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{{Adams
#REDIRECT [[GNRO-2012/00120, Responses to NRC Requests for Additional Information GGNS Criticality Safety Analysis License Amendment Request]]
| number = ML12276A152
| issue date = 10/01/2012
| title = Responses to NRC Requests for Additional Information GGNS Criticality Safety Analysis License Amendment Request
| author name = Ford B S
| author affiliation = Entergy Operations, Inc
| addressee name =
| addressee affiliation = NRC/NRR, NRC/Document Control Desk
| docket = 05000416
| license number = NPF-029
| contact person =
| case reference number = GNRO-2012/00120
| package number = ML122760208
| document type = Letter
| page count = 55
| project =
| stage = Response to RAI
}}
 
=Text=
{{#Wiki_filter:When Attachment 1 is removed from this letter, the entire document is NON-PROPRIETARY.
Attachment 1 contain s PROPRIETARY information.
GNRO-201 2/00120 October 1, 2012 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC  20555
 
==SUBJECT:==
Responses to NRC Requests for Additional Information
- GGNS Criticality Safety Analysis License Amendment Request Grand Gulf Nuclear Station, Unit 1 Docket No. 50
-416  License No.
NPF-29 
 
==REFERENCES:==
: 1. Entergy Operations, Inc. letter to the NRC (GNRO
-20 11/00076), License Amendment Request
- Criticality Safety Analysis and Technical Specification 4.3.1, Criticality, September 9, 2011 (ADAMS Accession No. ML1125321287
)  2. NRC e-mail to Entergy Operations, Inc., Grand Gulf Nuclear Station Request for Additional Information Regarding Criticality Safety Analysis Amendment (ME7540), July 25, 2012 (ADAMS Accession #ML12207A488)
 
==Dear Sir or Madam:==
 
In Reference 1, Entergy Operations, Inc. (Entergy) submitted to the NRC a license amendment request (LAR), which proposes to:  1) revise the criticality safety analysis (CSA) for the spent fuel and new fuel storage racks; 2) impose additional requirements for the spent fuel and new fuel storage racks in Technical Specification (TS) 4.3.1, Criticality
; and 3) delete the spent fuel pool loading criteria Operating License Condition 2.C(45). In Reference 2, the NRC transmitted to Entergy twenty-six (26) requests for additional information (RAIs) pertaining to the CSA LAR. Responses to these RAIs are provided in  to this letter.
Please note that the response to RAI 12 identifies the need to revise TS 4.3.1.1 as proposed in Reference 1. Entergy is currently processing this change and plans to submit it to the NRC by October 22, 2012.
Entergy Operations, Inc.
1340 Echelon Parkway  Jackson, MS  39213 Bryan S. Ford Senior Manager, Licensing Tel.  (601) 368
-5516 GNRO- 2012/00120 Page 2 of 2  Global Nuclear Fuel
- Americas, LLC (GNF) considers certain information contained in Attachment 1 to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding this information, executed by G NF, is provided in Attachment 2. The subject information contained in Attachment 1 was provided to Entergy in a G NF transmittal that is referenced in the affidavit. Therefore, on behalf of G NF, Entergy requests Attachment 1 be withheld from public disclosure in accordance with 10 CFR 2.390(b)(1).
A n on-proprietary, redacted version of Attachment 1 is provided in Attachment
: 3. This letter contains new commitment s, which are identified in Attachment 4
. If you have any questions or require additional information, please contact Guy Davant at (601) 368-5756. I declare under penalty of perjury that the foregoing is true and correct; executed on October 1 , 2012. Sincerely,    BSF/ghd  Attachment s: 1. Responses to NRC Requests for Additional Information
- GGNS Criticality Safety Analysis License Amendment Request (Proprietary Version)
: 2. Affidavit Supporting Request to Withhold Attachment 1 from Public Disclosure
: 3. Responses to NRC Requests for Additional Information  
- GGNS Criticality Safety Analysis License Amendment Request (Non
-Proprietary Version)
: 4. List of Regulatory Commitments cc: Mr. Elmo E. Collins, Jr.
Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 612 East Lamar Blvd., Suite 400 Arlington, TX  76011
-4005  State Health Officer Mississippi Department of Health P. O. Box 1700 Jackson, MS  39215
-1700  U. S. Nuclear Regulatory Commission ATTN: Mr. A. B. Wang, NRR/DORL (w/2)
ATTN: ADDRESSEE ONLY ATTN: Courier Delivery Only Mail Stop OWFN/8 B1 11555 Rockville Pike Rockville, MD  20852
-2378 NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS  39150
 
ATTACHMENT 2 GNRO-2012/00120 AFFIDAVIT SUPPORTING REQUEST TO WITHHOLD ATTACHMENT 1 FROM PUBLIC DISCLOSURE PROVIDED BY GLOBAL NUCLEAR FUEL
- AMERICAS, LLC 3901 CASTLE HAYNE ROAD WILMINGTON, NC 28401
 
LRW-ENO-JB1-12-115 Enclosure 1  Affidavit Page 1 of 3 Global Nuclear Fuel - Americas AFFIDAVIT I, Lukas Trosman, state as follows:
 
