump would take a finite amountof time to carry any debris to the core, it is clear that such instantaneous blockage couldonly be realized in theory. Blockage, should it actually occur, would require someamount of time to build up. The scenarios selected therefore represent outcomes fortheoretical extremes. The results are useful for success criteria in the Probabilistic RiskAssessment (PRA) and for safety margin asked for in RG 1.174. That is, by investigatingthese extreme theoretical scenarios, STP could see that the vast majority ofhypothesized LOCA scenarios would go to success. STP investigated additionalmarginal cases as further described in the LAR Enclosure 5 (Volume 6.2, page 117, Item5.a.14), where either there was flow through the barrel-to-baffle bypass region (withoutcredit for LOCA holes in the baffles) or through a small opening (one fuel channel, eitherperipheral or center) in the core.
NOC-AE-1 4003101Attachment 3Page 20 of 21SCVB, Containment and Ventilation Branch: RAI 9aIn support of Enclosure 2-2, "Request for Exemption from Certain Requirements ofGeneral Design Criterion 35," please provide the following:(a) Please list the specific STP plant systems that will not meet the requirements of GDC-35.STP Response:The license amendment request, Enclosure 2-2, Section 1 identifies the EmergencyCore Cooling System (ECCS) as the only system for which the proposed exemption toGDC 35 would apply. The ECCS is the only system credited for meeting GDC-35 thattakes suction from the containment sumps during the recirculation mode of accidentmitigation and is consequently subject to debris effects, which are the focus of the riskassessment provided in the STP licensing application. The specific ECCS subsystemsthat are affected are Low Head Safety Injection and High Head Safety Injection.
NOC-AE-1 4003101Attachment 3Page 21 of 21SCVB, Containment and Ventilation Branch: RAI 9b(b) Please describe the specific requirements of GDC-35 that will not be met by each ofthe plant systems listed in response to item (a) above.STP Response:The second paragraph of GDC-35 prescribes the deterministic approach to the analysisthat assures compliance with the criterion. The proposed exemption to GDC-35 utilizesthe application of a risk-informed approach per Regulatory Guide 1.174 for theassessment of the effects of debris. Debris effects could affect all three emergencysumps and consequently all three trains of the Containment Spray System (CSS). Inaccordance with the definition of single failure in 10CFR50, Appendix A, this would beconsidered a single occurrence that could result in the failure of multiple components.The risk assessment demonstrates the change in core damage or large early releasefrequency from the debris effects is very low in accordance with the guidance of RG1.174.
NOC-AE-1 4003101Attachment 4Attachment 4Response to SNPB Request for Additional Information:RAI 1,2,3,5 NOC-AE-1 4003101Attachment 4Page 1 of 25SNPB, Nuclear Performance and Code Review Branch: RAI laPlease provide the following information for the STP Nuclear Steam Supply Systems(NSSSs):(a) Volume of the lower plenum, core and upper plenum below the bottom elevation ofthe hot leg, each identified separately. Also, please provide the heights of theseregions and the hot-leg diameter.STP Response:The volume of the lower plenum is 638.7 ft3.The volume of the core is 715.1 ft3.The volume of the upper plenum below the bottom elevation of the hot leg is 520.51 ft3Diagrams displaying these volumes and their respective elevations are shown below.a. P.ottom elevation of sucllon leg iciossover -31 -ID D0wg 14%6-01 4Cr2J-O2O42-( WN. typ- 8 pkl)b, Top eft.vation of cold !e.<. Top e MI or iheight 0 core)d. Rooettore1va1oiknof down~crowrHot W9g d~afflete?Top Elevation of Cold Leg b_---- --tRCS leg,C. UpINoplenvm rj suclioneg -1x~go hot Nog(42S BOorIo ElevTiono 5wCtiorn Leg a.33'4.75 C(OP03-ZG '.,owci 121&54. Sh 1CFj23.73' (Top of rcui (wri pipo! ID) -r22'523l25'O-m2I&814, Si, 91, Suction Icdgirneler WIDIU;%vtOfreplateTcpfuel NetsTop [IVwik~o ýf Care,2.11O7TI~weq 6117E0.0i1V51VSTPPDI IUI Cie. ,d4'nvedi)2'W(%[SP NDR U I C 18 -tlri-od)CoreREL13I D Model:V = 715:10ft.... 1 7. F 11) LOvg 61 11E69-Clt. AeI~vvV Fu vI (141I lip kim fIdRr -- -Lowi~enu --1 Core~ support topA , RLS J)I Bto elevation of dawncomrne d. -113model5359&5,45 / .LerPesm(.lV. V=638.7 t3--. __________ ktoi elevration of lower plen umo -12.7%'(STP11 NDR U I C 18 -dvrivd)b2. 75'457P NDRU I C 19 -dmived)50.81 (Dwg 6121 E87 -derived)4 Sal' D~vg ()121 ES 7 IJFSARSec. ,. &rivd)
NOC-AE-1 4003101Attachment 4Page 2 of 25Volume (Vup) of the Upper Plenum Below Hot Leg:0.6039791Vup = V845 + V865 x 3.625005Vup = 464.1557298 + 0.16661469 x 338.2303665Vup = 464.1557298 + 56.35415Vup= 520.51 ft3El Chng = 3.625005'El Chng = 3.6416876'Volume 865Hot leg inside diametertt RCS legs --- 32.25'(Dwg6-C-18-9-N-S007) --- 29"1D-------- -31.04167'El Chng = 0.6039791'T -Vup520.51 ft3Volume 845-Top Elevation of Core T 26.796'(STP NDR Ul C18 -derived)Active Fuel (168')Bottom of Active Fuel I 12.796'(STP NDR U1 Cl8 -derived)CoreRELAP5ID Model:605 & 606V=715.1 ftL JReference:1. T. Crook, A. Franklin, J. Scherr, R. Vaghetto, A. Vanni, and Y. Hassan, "South TexasProject Power Plant RETRAN-RELAP5-3D Conversion Tables", July 2013.
NOC-AE-1 4003101Attachment 4Page 3 of 25SNPB, Nuclear Performance and Code Review Branch: RAI lb(b) Loop friction and geometry pressure losses from the core exit through the steamgenerators to the inlet nozzle of the reactor vessel for steady state full poweroperation. Also, provide the locked rotor reactor coolant pump (RCP) k-factor. Pleaseprovide the mass flow rates, flow areas, k-factors, and coolant temperatures for thepressure losses provided (upper plenum, hot legs, Steam Generators (SGs), suctionlegs, RCPs, and discharge legs). Please include the reduced SG flow areas due toplugged tubes. Also, provide the loss from each of the intact cold legs through theannulus to a single broken cold leg and the equivalent loop resistance for the brokenloop and separately for the intact loop. Please identify the flow area (hydraulicdiameter) on which the k-factors are based.STP Response:Nuclear Steam Supply System (NSSS) parameters are provided here for steady statefull power operation under nominal conditions. To examine the pressure losses at eachsection of the loop, individual pressures were taken from the RELAP5-3D steady-statesimulation (1). Figure 1 shows the RELAP5-3D nodalization of the plant from Ref. 1. forreference in the following tables.Figure 1. RELAP5-3D Nodalization Diagram of the Primary System (Accumulator not Shown)
NOC-AE-1 4003101Attachment 4Page 4 of 25The steady-state plant operating conditions under which the pressure losses werecalculated are summarized in Table 1.Table 1. Steady-State Plant Operating ConditionsPS.ar.eter Steady Stat.e C(n01ti669qunits)Loop Mass Flow Rate 10108.453 (Ibm/s)Upper Plenum Bypass Fraction 2.080 (%)Core Flow Rate 36998.386 (lbm/S)Corefessel Inlet Temperature 560.977 OFAverage Core Outlet Temperature 626.354 OFReduced Flow Area Due to SG tube Plugging 0% (No SG plugging)Table 2 shows the total pressure drop through different sections of the primary system,calculated as difference of the total pressure in the nodes identified in column "RELAP5-3D Nodes ID."Table 2. RELAP5-3D Loop Pressure DifferentialsLocation RELAP5-3D Nodes ID RELAP5-36,Pressu're Change (psid) Pump Inlet 11202-11301 50.3095Pump Outlet 11301-11601 26.0005Vessel Inlet 11601-50101 0.6746Downcomer 53501-50101 4.2771Upper Plenum Bypass 58501-50101 -41.9059Core 84501-54501 -37.4984Vessel Outlet 10001-84501 -17.648Hot Leg 10402-10001 -0.4353SG Plenum Inlet 10601-10402 5.4342SG U-tubes Inlet 10801-10601 -5.3809SG U-Tubes 10808-10801 -15.4235SG U-Tubes Outlet 11001-10808 1.4743SG Plenum Outlet 11202-11001 -9.3558Vessel Inlet to Exit 50101-86501 38.8255Table 3 shows the flow areas, hydraulic diameters, and k-loss coefficients at differentlocations of the primary system, identified as junctions between nodes in column"RELAP5-3D Junction ID."