(1) I am Engineering Manager, Reload Design and Analysis, Global Nuclear Fuel - Americas, LLC (GNF-A), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.
(2) The information sought to be withheld is contained in Enclosure 1 of GNF letter, LRW-ENO-JB1-12-115, L. Watts (GNF-A) to F. Smith (Entergy), entitled "Revised GNF Response to NRC RAIs for Grand Gulf Fuel Storage Criticality Safety Analysis," dated September 28, 2012. GNF-A proprietary information within the text and tables in Enclosure 1 is identified by a dotted underline inside double square brackets. [[This sentence is an example.
{3}]]  In all cases, the superscript notation
{3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.
(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GNF-A relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exempti on 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F2d 871 (DC Cir. 1992), and Public Citizen Health Research Group
: v. FDA, 704 F2d 1280 (DC Cir. 1983).
(4) Some examples of categories of information which fit into the definition of proprietary information are:
: a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GNF-A's competitors without license from GNF-A constitutes a competitive economic advantage over other companies;
: b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
: c. Information which reveals aspects of past, present, or future GNF-A customer-funded development plans and programs, resulting in potential products to GNF-A;
: d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. above.
 
LRW-ENO-JB1-12-115 Enclosure 1  Affidavit Page 2 of 3 (5) To address 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GNF-A, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GNF-A, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.
(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GNF-A. Access to such documents within GNF-A is limited on a "need to know" basis.
(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GNF-A are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.
(8) The information identified in paragraph (2), above, is classified as proprietary because it contains details of the nuclear fuel criticality licensing methodology for the GEH Boiling Water Reactor (BWR). Development of these methods, techniques, and information and their application for the design, modification, and analyses methodologies and processes was achieved at a significant cost GNF-A.
The development of the evaluation processes along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GNF-A asset.
(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GNF-A's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GNF-A's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value
 
derived from providing analyses done with NRC-approved methods.
The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by GNF-A.
 
LRW-ENO-JB1-12-115 Enclosure 1  Affidavit Page 3 of 3  The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to qua ntify, but it clearly is substantial.
GNF-A's competitive advantage will be lost if its competitors are able to use the results of the GNF-A experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.
The value of this information to GNF-A would be lost if the information were disclosed to the public. Making such information available to competitor s without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GNF-A of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.
I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.
Executed on this 28th day of September 2012.
Lukas Trosman Engineering Manager, Reload Design and Analysis Global Nuclear Fuel - Americas, LLC
 
ATTACHMENT 3  GRAND GULF NUCLEAR STATION GNRO-2012/00120 RESPONSES TO NRC REQUESTS FOR ADDITIONAL INFORMATION
- GGNS CRITICALITY SAFETY ANALYSIS LICENSE AMENDMENT REQUEST (Non-Proprietary Version)
 
to GNRO-2012/00120 Page 1 of 46    RESPONSES TO NRC REQUESTS FOR ADDITIONAL INFORMATION
- GGNS CRITICALITY SAFETY ANALYSIS LICENSE AMENDMENT REQUEST In a letter to the NRC 1, Entergy Operations, Inc. (Entergy) submitted a license amendment request (LAR), which proposes to:  1) revise the criticality safety analysis (CSA) for the spent fuel and new fuel storage racks; 2) impose additional requirements for the spent fuel and new fuel storage racks in TS 4.3.1, Criticality; and 3) delete the spent fuel pool loading criteria Operating License Condition 2.C(4 5). In an e-mail to Entergy 2, the NRC staff transmitted 26 requests for additional information (RAIs) pertaining to the CSA LAR. Responses to these RAIs are provided below.
RAI 1 In general, references are not provided to justify the degree of Boraflex degradation modeled. Relevant assumptions are stated in Section 3.6, but no justification is provided. Reading NEDC-33621P, Rev.
0, it would appear the analyst is merely assuming 30% loss of 10B and that the Boraflex panel shrinkage has reduced the panel width by 4.1%. Justify use of these values.
Response The 30% 10B loss is an analysis assumption, which is confirmed by the Boraflex monitoring program 3 , which determines if a cell meets the criteria for Region I storage.
The 4.1% panel shrinkage is the maximum Boraflex shrinkage reported in the EPRI report NP-6159, "An Assessment of Boraflex Performance in Spent
-Nuclear-Fuel Storage Racks."  This assessment is based on both a theoretical model of Boraflex's performance and post
-irradiation measurements of Boraflex performance up to 10 11 Rads. This parameter only applies to the Region I analysis, which is applicable to doses less than or equal to 2.3 E10 Rads. It is conservatively applied to all panels, at all elevations, independent of dose.
 
1 Entergy Operations, Inc. letter to the NRC (GNRO
-2011/00076), License Amendment Request
- Criticality Safety Analysis and Technical Specification 4.3.1, Criticality, September 9, 2011 (ADAMS Accession No. ML1125321287) 2 NRC e-mail to Entergy Operations, Inc., Grand Gulf Nuclear Station Request for Additional Information Regarding Criticality Safety Analysis Amendment (ME7540), July 25, 2012 (ADAMS Accession #ML12207A488) 3 The Boraflex Monitoring Program was previously discussed in Entergy letters GNRO-2011/00017, 3/9/2012 (see the response to RAI 2 and GNRO
-2011/00104, 11/21/2012 (see the response to RAI 4).
to GNRO-2012/00120 Page 2 of 46    RAI 2 The text in Section 4.3 includes the following statement:
"The fuel loadings considered for each lattice span a range of exposure, average enrichments, number of gadolinium rods, gadolinium enrichment, and void histories considered to be reasonably representative of any GGNS fuel design."
10 CFR 50.68(b)(4) requires that the k
-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95% probability, 95% confidence level. Use of "reasonably representative" fuel assembly design and depletion parameters appears to be inconsistent with these requirements of 10 CFR 50.68. Provide justification that explains how use of reasonably representative values is consistent with 10 CFR 50.68. Response The method used to establish the design basis lattice in NEDC
-33621P, Revision 0 includes studies of extreme lattice designs at different peak reactivity exposures.  [[
        ]] The method used to establish the design basis lattice in NEDC
-33621P, Revision 0 has been generically studied for a similar rack in Reference 2
-1. That study compared the same method used to establish a design basis lattice in NEDC
-33621P, Revision 0, with a method that assumed the appropriate Knominal should be calculated based on the linear fit of in
-core vs. in
-rack reactivities, with the uncertainty of that fit included in the square root of the sum of the squares of all other uncertainty contributors and included in the final K(95/95) result. The study demonstrated, in accordance with the NUREG
-6698 method, that the design basis lattice process used in NEDC
-33621P, Revision 0, effectively captures the uncertainty of the linear fit at 95% probability, 95
% confidence level.
Reference 2-1. Hannah, John C., Metwally, Walid A., & Mills, Vernon W. (2010). Uncertainty Contribution to Final In
-Rack K(95/95) from the In
-Core Kinf Criterion Methodology for Spent Fuel Storage Rack Criticality Safety Analyses.
PHYSOR 2010, Advances in Reactor Physics to Power the Nuclear Renaissance. Pittsburgh, Pennsylvania, USA May 9 - 14, 2010 to GNRO-2012/00120 Page 3 of 46    RAI 3 Confirm that the Region I spent fuel storage rack modules include Boraflex panels on the all outside faces of all rack modules. If Boraflex panels are not present on the outside of all Region I spent fuel storage racks, the evaluation of normal and credible abnormal conditions should be revised to include consideration that some peripheral locations may not have Boraflex panels.
Response Boraflex is present on the outside face of all rack modules.
RAI 4 The following questions concern the modeling of the Region I Boraflex gaps.
: a. Total panel loss and gap location would appear to be correlated with high radiation dose rates. The total panel loss would be higher in all four panels around an assembly that had a high source term and axially the gap locations in all four panels around a high
-dose assembly would be similarly correlated with  a higher axial peaking source term. It is not clear that randomly sampling total panel loss and gap locations, assuming no correlations, is appropriate. Provide justification for ignoring potential correlations between the distributions and between panels in a cell.
Response See response to RAI 4.e(iii), below.
: b. Monte Carlo sampling was performed for [[
  ]]. The reality of the situation is that a large number of the possible distributions will exist in the larger array of rack modules at the same time. Some of the more reactive configurations likely exist. Since criticality is a local phenomenon rather than an average phenomena, using the average keff value does not appear to be appropriate. It seems appropriate to use the highest keff value as the normal keff for the Region I racks. Justify not using the highest keff value as the normal keff value for Region I.
The text on page 33 of NEDC-33621P, Rev 0, describes derivation of a gap modeling total uncertainty. A frequency plot of the results is provided below. It is not clear that the results are normally distributed. What statistical techniques were used to determine the 95% probability/95% confidence keff value?
to GNRO-2012/00120 Page 4 of 46    [[                              ]] Response Table 16 on page 33 of NEDC
-33621P, Revision 0 presents
[[ 
 
    ]] c. The impact of gaps was estimated by simulating only [[  ]]. It is not clear that [[ ]] is enough to accurately quantify the impact of the gaps. One might check this by tracking the evolution of the mean keff value and keff population standard deviation as a function of the number of simulated configurations to assess convergence. Justify limiting the simulation to only
[[ ]] and, if necessary, perform additional simulations.
 