NOC-AE-1 4003101Attachment 4Page 5 of 25Table 3. RELAP5-3D Loop Flow Areas and Frictional k-loss Factors.* =.....:: ...... ... ' ... * .,, oatRELAP5-3D Flow Jno Forward ReverseLocation 2 HydraulicJunction ID Area (ft)kos k-lossDiameter (ft) ______,___Upper Plenum -Hot Leg jX21 (865 -XOO) 4.5869 2.41667 0.1194 0Hot Leg -SG Plenum Inlet jXO5 (X04 -X06) 4.5869 2.58583 0.46464 0.279639SG Plenum Inlet -U-tubes jX07 (X06 -X08) 15.2929 0.0506667 0.23 0.491709U-tubes -SG Plenum Outlet jX09 (X08 -X10) 15.2929 0.0506667 0.491709 0.23SG Plenum Outlet -Crossover Leg jXl 1 (Xl0 -X12) 5.241 2.58583 0.279639 0.46464Crossover Leg -RCP Inlet X12 -X13 5.241 2.583 0.001 2.09RCP Outlet -Cold Leg X13 -X14 4.1247 2.29167 2.09 0.001Cold Leg -Vessel Inlet jX19 (X18 -501) 4.1247 2.29167 0 0.1194Upper Core Plate -Upper Plenum 845 -865 51.0764 0.0365 0.5915 0.5915The locked rotor Reactor Coolant Pump (RCP) k-factor, calculatedsteady-state input model is shown below:with the RELAP5-3DKrotor = 5.86Reference:1. STP Power Plant RELAP5-3D Steady-State Model Verification. July 2013.
NOC-AE-1 4003101Attachment 4Page 6 of 25SNPB, Nuclear Performance and Code Review Branch: RAI Ic(c) Capacity and boron concentration of the RWST.STP Response:Refueling Water Storage Tank (RWST) (LAR Enclosure 4-3, Ref. 40, Page 6.3-32)Full tank volume, galMinimum volume (Technical Specification), galBoron concentration (as boric acid), ppm550,000*458,000*2,800-3,000*Volumes include unusable volume.
NOC-AE-1 4003101Attachment 4Page 7 of 25SNPB, Nuclear Performance and Code Review Branch: RAI ld(d) Capacity of the condensate storage tank (CST).STP Response:Unit 1 -Secondary Make-Up Tank, gal 300,000Unit 2 -Secondary Make-Up Tank, gal 300,000Unit 1 -Auxiliary Feedwater Storage Tank, TS min, gal 485,000Unit 2 -Auxiliary Feedwater Storage Tank, TS min, gal 485,000Secondary Make-Up Tanks provide normal, non-safety related make up water to thesecondary side.The Auxiliary Feedwater Storage Tank is the TS required safety-related water sourceused by the Auxiliary Feedwater System to remove heat from the RCS via the steamgenerators.
NOC-AE-1 4003101Attachment 4Page 8 of 25SNPB, Nuclear Performance and Code Review Branch: RAI le(e) Flushing flow rate at the time of switch to simultaneous injection.STP Response:Components of the flushing flow rate include the cold leg injection flow rate, hot leginjection flow rate, and vapor generation rate that were extracted from the RELAPS-3D/MELCOR simulation of a cold leg double ended guillotine (DEG) break scenario,under nominal operating conditions. Details on the simulation conditions applied areavailable in LAR Enclosure 4-3, Reference 5, page 12. The RELAP5-3D nodalizationdiagram adopted for the simulation is depicted in Figure 2 of LAR Enclosure 4-3,Reference 5, page 7.The total cold leg injection flow rate, the total hot leg injection flow rate, and the coreboil-off rate are plotted in Figure A. These thermal-hydraulic parameters have beenestimated as follows:" The total cold leg injection flow rate was estimated as the sum of the cold leginjection flow rate from each of the safety injection (Sl) trains (sum of mass flowrates of the valve components 149, 249 and 349 of Figure 2 of LAR Enclosure 4-3, Reference 5, page 7)." The total hot leg injection flow rate was estimated as the sum of the hot leginjection flow rate from each of the Sl trains (sum of mass flow rates of the valvecomponents 148, 248 and 348 of Figure 2 of LAR Enclosure 4-3, Reference 5,page 7)." The core boil-off rate was calculated as the sum of vapor generation rate in allthe nodes of the core (Pipe components 605 and 606 (LAR Enclosure 4-3,Reference 5, page 5), total 42 nodes). The vapor generation rate in each nodewas estimated by multiplying the value of the parameter "vapgen" by the volumeof the node.
NOC-AE-14003101Attachment 4Page 9 of 25200018001600-1400E 12001000U-FA 800InI(AS600400200-Cold Leg Injection (Ibm/s)-Hot Leg Injection (Ibm/s)-Vapor Generation (ibm/s)--- Switch to Simultaneous Injection0 1.,15000 17500 20000 22500 25000 27500Time (s)30000Figure A. Rate of Injection Parameters across the SimulationTable 1 summarizes the flow rates at the time to simultaneous injection (hot legswitchover time). The values reported are averaged over a period of time specified inthe table.Table 1. Average Injection and Vapor Generation Before and After Switch to Simultaneous InjectionI 15000s-Simultaneous Injection 1615.50 0.00 434.91After Simultaneous Injection 556.54 1107.42 0.75Reference:1. RELAP5-3D User's Manual, INEEL-EXT-98-00834.
NOC-AE-1 4003101Attachment 4Page 10 of 25SNPB, Nuclear Performance and Code Review Branch: RAI If(f) High pressure safety injection (HPSI} runout flow rate.STP Response:High Head Safety Injection Pumps (HHSI) (LAR Enclosure 4-3, Ref 40, Page 6.3-31)Max. (run-out) flow rate, gal/min 1,600 NOC-AE-1 4003101Attachment 4Page 11 of 25SNPB, Nuclear Performance and Code Review Branch: RAI 1g(g) Capacities and boron concentrations for high concentrate boric storage acid tanks, ifpart of system.STP Response:The boric acid storage tanks are not a part of the STP Emergency Core Cooling System(ECCS), (LAR Enclosure 4-3, Reference 40) and are not considered within the CASAGrande Analysis.
NOC-AE-1 4003101Attachment 4Page 12 of 25SNPB, Nuclear Performance and Code Review Branch: RAI lh(h) Flow rate into the RCS from the boric acid storage tanks, if applicable.STP Response:The boric acid storage tanks are not a part of the STP Emergency Core Cooling System(ECCS) (LAR Enclosure 4-3, Reference 40) and are not considered within the CASAGrande Analysis.
NOC-AE-1 4003101Attachment 4Page 13 of 25SNPB, Nuclear Performance and Code Review Branch: RAI 1i(i) Time to empty the RWST (all pumps operating).STP Response:The time to empty the Refueling Water Storage Tank (RWST) (time to initiate the sumpswitchover procedure) was calculated using the RELAP5 and MELCOR input modelsdescribed in (LAR Enclosure 4-3, Reference 5, page 5). The value reported below wascalculated under the following conditions:" Cold leg double ended guillotine (DEG) break (27.5 inch break in loop 3)" Nominal plant conditions (all Safety Injection (SI) and Containment Spray (CS)pumps operating) (See note)" Usable volume of the RWST (volume of the water until the low-low level alarm isreached) equal to 413,735 US gal.The time to empty the RWST was estimated to be:TRWST = 29.5 minNote: One of the three containment spray pumps manually secured at the beginning ofthe transient.Reference:1. STP MAAP 4.04 Input File NOC-AE-14003101Attachment 4Page 14 of 25SNPB, Nuclear Performance and Code Review Branch: RAI lj(j) Minimum containment pressure or containment pressure versus time graph.STP Response:Containment pressure was calculated using the MELCOR input model described in (LAREnclosure 4-3, Reference 5, page 8). The pressure reported was calculated under thefollowing conditions" Cold leg double ended guillotine (DEG) break (27.5 inch break in loop 3)" Nominal plant conditions (all Safety Injection (SI) and 2 Containment Spray (CS)pumps operating) (LAR Enclosure 4-3, Reference 5, page 14) (See Note)The pressure of the containment was extracted as the total pressure of the uppercompartment (node 4 of the MELCOR containment model nodalization diagram inFigure 4 of LAR Enclosure 4-3, Reference 5, page 9).The containment pressure response during the time between the sump switchover andthe hot leg switchover is plotted in Figure A.2019.51918.518 I17.5 ,-Containment Pressure---- Sump Switchover TimeIII--Hot Leg Switchover(U(00.0I..000I-0~IIII1716.5 I1615.5irIII17006700117001670021700Time [s]Figure A. Containment Pressure (from MELCOR Upper Compartment)Note: One of the three containment spray pumps manually secured at the beginning ofthe transient.