to GNRO-2012/00120 Page 5 of 46    Response The method for using
[[ ]] to accurately quantify the impact of the gaps is
[[  ]] outlined in the response to RAI 4.b, above. d. NEDC-33621P, Rev. 0, Figure 16 shows a
[[      ]]  This could indicate a problem introduced by post
-sampling adjustment or the need for additional simulations. Explain and justify the differences and, if necessary, perform additional simulations.
Response The input probability distributions for gap size per panel and total panel loss per panel conflict when modeling random distributions due to the variable number of gaps. 
[[                ]]  Given that the total panel loss per panel probability distribution was met within tolerance and the average gap size modeled is conservative to the target average gap size, additional simulations are not needed.
: e. Quantify the uncertainty associated with using the Monte Carlo sampling technique. Factors to be addressed include:
(i) Uncertainties associated with the measurements that were used to yield the sampling distributions. Issues of calibration and repeatability should be addressed.
Response Blackness tests were used to develop the Boraflex gap distributions that provide the bases for the analysis sampling distributions. The uncertainty in the Blackness test measurements for gap size is +/
- 20%. This uncertainty is based on the use of a GGNS-specific calibration cell that contains a broad range of gaps of known size. The uncertainty in the location of the gap is estimated to be a few inches. Of the 208 panels to GNRO-2012/00120 Page 6 of 46    measured, 20 panel measurements are repeated with the measurement tool oriented so a different detector is used. Seven Blackness test campaigns were conducted  over approximately 10 years. Comparison of gap measurement s over these campaigns shows good agreement in the location of gaps with gap sizes increasing as accumulate d dose increases.
The uncertainty in measurement in the gap size is significantly less that the conservative bias between the analysis gap size assumptions and the gap measurement distribution 4 so no adjustment is necessary. The gap location uncertainty is less than the distribution bin size and is insignificant relative to the bias between the analysis assumptions and the measurement distribution
: 5. Additional information is provided in response to RAI 4.f, below. (ii) Uncertainties associated with production of the sampling distributions from a finite set of measurements.
Response The gaps size distributions were developed based on the seventh GGNS blackness test campaign, which measured 208 panels. The panels contained a total of 362 individual gaps. This dataset is sufficiently large so that uncertainties associated with the sample size are negligible. This is demonstrated by comparing the change in the t
-distribution at the 5% probability level. The critical value for samples with 200 degrees of freedom is 1.653. It reduces to 1.645, for an infinite number of degrees of freedom. This ~0.5 % change is insignificant relative to margin between the measured and analyzed distributions. Additional information is provided in response to RAI 4.f, below. (iii) Uncertainties associated with correlations between sampling distributions.
Response The potential correlation between the Boraflex sample distributions was evaluated by a regression analysis of the sample distributions. Essentially no correlation is indicated based on the R
-Squared results for all combinations except the Gaps per Panel
-vs.-Total Panel Loss. As illustrated by the following graph, application of this correlation to the Boraflex gap sampling protocol would only produce a large total panel gap loss when many smaller individual gaps are present. The analysis sampling protocol limits the number of gaps to either one or two gaps while independently sampling the total panel loss. This approach produces larger total panel losses than would be obtained if the correlation was applied; therefore, not applying the correlation is conservative. Note that R-Squared is reduced to 0.12 when only one or two gaps/panel is considered. Therefore, no adjustments are merited based on the correlations between samples.
 
4 See Figures 3 and 4 provided in Attachment 1 to Entergy letter GNRO
-2010/00073, November 23, 2010. 5 See Figure 5 provided in Attachment 1 to Entergy letter GNRO
-2010/00073, November 23, 2010.
 
to GNRO-2012/00120 Page 7 of 46    These results are as expected. While panel Boraflex shrinkage is correlated to dose, gap size and location are also influenced by the location of interference points, which restrict panel movement. The panels in the GGNS racks are sandwiched between flat stainless steel sheets in a window frame design. In this configuration, interference points would be expected to occur where local variations in rack/panel geometry are sufficient to restrain the Boraflex. Gaps are expected to form at some location between two such interferences.
Distributions Compared R-Squared Axial Location vs. Gap Size 0.035 Gaps per Panel vs. Total Panel Loss 0.461 Gaps Per Panel vs. Gap Size 0.023 Gaps Per Panel vs. Axial Location 0.000 Total Panel Loss vs. Gap Size 0.125  (iv) Uncertainties associated with post
-sampling adjustment of sampled data.
Response The only post
-sampling adjustments are applied in the procedure used in develop the individual panel configurations used in the Monte Carlo N
-Particle (MCNP) calculations (see Section 6.2.1 of NEDC
-33621P, Revision 0). In cases where two gaps/panel are selected, individual gap sizes may be normalized to conserve the total panel loss. No additional uncertainty is attributable to this step since the total panel loss is conserved to maximize coupling.
 
to GNRO-2012/00120 Page 8 of 46    (v) Uncertainties associated with the continuing time
-dependent evolution of the true distributions.
Respons e The gap distribution s are applied to the Region I rack configuration
, which are limited to cells with an accumulated gamma dose less than or equal to 2.3 E10 Rads. The accumulated dose associated with the seventh blackness test campaign, used to develop the gap distributions, is significantly higher (averages ~3.0 E10 Rads). Since the Region I rack configurations will always have a lower accumulate d dose, no uncertainty associated with the additional evolution of the distribution is needed.
: f. In NEDC-33621P, Rev 0, the target probability distribution for the number of gap per panel, total panel loss per panel, the gap size per panel, and gap axial location per panel were shown in Table 12, Table 13, Table 14, and Table 15 respectively. Please provide the methodology and algorithm used to derive the probability distribution. In addition, explain the correlation of the probability distribution against the accumulated gamma dosage exposure in the Region I spent fuel pool.
Response The Boraflex probability distributions were developed by biasing the measured probability distribution in order to maximize the neutron coupling between gaps. The gaps per panel were reduced to either one or two per panel in order to maximize the gap sizes in a given panel. This approach conservatively includes the less
-reactive configurations (28%) in these two bins. The total panel loss and individual gap size distributions were significantly skewed toward larger gap sizes and axial location probability was compressed to force a ll gaps into the central six feet of the rack even though the measurements show a significant number above and below this region. Similar to many past Boraflex analyses, the extent of the conservatisms included in the analyzed configuration is largely based on judgment.
In order to illustrate the extent of the conservatism
, a systematic evaluation of the measured distributions has been performed. Gap distributions were generated from the measurement database by randomly sampling a subset (25%) of the database. Twenty
-five such sets of randomly generated Boraflex distributions were developed, with replacement. The 95
% percentile upper limit for each bin in each distribution was determined. The gaps sizes and total panel loss for all panels were bias ed by 20% to address the gap size measurement uncertainty noted in the response to RAI 4.e , above. In order to compare these upper limit measurement distributions to the analyzed distributions
, cumulative probability distributions were developed. The summation begins from the most reactive bin in a given probability distribution. The following graphs show that the Boraflex assumptions used in the analysis are significantly more conservative than the upper limit of the measured distributions.
 
to GNRO-2012/00120 Page 9 of 46 to GNRO-2012/00120 Page 10 of 46      The relationship between the gamma dose limit and the gap probability distributions is based on the change in gap
-size distributions between the sixth and seventh blackness campaign s. The gap sizes observed during the sixth campaign were consistent with the EPRI-developed shrinkage model with the cumulative panel loss except for the three highest dose panels. All three of these panels had greater than 2.3 E10 Rads with the lowest at 2.31 E10Rads. Following an additional cycle of irradiation from freshly-discharged fuel, the total panel gap loss accelerated rapidly. This change is attributed to the presence of large gaps already in the panels in combination with gap edge erosion. The presence of the large gaps provides more panel cavity volume and less flow resistance. The dose limit of 2.3E10 Rads is established to limit the presence of large gaps and the potential accelerated gap growth due to edge erosion.
 
to GNRO-2012/00120 Page 11 of 46    g. In NEDC-33621P, Rev 0, section 6.8, the reactivity
[[  ]]. It was not clear how the value was derived.
It was not consistent with the In
-rack K in the Table 18
- Spent fuel Storage Rack In
-Rack K results - Region I. Please explain and justify how this value was derived.
Response [[ 
  ]]  The value and associated uncertainty are calculated per NUREG/CR-6698 guidance.
: h. In NEDC-33621P, Rev 0, Table 13 states the target probability distribution for the total panel loss per panel. According to the probability distribution in the table, it indicates that the cumulative probability more than 50% of the panel have more than 10 inches (or more) of panel loss. However, none of twenty cases was studied with panel loss exceed 10 inches. Please provide explanation of the discrepancy between Table 13
- the target probability distribution and Figure 11
- [[ ]] in the report.
Response Table 13 describes the distribution for
[[ ]]  The cumulative distribution function (CDF) states that 54% of the total panel loss per panel values will fall in any of the bins less than or equal to 10 inches. 
[[      ]]
to GNRO-2012/00120 Page 12 of 46    Table 4-1 Target Probability Distribution for the Total Panel Loss per Panel (Adjusted Bins)
Bin Target Probability Cumulative Probability Bin x  Target Probability 0 0 0 4.5 0.05 0.05 0.225 5.5 0.125 0.175 0.6875 6.5 0.125 0.3 0.8125 7.5 0.08 0.38 0.6 8.5 0.08 0.46 0.68 9.5 0.08 0.54 0.76 10.5 0.08 0.62 0.84 11.5 0.125 0.745 1.4375 12.5 0.125 0.87 1.5625 13.5 0.04 0.91 0.54 14.5 0.04 0.95 0.58 15.5 0.04 0.99 0.62 16.5 0.01 1 0.165  9.51 Tolerance
- +/-.05 
[[      ]]  Figure 14 shows
[[    ]] i. On NEDC-33621P, REV 0, Page 26, it indicated the
[[  ]]  Please provide the explanation on the mathematical model used to generate
[[  ]. Please provide justification for how the output is compatible with MCNP
-05P? Response [[
to GNRO-2012/00120 Page 13 of 46                                            ]] j. According to the response to RAI 32 in the licensee's April 21, 2011, letter only 18 of approximately 9000 Region I Boraflex panels were tested. Justify this limited testing in establishing the Boraflex modeling methodology and any additional bias and uncertainty that are inherent in that limited testing scope.
 