NOC-AE-1 4003101Attachment 4Page 15 of 25SNPB, Nuclear Performance and Code Review Branch: RAI 1k(k) Sump boric acid concentration versus time.STP Response:The sump boric acid concentration over time was not computed. However, the sumpboric acid concentration is tightly controlled. Approximately 88% of the sump boric acidconcentration is provided by the Refueling Water Storage Tank (RWST) (between 2800and 3000 ppm) and accumulators (between 2700 and 3000 ppm). The remainingcontribution to boric acid concentration comes from the RCS which varies between 0and 3500 ppm.Reference:1. South Texas Project Nuclear Operating Company. Westinghouse. CN-CRA-97-094 Rev. 1. Required Mass of TSP for LOCA Sump Solution pH Adjustment,November 2008 NOC-AE-1 4003101Attachment 4Page 16 of 25SNPB, Nuclear Performance and Code Review Branch: RAI 11(I) Minimum RWST temperature.STP Response:The minimum temperature for deterministic LOCA analysis is 50 0 F. This temperature isthe design range minimum for the building ambient temperature.Reference:1. South Texas Project Electric Generating Station. MAB HVAC Design BasisDocument 5V1 09VB001 10. Rev. 3. Table T-8. Page 266.
NOC-AE-14003101Attachment 4Page 17 of 25SNPB, Nuclear Performance and Code Review Branch: RAI 1m(m) Injection temperature versus time from sump during recirculation.STP Response:The High Head Safety Injection (HHSI) discharge temperature and the Low Head SafetyInjection (LHSI) discharge temperature were extracted from the simulation results of a27.5 inch cold leg double-ended guillotine (DEG) break performed with RELAP5-3D/MELCOR. Information on the RELAP5 and MELCOR simulation conditions arereported in the Sump Sensitivity Analysis (LAR Enclosure. 4-3, Reference. 5, page. 14)for nominal conditions.The RELAP5 nodalization diagram of the Emergency Core Cooling System (ECCS) isdepicted in Figure A to facilitate the identification of the volumes (nodes) where the liquidtemperature was read.Hot Leg0,SUx=2,3, and 4RWST IContainmentHHSI Pump sumpExchangerFigure A. RELAP5-3D Safety Injection Nodalization Diagram (X = loop number = 2, 3, or 4) (LAR Enc. 4-3, Ref. 5, Page 7)Based on the diagram of Figure A, the discharge temperature of the HHSI (time-dependent junction x45) is the same as the sump pool temperature (temperature of theliquid in the time-dependent volume x91) because no heat structures were modeledalong the HHSI flow path.
NOC-AE-14003101Attachment 4Page 18 of 25The discharge temperature of the LHSI (time-dependent junction x46) was read as theliquid temperature at the exit of the pipe component x47, simulating the primary side ofthe Residual Heat Removal (RHR) heat exchanger.The injection temperature, resulting from mixing of the liquid flows from the HHSI andLHSI, was read as the temperature of the liquid in the mixing branch (node x60).Figure B shows the discharge temperature of the HHSI, the discharge temperature ofthe LHSI, and the temperature of the mixed liquid injected in the primary system duringthe time between the sump switchover and the hot leg switchover.200 -LHSI190 Mix18SO -- HHSI170170 --- SumpSwitchoverTime-Hort Leg Switchover Time146 160150CL 140EO 1301201101001250 3750 6250 8750 11250 13750 16250 18750 21250Time (s)Figure B. Safety Injection Temperature (From Sump Switchover to Hot Leg Switchover)
NOC-AE-14003101Attachment 4Page 19 of 25SNPB, Nuclear Performance and Code Review Branch: RAI 2aPlease provide the following elevation data:(a) bottom elevation of the suction leg horizontal leg piping and cold leg diameterSTP Response:Inner Diameter 31",Center Line El. 22' 5-5/16"Bottom EL. of ID 21.151'Top EL. of ID 2 3.7 3'(See response to SNPB-RAI-2b for cold leg information).
NOC-AE-14003101Attachment 4Page 20 of 25SNPB, Nuclear Performance and Code Review Branch: RAI 2b(b) top elevation of the cold leg at the RCP dischargeSTP Response:Center Line El.Bottom EL. of ID,-Jr..31' 1.25"Top EL. of ID 33' 4.75" NOC-AE-1 4003101Attachment 4Page 21 of 25SNPB, Nuclear Performance and Code Review Branch: RAI 2c(c) top elevation of the core (also height of core)STP Response:Top Elevation of Core, ft 26.796'Height of Core, ft 14'(See the response to SNPB-RAI-la)
NOC-AE-1 4003101Attachment 4Page 22 of 25SNPB, Nuclear Performance and Code Review Branch: RAI 2d(d) bottom elevation of the downcomerSTP Response:Bottom Elevation of Downcomer, ft 10.81'(See the response to SNPB RAI la)
NOC-AE-1 4003101Attachment 4Page 23 of 25SNPB, Nuclear Performance and Code Review Branch: RAI 3Please provide the limiting bottom and top skewed axial power shapes.STP Response:The maximum core axial offsets (AOs) for LOCA analyses must meet the requirementsof the reload safety analysis checklist (RSAC). The RSAC requires a minimum AO of-20% and a maximum AO of 13% for LOCA. The cores are designed to meet TechnicalSpecification limits of Fxy and FnAHPower shapes are not a core design parameter, but are verified to be within the RSAClimits. Based on the design limits on AO, shapes (maximum allowed design LOCA AO)can be developed for bottom and top skewed axial power shapes using the followingconstraints:1 ) Power=0 at bottom and top of core (no power generation outside of core)2) The integral from 0 to 0.5 is 0.6 and the integral from 0.5 to 1.0 is 0.4 (bottom-skewed). The integral from 0.5 to 1.0 is 0.565 and the integral from 0 to 0.5 is 0.435(top-skewed).3) The slope is maximum (infinite) at x=0 (bottom-skewed), infinite at x = 1.0 (top-skewed)4) The slope at x=1.0 is negative (bottom-skewed), slope at 0 is positive (top-skewed)5) Uniform radial power distribution (fxy is everywhere 1.0), all channels have exactlythe same axial profile.The parameters above are chosen to maximize the power at the lowest (or highest fortop-skewed) by setting the slope maximum at x=0 or x=1.0. Also, uniform radial peakingwould minimize channel-to-channel mixing.The resulting function for bottom-skewed profiles is:f(x)= 4.05499*(x^O.574847)*(1 -x).The resulting function for top-skewed profiles is:f(x)= 4.66256*(x)*(1 -x)^0.717919A conceptual illustration of the functions is shown below.
NOC-AE-14003101Attachment 4Page 24 of 25fix) = 4.05499(x°374"7) (1 -x)f(x) = 4.66256 (x)(1 -x) 0.71791900.5Bottom-peaked power shape100:5 1Top-peaked power shapeThe actual minimum and maximum AOs measured for several core cycles has been wellwithin the core design requirements. STPNOC Guideline REM-2 "Core PerformanceTrending Program" requires measured axial offsets to be plotted for over the cycle forUnits 1 and 2. The minimum measured AO for was -7.5% in Unit 2 and the maximummeasured AO was 6% in Unit 1.
NOC-AE-1 4003101Attachment 4Page 25 of 25SNPB, Nuclear Performance and Code Review Branch: RAI 5Justification and description of the methodology used to compute the sump boric acidconcentration versus time.STP Response:The sump boric acid concentration over time was not computed. However, the sumpboric acid concentration is relatively constant as the concentration is tightly controlled.Approximately 88% of the sump boric acid concentration is provided by the RefuelingWater Storage Tank (RWST) (2800 and 3000 ppm) and accumulators (2700 and 3000ppm). The remaining contribution to boric acid concentration comes from the ReactorCoolant System (RCS) which varies between 0 and 3500 ppm. The RCS contribution tothe sump boric acid concentration has little effect.Reference:1. South Texas Project Nuclear Operating Company. Westinghouse. CN-CRA-97-094 Rev. 1. Required Mass of TSP for LOCA Sump Solution pH Adjustment.November 2008.