to GNRO-2012/00120 Page 14 of 46    Response The selected panels are representative of the Region I storage cells. Three primary factors impact Boraflex loss, spent fuel pool water environment, rack design, and irradiation history. The first two items are common to all panels in the GGNS pool while the third is dependent on the fuel discharge history. The majority of the panels selected for the Region I tests (10 of 18) have been irradiated since the first fuel discharge into the spent fuel pool, which occurred in 1986. Four of the panels began irradiation in 1996, one in 2004
, and three are in storage cells where fuel has not been stored. The selected panels cover the range of operating times experienced in the balance of Region I. However, they are somewhat biased toward longer operating times and
, therefore , higher panel loss. Additionally, the dose accumulated by these panels covers the full range allowed for Region I storage, so dose levels are also representative of the balance of Region I panels. The difference between RACKLIFE and BADGER areal density results was examined to see if a correlation exists with either panel operating time or dose. No significant correlation was observed. The selected panels are representative of the balance of the Region I panels and the confidence interval used to establish the Boraflex uncertainty accounts for the selected test size. k. Enclosure 1 to the licensee's April 21, 2011, letter indicates un
-irradiated Boraflex panels are degrading. With respect to this degradation provide the following information:
(i) Explain the degradation mechanism for un
-irradiated panels.
Respons e As described in EPRI report TR-107333, "The RACKLIFE Boraflex Rack Life Extension Computer Code:
Theory and Numerics," Boraflex loss does occur at low loss rates when the accumulate d dose is less than 5 E08 Rads. The loss rates increase as accumulate d dose increase s above this value. The presence of a small reduction in areal density for panels with no accounted for dose is not a new mechanism.
Two of the panels with zero
-reported dose, show some gapping. The RACKLIFE model accounts for dose accumulation when fuel is stored in face adjacent cells. The dose accumulation rate accounts for the impact of fuel stored in diagonally
-adjacent cells. However, if no fuel is stored in a face
-adjacent cell, the dose from diagonally
-adjacent fuel is not accounted for. A review of the history of cells ZQ14 and ZQ16 show that fuel stored in diagonally
-adjacent locations would account for the observed gaps. Cell ZR15 has not had fuel stored in diagonally
-adjacent locations and no gaps were observed.
(ii) Explain how this degradation mechanism is incorporated into the Boraflex monitoring program. Response The mechanisms used to determine B4C loss are currently included in the RACKLIFE code; therefore, no adjustments are necessary. This is additionally supported by noting that the difference between the predicted areal density and the BADGER measured results are within one
-sigma of the mean for the zero dose panels.
 
to GNRO-2012/00120 Page 15 of 46    While the contribution to the calculated panel dose from diagonally adjacent fuel assemblies is relatively small, an adjustment to the dose calculations will be augmented to include this effect for empty cells. The RACKLIFE vendor has been contacted to evaluate this impact on other RACKLIFE users. The three panels noted above will be included in the upcoming BADGER test campaign. (iii) Explain how this degradation mechanism is incorporated in the nuclear criticality safety (NCS) analyses for the GGNS SFP.
Response No adjustments are needed since the monitoring program will be modified to account for the effect.
RAI 5 NEDC-33621P, Rev. 0, Sections 6.1 and 6.2 describe the Region II storage configuration and models. As presented, the analysis appears to support only repeated 4x4 sections that are similar to the configuration shown in Figure 18 and, for example, would not support a 3x4 Region II storage area. Describe the minimum acceptable Region II storage area.
Response A 4x4 region is the minimum acceptable area for storing fuel in a Region II configuration.
RAI 6 NEDC-33621P, Rev. 0, Section 6.2.3 presents the model that was used to evaluate the interface condition. 
[[  ]]  This model likely does not yield the most reactive interface keff value. Confirm that the most limiting interface condition has been modeled.
Response [[ 
  ]] RAI 7 NEDC-33621P, Rev. 0, Section 6.3 describes the selection of the "design basis lattice."  From the text and Table 17, it appears that only 15 lattices were evaluated. Justify the selection of the design basis lattice (DBL) from such a limited set of lattices. Include in the justification the logic supporting that using the DBL will ensure that limiting all past, present and future lattice variations that meet the SCCG k limit will also meet the limits described in 10CFR50.68, including the 95% probability and 95% confidence requirement. Justify not specifying a range of acceptable assembly gadolinium (Gd) loading consistent with the evaluated lattices.
 
to GNRO-2012/00120 Page 16 of 46    Response As mentioned in the response to RAI 2, above, the design basis lattice approach used in NEDC
-33621P, Revision 0 has been studied previously for a similar rack analysis (Reference 7
-1). The study used 12 unique lattices with a peak in
-core kinf greater than the limit. The 12 lattices were studied in
-rack at peak reactivity as well as at other exposures to establish the linear relationship of in
-core vs. in
-rack reactivity. The study demonstrated that the design basis lattice process effectively captures the uncertainty of the linear fit at 95
% probability , 95% confidence level. 
[[    ]] The [[ ]] lattices studied in NEDC
-33621P, Revision 0
[[            ]] References 7-1. Hannah, John C., Metwally, Walid A., & Mills, Vernon W. (2010). Uncertainty Contribution to Final In
-Rack K(95/95) from the In
-Core Kinf Criterion Methodology for Spent Fuel Storage Rack Criticality Safety Analyses. PHYSOR 2010, Advances in Reactor Physics to Power the Nuclear Renaissance. Pittsburgh, Pennsylvania, USA May 9-14 2010. RAI 8 NEDC-33621P, Rev. 0, Section 6.4.1 notes that
[[  ]]. The impact of
[[  ]] should have been evaluated. Justify the limited evaluation of
[[ ]]. Response [[    ]]
to GNRO-2012/00120 Page 17 of 46    RAI 9 NEDC-33621P, Rev. 0, Section 6.4.1 includes the following statement:
[[   
  ]] possibly increasing keff. In fact, many of the parametric variations and abnormal conditions should be evaluated [[
]] to ensure that the most reactive condition is identified. Provide justification for limiting the analysis to [[
  ]]. Response [        ]] RAI 10 NEDC-33621P, Rev. 0, Table 18 shows that impact of
[[    ]]. Provide justification for use of this
[[  ]]. Tables 18 and 20 include different keff values for the Region I rack keff with [[    ]]. The value in Table 20 appears identical to the value in Table 18 for
[[ ]]. Additionally, the value listed in Table 20 for
[[  ]] appears to be identical to the value in Table 18 for
[[ ]]. Describe and justify the differences, or correct any errors in the data and report.
Response [[ ]] provide limiting values for the range of possible temperatures in the spent fuel pool. 
[[ 
  ]] Discussions with GNF indicate they plan to correct an error in NEDC
-33621P, Revision 0 Table B6 B7 cases will be corrected to provide the following data consistent with the results in NEDC
-33621P, Revision 0 Table 18. See the response to RAI 16, below for an updated Table
: 20.
to GNRO-2012/00120 Page 18 of 46    RAI 11 NEDC-33621P, Rev. 0, Table 18 shows the impact of modeling variations on the Region I keff value.  [[
    ]]  Justify not evaluating the impact of parameter variations with [[
  ]]. Response The gaps were modeled for all parameter variation cases in Table 18. As stated in the last paragraph of Section 6.2.1 in NEDC
-33621P, Revision 0:
[[    ]] RAI 12 NEDC-33621P, Rev. 0, Table 19 shows the impact of modeling variations on the Region II keff value. The following issues are noted concerning the information provided in Table 19.
: a. Parameter variations should have been examined
[[  ]]. Justify not examining parameter variations
[[ ]]. b. Considering which storage cell positions may be used in Region II, it is not clear that
[[  ]] will yield the highest keff value. Justify considering only the
[[  ] case. c. In Region I, bundle rotations were evaluated
[[ ]]. If [[  ]] were evaluated in Region II the same way, additional
[[  ]] should have been considered. Justify examining only the
[[    ]]. d. The impact on keff of moderator density appears to be significantly higher
[[  ]]. Since the impact was evaluated only at
[[ ]], it is not clear that the peak value has been identified. Justify evaluating moderator densities at only
[  ]]. Response Region II has been redefined to require a full checkerboard pattern (8 empty out of 16 cells) as shown in Figure 1 2-1, below. The change in configuration was necessary to produce a final kmaximum value below the 0.95 limit with the inclusion of a misloaded assembly case as an abnormal bias condition.
6  Additional updates to the Region I and Region II analyses have be en incorporated and are described below.
 
6 The CSA LAR submitted via Entergy letter GNRO
-2011/00076 (ADAMS Accession #ML1125321287) proposed a change to TS 4.3.1 that specified six of 16 cells in a Region II 4x4 cell array will be to GNRO-2012/00120 Page 19 of 46    Actinide Validation As stated in the response to RAI 19.e, below, an updated MCNP validation has been performed
[    ]] Fission Product Validation Section 5 of NUREG/CR
-6698 provides guidance on applying appropriate factors to a validation study to justify an extension of the area of applicability beyond the range of the critical experiment bounds. In this section, the following guidance is provided:
"Extension of the area of applicability may require an additional margin of subcriticality to account for increased uncertainty in the bias results due to extrapolation of the validation results. Determination of this additional margin, AOA, should be based on the results of the sensitivity study (bias trends) as well as engineering judgment."
[[              ]] As specified in NUREG/CR
-6698, the required additional margin to account for this validation extension can then be applied by increasing the uncertainty in the bias. 
[      ]]                                       
 
blocked. In order to reflect that eight rather than six cells will be blocked as determined in the criticality safety analysis , Entergy is processing a change to proposed TS 4.3.1.
This revision will be submitted under separate letter on or before October 22, 2012.
to GNRO-2012/00120 Page 20 of 46    Fuel Depletion Uncertainty The uncertainty applied for fuel depletion calculations has been [[treated as an uncertainty{3}]] in the updated analyses for Region I and Region II, consistent with the following technical guidance from NRC Division of Safety Systems Interim Staff Guidance Document DSS
-ISG-2010-01 Section IV.2:
"A reactivity uncertainty due to uncertainty in the fuel depletion calculations should be developed and combined with other calculational uncertainties. In the absence of any other determination of the depletion uncertainty, an uncertainty equal to 5 percent of the reactivity decrement to the burnup of interest is an acceptable assumption". 
[[ 
                        ]] Figure 12-1: Region II New Configuration All Region II cases were rerun using the full checkerboard configuration. Updates to NEDC-33621P, Revision 0, Tables 17, 19, 21, 23, 25, and 27 are provided below.
 
to GNRO-2012/00120 Page 21 of 46    Table 17 -Fuel Parameter Ranges Studied in Spent Fuel Rack Case Lattice Type Void Average Lattice Enrichment (U235 wt%)
Number of Gad Rods* Gad Enrichment (Gd wt %)*
Peak-Reactivity Exposure (GWd/ST) TGBLA06 Defined In-Core k Region I Region II MCNP-05P Defined In-Rack k Rack Efficiency MCNP-05P Defined In-Rack k Rack Efficiency
[[                                                                                                                                                                                  ]]  [[ ]]
to GNRO-2012/00120 Page 22 of 46    Table 19 - Spent Fuel Storage Rack In
-Rack K Results - Region II Configuration In-Rack k Error (1) [[                          ]]  [[  ]] Table 21 - Spent Fuel Storage Rack Bias Summary
- Region II Term Description Keff Error (1)    (2) [[                                                                                ]] [[ ]]
to GNRO-2012/00120 Page 23 of 46    Table 23 - lts - Region II Term Description Keff Error (1) K K Uncertainty (2) [[                                                                                      ]]  Table 25 - - Region II Term Description Value [[                      ]] 
[[  ]]
to GNRO-2012/00120 Page 24 of 46    Table 27 - Spent Fuel Storage Rack Results Summary
- Region II Term Value [[        ]]  Individual responses to items 12a through 12d are provided below.
Response to RAI 12a As demonstrated in Table 19, above, [[      ]] Response RAI 12b Region II has been reanalyzed with a full checkerboard pattern. 
[[      ]] as shown above in the updated Table 21, above.
 