NOC-AE-1 4003101Attachment 5Attachment 5Response to SSIB Request for Additional Information:a.b.C.d.e.f.Transport: RAI 12Head Loss and Chemical Effects Bump Up: RAI 25, 26NPSH and Degasification: RAI 30, 31, 32, 34, 35In-Vessel and Boric Acid Precipitation: RAI 37Debris Bypass: RAI 39Defense in Depth and Mitigative Measures: RAI 41 NOC-AE-1 4003101Attachment 5Page 1 of 30SSIB, Safety Issue Resolution Branch-Transport: RAI 12Based on description of Item 5.a.4 of Volume 6.2 (Page 47), it is assumed that debris willremain in the vicinity in which it was washed down until recirculation starts. Pleaseprovide additional justification for this assumption. Please state if the debris would beredistributed during pool fill including by potential sheeting flow and if this would affectthe assumption that debris is mixed homogeneously in the pool at the start ofrecirculation. If so, please describe which types and sizes of debris are affected.STP Response:Item 5.a.4 of LAR Enclosure 5 (Vol. 6.2) was added as supplementary informationdescribing details and assumptions of the supporting debris transport calculation. Adescription of the homogenous pool mixing assumption used in the STP CASA Grandeevaluation is provided in the response to SSIB-RAI-1 1C in STP letter to NRC dated May22, 2014, NOC-AE-14003103 (ML14149A434).Because homogeneous mixing of all fine and small debris is assumed, the statement oflocal retention made in Vol. 6.2 has no impact on debris transport. Debris transportdirectly to the sumps under potential sheeting flow during pool fill is assessed as afraction of the debris that is calculated to reside on the floor immediately after the break.Similarly, the assumption of homogeneous mixing for all fine and small debris is notaffected by the debris location following pool fill.
NOC-AE-1 4003101Attachment 5Page 2 of 30SSIB, Safety Issue Resolution Branch-Head Loss and Chemical Effects Bump-up:RAI 25aVolume 3, Assumption 7.f, states that it is assumed that fiberglass will accumulateuniformly on the strainers but also states that the amount of debris that can collect onthe bottom of the strainer is limited to two inches. This assumption seems to contradictitself.(a) Please explain how the assumption is accounted for in head loss calculations orprovide information that shows it is not significant to the results. Please explain howa non-uniform accumulation of fibrous debris, limited by the floor or pool height,would affect the head loss calculation.STP Response:Assumption 7.f is accurate. There is a gap of 2 inches from the bottom of the strainer tothe floor. Initially, the strainer accumulates fiber uniformly, including the 2 inch gap.Once the loading transitions to the circumscribed area and the 2-inch gap is filled, (itphysically can no longer accumulate fiber), the remainder of the strainer continues toaccumulate fiber in a uniform manner. The existing STP strainer layout and designensures that the assumption of uniform accumulation of the transported debris over allactive portions of the strainer is valid.Equations 42 and 43 (LAR Enclosure 4-3, page 180) are used to calculate theincremental thickness increase and corresponding debris surface area. Once the 2-inchgap is filled, flow is considered to be restricted and Equations 42 and 43 calculateuniform debris accumulation on all sides of the strainer except the bottom where flow isset to zero. This decrease in flow surface area initially increases the head loss comparedto previous time steps when considering Equation 33 (LAR Enclosure 4-3, page 175).Assumption 7.f (LAR Enclosure 4-3, page 79) is used to calculate the incremental debrisbed thickness increase and debris (flow) areas used in the head loss calculationsaccording to Equations 40 through 43 (LAR Enclosure 4-3, page 180). Total strainervolumetric flow rate divided by available flow area determines approach velocity used inthe head-loss correlation.CASA Grande does not compare the debris height on the top of the strainer to the pooldepth. In the rare condition that the bed can exceed this height, the current evaluationallows flow area to increase unrealistically at the top of the strainer. Although unconfinedbed growth above the pool is non-conservative, it was stated in Assumption 7.f that thisis an unlikely occurrence. Assumption 7.f can be validated by comparing the minimumpool height (input distribution lower bound) and corresponding fiber accumulationnecessary to exceed the pool height to the probability distribution of debris volume (seeresponse to SSIB-RAI- 26d).
NOC-AE-1 4003101Attachment 5Page 3 of 30SSIB, Safety Issue Resolution Branch-Head Loss and Chemical Effects Bump-up:RAI 25b(b) Please provide an evaluation of how this affects Assumption 7.e. of Volume 3regarding homogeneous bed formation.STP Response:This does not affect Assumption 7.e. Assumptions 7.e and 7.f are consistent in thathomogeneously mixed debris will accumulate uniformly on the as-designed STPstrainer.Initially, the strainer accumulates the homogeneously mixed fiber uniformly, including the2-inch gap. Once the 2-inch gap is filled, (it physically can no longer accumulate fiber),the remainder of the strainer continues to accumulate fiber in a uniform manner, and theflow area is adjusted to account for loss of flow through the bottom. Because there is noflow simulated through the bottom of the strainer after the two-inch gap is filled, debriswill be homogenously mixed on all remaining sides of the strainer that do support flow.The existing STP strainer layout and design ensures that the assumption of uniformaccumulation of the transported debris over all active portions of the strainer is valid.
NOC-AE-1 4003101Attachment 5Page 4 of 30SSIB, Safety Issue Resolution Branch-Head Loss and Chemical Effects Bump-up:RAI 26aThe submittal calculates circumscribed bed surface areas based on debris loading(Reference: Volume 3, Section 5.6.2). Please provide the following information:(a) Please state if areas calculated for beds transitioned from thin bed to circumscribed.STP Response:Yes, the "strainer loading table" of Section 5.6 (LAR Enclosure 4-3, page 181, Table5.6.3) includes areas calculated for all debris loadings including the transition from thinbeds to circumscribed loads. This table was used to determine if the thin-bed loadingcriterion was exceeded. To evaluate surface areas of debris beds transitioning from thinbed to circumscribed, the debris was linearly interpolated using the arriving debrisvolume calculated using manufactured density. For thin beds, the interpolated debrisvolume range was between the 0 and 81.79 ft3 with a corresponding area range of1,818.5 and 419.0 ft2, respectively from Table 5.6.3 (LAR Enclosure 4-3, page 181).
NOC-AE-1 4003101Attachment 5Page 5 of 30SSIB, Safety Issue Resolution Branch-Head Loss and Chemical Effects Bump-up:RAI 26b(b) When fibrous debris is deposited on the strainer its density will be significantlyincreased from the manufactured value. Please state how was this accounted for(Volume 3, Page 696, Section 5.6.2).STP Response:The increase in density from the manufactured value is not accounted for in the "Strainerloading table" (LAR Enclosure 4-3, page 181). This table is based on thickness alone,and reports associated areas and volumes. Debris compression is accounted for by thehead-loss correlation that provides an effective thickness that can be used to interpolatethe table to find effective surface area.
NOC-AE-1 4003101Attachment 5Page 6 of 30SSIB, Safety Issue Resolution Branch-Head Loss and Chemical Effects Bump-up:RAI 26c(c) Please clarify if there are any objects around the strainer that would prevent thedebris bed from accumulating uniformly as assumed in the strainer loading(Volume 3, Table 5.6.3).STP Response:The general arrangement of the strainers is that one side of all 3 strainer trains faces theouter containment wall, and the bottom side faces the containment floor. One of thestrainer trains additionally has a structural wall approximately 2 to 3 feet away. Visualinspection of the Emergency Core Cooling System (ECCS) strainer performance testing(LAR Enclosure 4-3, Reference 53, Figure 8-6, Figure 8-7, page 47) shows that even inconfined spaces (test flume) the strainers load evenly.The floor represents an obstruction that prevents circumscribed loading from exceedingthe 2-in. gap that exists between the floor and the lowest edge of the strainers. Whenthis gap is filled, the lower surface is no longer available for flow. The strainer loadingtable then assumes that debris accumulation continues uniformly in all unimpededdirections.
NOC-AE-1 4003101Attachment 5Page 7 of 30SSIB, Safety Issue Resolution Branch-Head Loss and Chemical Effects Bump-up:RAI 26d(d) The NRC staff is of the opinion that it is not realistic to assume the thickness of thedebris bed on the strainer can be such that it will exceed the height of the water levelin the pool. Please explain how this affects the debris loading calculation (Volume 2,Section 5.6.2).STP Response:CASA Grande does not compare the debris height on top of the strainer to the pooldepth. In the rare condition that the bed can exceed this height, the currentimplementation allows flow area to increase unrealistically at the top of the strainer.Although unconfined bed growth above the pool is non-conservative, it was stated inAssumption 7.f that this is an unlikely occurrence (LAR Enclosure, 4-3, page 79).Assumption 7.f is validated by comparing the minimum pool height (input distributionlower bound) and corresponding fiber accumulation necessary to exceed the pool heightto the probability distribution of debris volume.A minimum containment pool volume of 39,533 ft3 (input distribution lower bound), andpool area of 12,301 ft2 were used in the CASA Grande evaluation (LAR Enclosure 4-3,page 45). Dividing the pool minimum volume by the pool area yields the minimumpossible sampled pool level of 3.2 feet or 38.6 inches. Subtracting the height of the topof the strainer 28.5 inches (LAR Enclosure 4-3, page 63) from the minimum pool levelgives the minimum thickness needed for the debris to reach the surface of the pool (10inches of fiber accumulation). This thickness equates to a volume of 328 ft3 necessary toaccumulate on the strainer to meet or surpass the minimum pool level when interpolatedfrom Table 5.6.3.Complementary cumulative density distributions of the total fiber amount (beforetransport fractions are applied) are illustrated in Figure A for many Latin HypercubeSampling (LHS) replicates, taken over the full break size range (SBLOCA, MBLOCA,and LBLOCA), using the Zone of Influence (ZOI) sizes described in Table 2.2.0 (LAREnclosure 4-3, page 56). This quantity includes latent fiber and all ZOI destroyed fibrousinsulation quantities. Figure B shows that on a closer scale the conditional probability ofexceeding 328 ft3of fibrous debris is less than 10-14.
NOC-AE-14003101Attachment 5Page 8 of 30Distribution of LDFG Volume Before Transport10010xIIAa)E> 10CDLL0101020,10 1 ........ 0 .10., 101 102 1 104Total LDFG Volume Generated -Before Transport (ift3)Figure A: CCDF of LDFG debris generated, before transport101~AS10,1Total LDFG Volume Generated -Before Transport (ft3)Figure B: CCDF of LDFG present and generated, before transport (zoomed in)
NOC-AE-14003101Attachment 5Page 9 of 30SSIB, Safety Issue Resolution Branch-Head Loss and Chemical Effects Bump-up:RAI 26e(e) Please state how often the debris loading algorithm results in a circumscribed bed orone that is transitioning to circumscribed (fully or partially filled interstitial volume).STP Response:The interstitial gaps of a single strainer are filled (0.5 inch debris thickness) when thevolume of debris meets or exceeds 81.79 ft3 (LAR Enclosure 4-3, Table 5.6.3).Complementary cumulative density distributions of the total fiber amount (beforetransport fractions are applied) are illustrated in Figure A for many Latin HypercubeSampling (LHS) replicates taken over the full break size range (SBLOCA, MBLOCA, andLBLOCA), using the Zone of Influence (ZOI) sizes described in Table 2.2.0 (LAREnclosure 4-3, page 56). This quantity includes latent fiber and all ZOI destroyed fibrousinsulation quantities. Figure B shows on a closer scale that the conditional probability ofexceeding 81.79 ft3 of fibrous debris is less than 1014.A data tick has been added to thefigure near the 81.79 ft3 debris volume. Note that Low Density Fiberglass (LDFG)volumes in these figures represent generated debris volumes. Corresponding debrisvalues that reach the strainer are smaller, so the associated probability of reaching thecircumscribed load is also smaller.Distribution of LDFG Volume Before Transport100AE00-J1o'ik1010k-10"101X: 81.06Y: 3.331e-15102 103Total LDFG Volume Generated -Before Transport (0t3)10'Figure A: CCDF of LDFG debris generated, before transport NOC-AE-14003101Attachment 5Page 10 of 30Distribution of LDFG Volume Before Transport10 ,10`-1x1&OIIA4)E10-120LL8.In9 14 81.0610- 124 1- 1101 102Total LDFG Volume Generated -Before Transport (f03)Figure B: CCDF of LDFG debris generated, before transport (zoomed in)
NOC-AE-1 4003101Attachment 5Page 11 of 30SSIB, Safety Issue Resolution Branch-Head Loss and Chemical Effects Bump-up:RAI 26f(f) Please explain the significance of cases that result in the interstitial volume of thestrainer becoming partially or completely filled with debris.STP Response:As the debris fills the interstitial volume of the strainers the flow area decreases from theclean strainer area to the limiting lowest flow area of the circumscribed facessurrounding the strainers. As the flow area decreases, the fluid approach velocity (U)increases (LAR Enclosure 4-3, Equation 34, page 176). This increase in fluid approachvelocity (U) directly increases the calculated head loss (LAR Enclosure 4-3, Equation 33,page 175).
NOC-AE-1 4003101Attachment 5Page 12 of 30SSIB, Safety Issue Resolution Branch -NPSH and Degasification: RAI 30The STPNOC submittal states that the degasification caused by the pressure dropthrough the debris bed is calculated to determine if a pump failure criterion is met(References: Volume 1, Section 1.1, "Structured Information Process Flow"; Volume 3,Assumptions 8 a. through i.; Volume 3, Section 5.7.2, "Degasification"; and Enclosure 6,Table 1). Please state if the degasification calculation credits containment accidentpressure. If so, please explain how the pressure for each case or condition is calculated.Please state what temperature is used for the degasification calculation and how thistemperature was calculated for each case.STP Response:No, the degasification calculation does not credit containment accident pressure.In all degasification calculations, bulk sump water temperatures below 2120 F assumeatmospheric pressure (14.7 psia). At sump temperatures above 2120 F, the containmentpressure is assumed to be equal to the vapor pressure of the sump water.The overall temperature range used for the degasification calculation(s) is 102.50 F to177.50 F for SBLOCA and MBLOCA. The overall temperature range for LBLOCAcalculations is 86.00 F to 255.80 F for LBLOCA .The range was determined by break sizebasis and is based on time-dependent changes in the bulk sump water temperature fromthe time of recirculation through steady-state long-term cooling (LAR Enclosure 4-3,Reference 5).
NOC-AE-1 4003101Attachment 5Page 13 of 30SSIB, Safety Issue Resolution Branch -NPSH and Degasification: RAI 31The STPNOC submittal does not seem to evaluate the possible effects of the collection ofgas bubbles in the strainer or ECCS pump suction piping (Reference: Volume 3,Assumption 8.h. and Section 5.7.3, "Gas Transport and Accumulation"). Please explainhow it was determined that gas bubbles would not collect in the strainer, or pipingbetween the strainer and ECCS and CSS pumps and eventually transport as large voids.If gas pockets can become trapped in these locations, please explain its effect.STP Response:Reference 56, TDI-6005-07, Vortex, Air Ingestion & Void Fraction South Texas ProjectUnits 1 & 2. Revision 3: November 24, 2008, evaluates the possibility of the collection ofgas bubbles in the STP strainer. While Reference 56 concludes that there is no airingestion or void formation, CASA Grande calculates a void fraction and applies theresults in the calculation Emergency Core Cooling System (ECCS) and ContainmentSpray System (CSS) pump NPSHr.As stated in Section 2.2.28, the acceptance criterion for a steady-state gas void fractionat the pump suction inlet is 2%. CASA Grande conservatively assumes that any voidformed at the sump strainer is fully transported to the ECCS or CSS pump suction,Assumption 8.i.The general transport of gas voids in the piping between the strainer and ECCS andCSS pumps is explained in Reference 58, VTD-G927-0001. Units 1 and 2 AcceptableGas Void Volumes in ECCS and RHR Suction Piping.
NOC-AE-1 4003101Attachment 5Page 14 of 30SSIB, Safety Issue Resolution Branch -NPSH and Degasification: RAI 32The NRC staff could not determine whether the calculation of NPSH Available (NPSHA)includes containment pressure greater than the saturation pressure of the sump fluid.Volume 3, Assumption 1.c indicates that containment pressure greater than thesaturation (above 14.7 pounds per square inch absolute (psia)) is not credited in theNPSH calculations (Reference: Volume 1, Section 1.1, "Structured Information ProcessFlow"; Volume 3, Sections 3, "Assumptions," and 5.7.2, "Degasification"; andEnclosure 6, Table 1). Please clarify if the calculation for NPSHA includes containmentpressure above the saturation pressure of the fluid. If containment pressure greater thanthe saturation pressure of the fluid is credited in the NPSHA calculation, please providejustification for its use and provide the methodology used to calculate the containmentpressure and sump fluid temperature for each case.STP Response:The NPSHA module of CASA Grande does not include containment pressure above thesaturation pressure, for coolant vapor pressure conditions greater than standardatmospheric pressure. For temperatures equivalent to or above boiling at standardatmospheric pressure, the containment pressure is set equal to the saturation pressureof the fluid for the NPSHA calculation. For containment coolant vapor pressures belowboiling, the standard atmospheric pressure of 14.7 psi is used as the containmentpressure.
NOC-AE-1 4003101Attachment 5Page 15 of 30SSIB, Safety Issue Resolution Branch -NPSH and Degasification: RAI 34The submittal lists minimum and maximum values for containment spray flow rates(Reference: Volume 3, Section 2.2.8, "ECCS and CCS Flow Rates"). Please state howthese values are used in the evaluation. If flow rates other than the maximum are used,please explain how the appropriate flow rate was determined for each case.STP Response:User entered minimum and maximum containment spray system (CSS) flow rates areapplied as probability distribution bounds; the small, medium, and large break scenarioshave the same bounds. For each simulated pipe break, the probability space betweenthe user entered minimum and maximum system flow rates are randomly sampled todetermine one individual CSS pump flow rate for the scenario. Total CSS flow rate isdetermined by multiplying the random pump flow rate by the number of operable CSSpumps.CSS flow rates used in the CASA Grande evaluation are entered as probabilitydistributions with equal probability between user entered CSS pump minimum andmaximum flow rates. The maximum value is set to the FLOMAP calculated averagedesign flows for train A and B operation during recirculation (LAR Enclosure 4-3Reference 42, page A-39).Values other than the maximum are appropriate because the bounding minimum flowrates, used as inputs, were selected from simulated probable events (LAR Enclosure 4-3, Reference 42, page A-40) for each operable-train state (i.e. 3, 2, or 1 train operable).CASA Grande uses the higher two-trains-operation flow rates (LAR Enclosure 4-3, Table2.2.15, page 54) for all events with two or three trains in operation (Cases 01, 09, 22 and26). Events with one train operation (Case 43) used their respective minimum andmaximum values from Table 2.2.15 (LAR Enclosure 4-3, page 54) to bound theirprobability distribution.
NOC-AE-1 4003101Attachment 5Page 16 of 30SSIB, Safety Issue Resolution Branch -NPSH and Degasification: RAI 35The STPNOC submittal calculates an equivalent break size of 38.9 inches for a27.5 inch-DEGB in Volume 3, Section 2.2.8. Please describe how the equivalent breaksize of 38.9 inches was calculated and why it was necessary to calculate this value.STP Response:The double ended guillotine break (DEGB) values computed by Equation 22 of LAREnclosure 4-3 (Section 5.3.1, Page 125) are used only to assign DEGB breaks to aLOCA category (S, M, L).For a double ended guillotine break, the break exit diameter (De) is equal to the innerdiameter (D,) of the pipe. A DEGB with full separation results in two jets (one from eachruptured side of the pipe). Thus, there are conceptually two break areas and two breakvolumes. The equivalent diameter (DDEGB) is determined by doubling the cross sectionalarea of the broken pipe and finding single equivalent diameter to represent the totalarea:A DEGB 2 -( i D24,4,DDEGBz 2' D i2)4 2 4 )Jon'Ten = V2 -D,Equation 22 NOC-AE-1 4003101Attachment 5Page 17 of 30SSIB, Safety Issue Resolution Branch -In-Vessel and Boric Acid Precipitation: RAI 37The STPNOC submittal uses 7.5 grams per fuel assembly as the fiber acceptance limit forcold-leg breaks (References: Volume 1, Section 1.1, "Structured Information ProcessFlow," Step 18; Volume 1, Sections 1.2.10, "Boric Acid Precipitation," and 1.2.11,"In-Vessel Fiber Limits"; Volume 3, Assumption 11.b; Volume 3, Section 4.2, "StructuredInformation Process Flow," Step 18; Volume 3, Section 5.11.2, "Acceptance Criteria:Debris Loads"; and Volume 6.2, Items 5.a.13 and 5.a.15). The NRC staff stated in its SEon "Evaluation of Long-Term Cooling Considering Particulate, Fibrous, andWCAP-16793, Revision 2, "Chemical Debris in the Recirculation Fluid," October 2011(ADAMS Accession No. ML13084A154), that the maximum amount of fiber that would bepresent in the limiting reactor design following a cold-leg break would be expected to beabout 7.5 grams, if the hot-leg break fiber amount did not exceed 15 grams. The staff didnot conclude that a fiber load of 7.5 grams was adequate to ensure that boric acidprecipitation would not occur. The amount was projected as the potential maximum inthe short term until industry completed a separate program on boric acid precipitation(BAP). In its evaluation, the staff considered that the plant calculation of the in-vesseldebris load included the worst case debris load for the plant and that most plants wouldhave much less than 7.5 grams of debris following a cold-leg break. Note that testing forthe WCAP did show that the flow required to match decay heat boil off would reach thecore following a cold-leg break with debris loads greater than 7.5 grams, but did notshow that mixing credited to prevent BAP would not be affected. The limit of 7.5 gramsper fuel assembly has not been technically justified as an acceptance criterion for BAP.Please provide the technical basis for assuming that 7.5 grams is an acceptable limit fora cold-leg break at STP when considering the potential for boric acid precipitation.STP Response:7.5 g/fuel assembly (FA) of fiber is chosen as a threshold of concern for boric acidprecipitation (BAP) based on previous results (LAR Enclosure 4-3, Reference [62]) thatshowed very little head loss when 15 g/FA with a full amount of chemical precipitateswere applied during hot leg break flow conditions. With a debris load of 7.5 g/FA, thecore is expected to remain full of water during a cold leg break (CLB) even though thereis no opportunity for bypass flow credited in the analysis. However, during a CLB, atthese low debris amounts (7.5 g/FA), STP would have significant flow through bypasspathways (LOCA holes and over the top of the core from barrel-to-baffle bypass region)as described in the LAR Enclosure 4-1, Section 2.1.2. LAR Enclosure 4-1, Section 2.1.2describes thermal hydraulic analysis of extreme scenarios that show the core wouldcontinue to be supplied with adequate flow such that cooling is preserved, and the corewould be reflooded early in the transient (ADAMS Accession No. ML14029A533).The 7.5 g/FA threshold is applied under the assumption of full debris deposition on thefuel and takes no credit for debris that may actually deposit in the barrel-to-bafflebypass. The CASA Grande analysis records a scenario failure whenever an equivalentinventory of 7.5 g/FA enters the core. STP fuel assemblies are designed with asignificant gap below the bottom tie plate that provides a large flow plenum between thebottom of the active fuel and the top of the bottom core plate. The flow-channel to barrel-to-baffle region has a large gap (approximately 2 inches) around the entire coreperiphery that is not credited for possible debris retention or for allowing lowconcentration flow to circulate though the bypass region through the Loss of CoolantAccident (LOCA) holes or over the top of the core.
NOC-AE-1 4003101Attachment 5Page 18 of 30STP has shown the chemical contribution to head loss is insignificant prior to hot legswitchover (about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />). Based on the STP reactor design, based on boundingexperimental results from both the work done in support of LAR Enclosure 4-3, page 82of 248, and from Reference (1), and based on bounding thermal hydraulic simulationresults for blockage, it is reasonable to use 7.5 g/FA, defined by tests using aconservative chemical load, as a threshold of concern for BAP. Note that this limit isapplied to all scenarios at all times regardless of realistic chemical loading and istherefore conservative even though higher thresholds could be considered, particularlyearly in the scenario. Although the core-fiber boron precipitation threshold could bereasonably described as a distribution having a mean much higher than 7.5 g/FA, theSTP LAR applied a sharp, single-value threshold to maintain clarity on this reactorperformance metric.References:1. CHLE-012 T1 MBLOCA Test Report Rev 4. Albuquerque, NM: University of NewMexico, February 18, 2014. (ML14072A084)2. CHLE-014 T2 LBLOCA Test Report Rev 3. Albuquerque, NM: University of NewMexico, February 22, 2014. (ML14072A085)
NOC-AE-1 4003101Attachment 5Page 19 of 30SSIB, Safety Issue Resolution Branch -Debris Bypass: RAI 39aThe submittal states that debris bypass or penetration testing was completed to supportmodeling of the bypass of debris past the STP strainer (Reference: Volume 6.2,Item 5.a.16). Please provide additional details on how debris penetration testing for fiberwas conducted. Specifically, please provide the following information:(a) Provide details on the characteristics of the fiber that was added to the test facility.i. How the fiber was prepared.ii. State What the percentages were of each fiber classification as described inNUREG/CR-6808, "Knowledge Base for the Effect of Debris on Pressurized WaterReactor Emergency Core Cooling Sump Performance," February 2003 (ADAMSAccession No. ML030780733), Table 3-2 after the fiber was prepared.iii. How was it ensured that agglomeration of the fiber did not occur prior to additionto the test loop?STP Response:i. The fiber was prepared with an ARL modified NEI protocol (LAR Enclosure 4-3,Reference 26, page 16). The general fiber preparation procedure was (LAREnclosure 4-3, Reference 26, Attachment A):1. Ensure drain valve in debris preparation tank is closed.2. Ensure the debris filter is installed on the return line from the tank.3. Weigh out the predetermined batch size of fiber and place into the scaled updebris preparation tank.4. Record quantity and debris lot information in Table 4.5. Fill debris preparation tank with water to at least the minimum dilution specified inthe NEI debris preparation protocol (< 0.72 Ibm/gal). For 2.4 Ibm of fiber, therecommended dilution based on the shakedown testing results in 30 gal.6. Start recirculating flow pump for the debris preparation tank.7. Ensure the pressure washer nozzles are connected.8. Change the height of the pressure washer nozzles to be within 2 in above orbelow the water surface.9. Start the pressure washer and run approximately 15 min (determined fromshakedown testing for 2.4 Ibm of fiber).
NOC-AE-1 4003101Attachment 5Page 20 of 3010. Examine debris characteristics for the batch and note the "comments" section onthe following page. If debris characteristics do not meet expectations, apply highpressure spray for an extended duration as needed, and document in commentssection.11. Stop the pressure washer and record the actual time over which the spray wasapplied in Table 4.12. Place a clean barrel under the debris preparation tank and drain debris throughthe valve in the bottom of the tank.13. Rinse tank walls and ensure filter is free of debris.14. Label the barrel containing the debris with the batch number and debris weight.Any deviations from this procedure were documented for each test in Attachment Aof the South Texas Penetration Test Report (LAR Enclosure 4-3, Reference 26). Forexample, Tests 1 through 4 interchanged steps 10 and 11 which were noted with "inkand initial" modifications during testing.ii. Analysis to quantitatively characterize the fiber after preparation and prior to testingwas not conducted. However, step 10 of the fiber preparation procedure, asdescribed above allows the fiber to be high pressure sprayed multiple times if theexpectation of the mostly class 2 fibers was not met.iii. Step 15 of fiber penetration test procedure requires "gently re-mixing the debris usinga mixing paddle" before the debris is introduced into the hopper/test flume (LAREnclosure 4-3, Reference 26, Attachment A).The debris hopper also introduces additional mixing energy to the debris before itenters the flume.
NOC-AE-1 4003101Attachment 5Page 21 of 30SSIB, Debris Bypass: RAI 39b(b) For tests that had more than one batch of fiber added to the test, please state whatthe timing was of each debris addition.STP Response:The subsequent batch of fiber was not added to the test until at least 5 pool turnoverswere completed (LAR Enclosure 4-3, Reference 26, page 26). Since the tests wereconducted at different flow rates, the duration of a pool turnover also varied. Thedurations of 5 pool turnovers for Tests 1 -5, Test 6, and Test 7 were 11.8 minutes, 50.5minutes, and 19.1 minutes, respectively (LAR Enclosure 4-3, Reference 26, AttachmentA).
NOC-AE-1 4003101Attachment 5Page 22 of 30SSIB, Debris Bypass: RAI 39c(c) Please describe the design of the test facility.STP Response:A schematic of the test configuration is below (LAR Enclosure 4-3, Reference 26, page21).A table of the dimensions critical to quality for Test 1 is shown on the subsequent page(LAR Enclosure 4-3, Reference 26, Attachment A).
NOC-AE-1 4003101Attachment 5Page 23 of 30Dimension Description Prescribed Assigned Reference Figure MeasurementDimension Descriptio Dimension ToleranceInflow Section Width 36" 1/2" Figure 2 35-7/8"Inflow Section Length 72" 1/2" Figure 2 71-3/4"Tapered Section Length 48" 1/2" Figure 2 48"Tapered Section Width 36" 1/2" Figure 2 35-7/8"(start)Tapered Section Width 21-13/16" 1/2" Figure 2 21-7/8"(end)Strainer Section Width 21-13/16" 1/2" Figure 2 21-7/8"Strainer Section 32" 1/2" Figure 2 31-3/4"LengthDistance from EasternStrainer Edge to 2" 1/2" Figure 2 2-1/8"Eastern Tank WallDistance from NorthernStrainer Disk to Northern 2-112" 5/8" Figure 2 2-3/4"Tank WallDistance from SouthernStrainer Disk to Southern 2-1/2" 5/8" Figure 2 2-1/4"Tank WallStrainer Submergence 9-1/2" 1" Figure 3 ---Total Strainer Height 30-1/2" 1/8" Figure 4 30-1/5"Strainer Disk Height 25" 1/8" Figure 4 25"Strainer Disk Width 28" 1/8" Figure 4 28"Total Strainer Width 31" 1/8" Figure 4 30-15/16"Disk Height off Ground 2-1/4" 1/8" Figure 4 2-3/16"+ 1/16"Active Module Length 16-13/16" -3/16" Figure 4 16-5/8" NOC-AE-1 4003101Attachment 5Page 24 of 30SSIB, Debris Bypass: RAI 39d(d) Was the circulation of fluid within the tank turbulent? Did debris settle? If somedebris did not reach the strainer, how was this accounted for?STP Response:Turbulence was introduced into the test tank via mixers. The mixers provided enoughturbulence to suspend debris in the test tank and prevent debris from settling to the floor(LAR Enclosure 4-3, Reference 26, page 16). Since the fiber remained suspended, alldebris reached the strainer.
NOC-AE-1 4003101Attachment 5Page 25 of 30SSIB, Debris Bypass: RAI 39e(e) How was it ensured that fiber did not bypass the filters during the test?STP Response:The test setup as described in response SSIB RAI 39c ensures full flow through the filterbags during the test. Additionally, the NUKON was prepared as fines having acharacteristic diameter of 7 microns (LAR Enclosure 4-3, Reference 44, Table 3-2). Thenominal pores of the bags used for fiber collection was (were) 5 microns. The largerdiameter fines, combined with the random orientation of the fiber as it contacted the filterbag, suggests that debris did not bypass the filters.
NOC-AE-1 4003101Attachment 5Page 26 of 30SSIB, Debris Bypass: RAI 39f(f) Was the design of the strainer and the design of the test facility (flow rate, etc.)prototypical with respect to the STP strainer?STP Response:Yes, the penetration test was prototypical with respect to the STP strainer. A STPprototypical PCI Sure-Flow strainer module was tested. The flow rate was scaled suchthat the maximum approach velocities of the tests and the STP strainer were equivalentat 0.0086 ft/s (LAR Enclosure 4-3, Reference 26, Table 3, page 19). Tests 1 through 4were tested with a total of 2.4 lb of fine fibrous debris; Tests 5 through 7 were tested witha total of 9.6 lb of fine fibrous debris (LAR Enclosure 4-3, Reference 26, Table 3, page19). The Design Basis Accident (DBA) test at ARL in July was conducted with a total of5.5 lb of fine fibrous debris (1).Reference:1. 0415-0100071WN / 0415-0200071WN. "South Texas Project Test Report forECCS Strainer Testing July 2008." Revision A. 11/24/2008.
NOC-AE-1 4003101Attachment 5Page 27 of 30SSIB, Safety Issue Resolution Branch -Defense in Depth and Mitigative Measures:RAI 41aVolume 1, Appendix C, Section C.5.4, lists mitigative measures that can be taken if thestrainer becomes blocked. It is not clear how the mitigative measures identified toaddress strainer blockage are implemented at STP (note that these actions are alsocredited for prevention of inadequate core flow). Please explain the following to explainhow the mitigative measures are capable of providing alternate flow to the requiredequipment.(a) The mitigative actions identified to reduce flow through the strainers appear toactually be designed to conserve RWST volume. These measures may delay theinitiation of recirculation, but except for securing CSS pumps will not reduce flowthrough the strainer. Please state at what point in the recovery these actions areperformed. If not performed immediately, will the RWST inventory be conserved? Ifthe reductions of flow through the strainer do not occur until after strainer blockageis evident, please state if these actions are effective.STP Response:A reduction in flow will occur (one (1) Containment Spray System (CSS) pump secured)before switchover to recirculation as directed by the conditional information page (CIP) inSTP procedure OPOP05-EO-EOOO, "Reactor Trip or Safety Injection" (LAR Enclosure 4-3, Reference 32). Securing a single CSS pump will conserve Refueling Water StorageTank (RWST) inventory. Per EOP OPOP05-EO-EO10, "Loss of Reactor or SecondaryCoolant" all sprays can be secured after 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> based on Iodine levels low enough tosupport habitability but containment pressure would need to be less than 6.5 psig andTSC concurrence. Conservation of RWST volume and reduction of strainer flow are bothbeneficial strategies for LOCA response that can be achieved by securing spray pumps.The third CSS pump is stopped shortly after the LOCA occurs, before the RWST isempty. As a consequence, when recirculation starts, that train will have approximately40% less flow through the strainer resulting in much less debris accumulation andtherefore head losses (on that train's strainer).Stopping a CSS pump in the most likely plant state scenario (all CSS trains running) isintended to conserve RWST inventory. Additionally, the flow through the EmergencyCore Cooling System (ECCS) strainer will be reduced by approximately 40% in the trainthat the CSS pump is stopped. This reduction in total flow through the strainer has theadditional benefit of reducing debris buildup on that specific strainer. A potentiallyadverse effect of securing spray flow is that the lower debris bed inventory allows morefiber penetration to the core.In the risk-informed methodology, the effect of the reduction in flow is taken into accountas described in the LAR Enclosure 4-2, Section A.4.2, page 83 of 257 in the descriptionof Top Event OFFS. The strainer loading is accounted for as found in LAR Enclosure 4-3, Equations 87 through 93 page 210 of 248 and Section 3 page 78 of 248 (6e: "It wasassumed that the debris transport to each of the strainers is proportional to the flow ratethrough each strainer divided by the total flow rate through all of the strainers. This is areasonable assumption since the debris transports with the flow.")
NOC-AE-1 4003101Attachment 5Page 28 of 30SSIB, Defense In Depth and Mitigative Measures: RAI 41b(b) Please state if STP has implemented operating procedures to secure the third train ofECCS/CSS if all three are initiated following a LOCA.STP Response:STPNOC did not revise the Emergency Operating Procedure (EOP) Emergency CoreCooling System-(ECCS) termination criteria to secure any trains of safety injection. TheEOPs were modified to secure one train of Containment Spray System (CSS) if all threetrains are injecting.
NOC-AE-1 4003101Attachment 5Page 29 of 30SSIB, Defense In Depth and Mitigative Measures: RAI 41c(c) Please clarify if STP implemented operating procedures or other guidance tobackwash the strainers, if necessary. If so, please provide details on the proceduralcontrols for this action.STP Response:No. STPNOC does not have a procedure or guidance to allow backwashing of theECCS strainers.
NOC-AE-1 4003101Attachment 5Page 30 of 30SSIB, Defense In Depth and Mitigative Measures: RAI 41d(d) Please state when the RWST refill is started and how long it takes to refill RWST tothe point where injection from the tank is viable. Please note that if the tank is notready for injection when blockage occurs, this action may not be effective. The NRCstaff notes that the STPNOC submittal states that most strainer blockage eventsoccur within the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the LOCA recovery.STP Response:As stated in the LAR Enclosure 4-1, page C14 the STP Emergency OperatingProcedures (EOPs) contain steps to refill the Refueling Water Storage Tank (RWST).STP procedure OPOP05-EO-EO10 "Loss of Reactor or Secondary Coolanf' directs theoperator to enter procedure OPOP05-EO-ES13 "Transfer to Cold Leg Recirculation."OPOP05-EO-ES13 directs the operators to refill the RWST in the step followingcompletion of transfer to cold leg recirculation. Following a hypothesized large breakLOCA, this would occur about 20 to 25 minutes following the start of the event.The makeup flow rate to RWST is approximately 150 gpm. This equates toapproximately 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> to make up 100,000 gallons. The minimum volume allowed inthe RWST before securing the ECCS/CSS pumps is 32,500 gallons by the procedure.In the unlikely event that debris prevents recirculation in all Emergency Core CoolingSystem (ECCS) trains in 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />, the RWST would be sufficiently full of water. The refillrate (plus RWST drain down) would be able to meet core cooling requirements withoutinterruption over a long period of time as shown in the LAR Enclosure 4-3, Table 5.10.1page 224 of 248. Thus the actions to re-fill in the RWST will be effective for providing analternative method of cooling water supply.
NOC-AE-1 4003101Attachment 6Attachment 6Definitions and Acronyms NOC-AE-1 4003105Attachment 6Page 1 of 2Definitions and AcronymsARLBABAPBCBEPB-FB-JBWRCADCASACCDFCCWCDFCETCHLECHRSCLBCRMPCSCSHLCSSCVCSDBADBDD&CDEGBDIDDMECCECCSECWSEOFAlden Research LaboratoryBoric AcidBoric Acid PrecipitationBranch ConnectionBest Efficiency PointBimetallic WeldsSingle Metal WeldsBoiling Water ReactorComputer Aided DesignContainment AccidentStochastic AnalysisComplementary CumulativeDistribution Function orConditional Core DamageFrequencyComponent Cooling WaterCore Damage FrequencyCore Exit Thermocouple(s)Corrosion/Head LossExperimentsContainment Heat RemovalSystemCold Leg Break or CurrentLicensing BasisConfiguration RiskManagement ProgramContainment SprayClean Strainer Head LossContainment Spray System(same as CS)Chemical Volume ControlSystemDesign Basis AccidentDesign Basis DocumentDesign and ConstructionDefectsDouble Ended GuillotineBreakDefense in DepthDegradation MechanismEmergency Core Cooling(same as ECCS)Emergency Core CoolingSystemEssential Cooling WaterSystem (also ECW)Emergency OperationsFacilityEOP Emergency OperatingProcedure(s)EPRI Electric Power ResearchInstituteESF Engineered Safety FeatureFA Fuel Assembly(s)FHB Fuel Handling BuildingGDC General Design Criterion(ia)GL Generic LetterGSI Generic Safety IssueHHSI High Head Safety Injection(ECCS Subsystem)HLB Hot Leg BreakHLSO Hot Leg SwitchoverID Inside DiameterIGSCC Intergranular StressCorrosion CrackingISI In-Service InspectionLAR License AmendmentRequestLBB Leak Before BreakLBLOCA Large Break Loss of CoolantAccidentLDFG Low Density FiberglassLERF Large Early ReleaseFrequencyLHS Latin Hypercube SamplingLHSI Low Head Safety Injection(ECCS Subsystem)LOCA Loss of Coolant AccidentMAAP Modular Accident AnalysisProgramMAB/MEAB Mechanical AuxiliaryBuilding or MechanicalElectrical Auxiliary BuildingMBLOCA Medium Break Loss ofCoolant AccidentNIST National Institute ofStandards and TechnologyNLHS Non-uniform LatinHypercube SamplingNPSH Net Positive Suction Head,(NPSHA -available,NPSHR -required)NRC Nuclear RegulatoryCommissionNSSS Nuclear Steam SupplySystem NOC-AE-1 4003101Attachment 6Page 2 of 2Definitions and AcronymsODPCIPDFPRAPWRPWROGPWSCCQDPSRAIRCBRCFCRCSRGRHRRI-ISIRMIRMTSRVWLRWSTOuter DiameterPerformance Contracting,Inc.Probability Density FunctionProbabilistic RiskAssessmentPressurized Water ReactorPressurized Water ReactorOwner's GroupPrimary Water StressCorrosion CrackingQualified Display ProcessingSystemRequest for AdditionalInformationReactor ContainmentBuildingReactor Containment FanCoolerReactor Coolant SystemRegulatory GuideResidual Heat RemovalRisk-Informed In-ServiceInspectionReflective Metal InsulationRisk Managed TechnicalSpecificationsReactor Vessel Water LevelRefueling Water StorageTankSBLOCA Small Break Loss of CoolantAccidentSC Stress CorrosionSI Safety Injection (same asECCS)SIR Safety Injection andRecirculationSRM Staff Requirements.MemorandumSTP South Texas ProjectSTPNOC STP Nuclear OperatingCompanyTAMU Texas A&M UniversityTF Thermal FatigueTGSCC Transgranular StressCorrosion CrackingTS Technical Specifcation(s)TSB Technical SpecificationBasesTSC Technical Support CenterTSP Trisodium PhosphateUFSAR Updated Final SafetyAnalysis ReportUSI Unresolved Safety IssueUT University of Texas (Austin)V&V Verification and ValidationVF Vibration FatigueWCAP Westinghouse CommercialAtomic PowerZOI Zone of Influence