to GNRO-2012/00120 Page 25 of 46    [[     
 
  ]] Figure 12-2: [[ ]]
to GNRO-2012/00120 Page 26 of 46    [[   
 
      ]] Figure 12-3: [[ ]] Response to RAI 12c
[[    ]] as shown above in the updated Table
: 21. Response to RAI 12d
[[ ]] provide limiting values for the range of possible temperatures in the spent fuel pool. 
[[
    ]] as shown above in the updated Table 21.
 
to GNRO-2012/00120 Page 27 of 46    RAI 13 NEDC-33621P, Rev. 0, Section 6.5.1 lists "[[
  ]]" to be an abnormal or accident condition.
: d. Describe eff value of the insert was derived. If it does not include an insert in all cells required to be empty, then adjust it as appropriate so that it does.
Response [[    ]] e. No description of the rack inserts is provided. Provide a description of the permitted rack inserts. Response The inserts are aluminum boxes with an outside dimension (OD) of 5.4 inches and an inside dimension (ID) of 5.0 inches. Inserts may be made of aluminum, iron, or other material with neutron absorption properties greater than or equal to aluminum with a n equal to or smaller OD and an equal to or smaller total cross
-sectional area.
: f. The presence of the rack inserts are a normal condition. Consequently, their presence should be considered when evaluating the impact of tolerances, parameter variations, and accidents/abnormal conditions. Justify not evaluating the impact of tolerances, parameter variations, and accidents/abnormal conditions.
Response [[ 
  ]] RAI 14 NEDC-33621P, Rev. 0, Section 6.5.1 addresses abnormal conditions evaluated. The following items are related to abnormal conditions.
: c. No abnormal conditions are identified related to the degraded Boraflex that is credited in Region I. There are at least three abnormal conditions that should be considered. They are (1) the impact of a seismic event on the continued effectiveness of credited Boraflex, (2) the impact of a dropped load on the continued effectiveness of credited Boraflex, and (3) that the Boraflex degradation is more extensive generally or locally than measurements and analysis indicate. Provide justification for not evaluating abnormal conditions related to Boraflex degradation.
 
to GNRO-2012/00120 Page 28 of 46    Response Seismic Event EPRI evaluated the performance of Boraflex under seismic conditions in report TR-109927, "The Performance of Irradiated Boraflex Under Seismic Conditions."  Boraflex samples that were irradiated up to 3.0 E10 Rads were tested to determine their mechanical properties (Flexural Strength and Young's Modulus). The peak strains in the Boraflex were calculated assuming limiting safe shutdown earthquakes for several rack designs including the GGNS rack design. In all cases, the calculated Boraflex stresses were less than the Boraflex threshold failure stress by a substantial margin. Therefore, a seismic event will not unduly impact the Boraflex structure.
The Boraflex in the GGNS rack design is sandwiched between steel plates with little clearance. Interferences to movement of the Boraflex panels are sufficient to preclude movement demonstrated by the formation of gaps from panel shrinkage. These interferences would provide a significant resistance to the panels slumping downward during a seismic event. As a conservative approach, calculations were performed to demonstrate criticality margins would be maintained should slumping occur. These calculations were previously described in the response to RAI 30 provided in Entergy letter GNR0
-2011/00025 (ADAMS Accession #ML11112A098).
Fuel Drop Event A GGNS-specific analysis was not readily available; however, an analysis for a similar rack was evaluated and determined to be bounding for the GGNS spent fuel racks and fuel handling equipment. The consequences of dropping a fuel bundle from the maximum height achievable with the refueling building 5
-ton crane (~36 feet) was evaluated to determine the impact on the spent fuel racks. This drop does not credit fuel handling procedure limits. A drop from this height significantly bounds a drop from the fuel mask that is used to move fuel into and out of spent fuel pool storage locations. The drop assumed a channeled assembly and the supporting attachment (1200 lbs.) fell onto a corner of the spent fuel rack. The impact on the rack structure was evaluated using the LS-DYNA code. The two panels of the rack corner showed some amount of plastic deformation in the top 24 inches of the rack.
The top of the GGNS Boraflex panels are located ~17 inches below the top of the rack structure; therefore, some change in geometrical structure of the Boraflex would be expected. The reactivity consequence was determined by assuming that the event actually occurred in the center of the rack and the Boraflex material in the top 7 inches was removed from 4 panels arranged in a cross configuration. Since the top of active fuel is 3 inches above the top of the Boraflex, 10 inches of active fuel would be uncovered in the four impacted panels. The first six inches of active fuel is naturally
-enriched Uranium, so only 4 inches of enriched fuel is impacted. Reactivity does not increase due to the high neutron leakage in the impacted area.
 
to GNRO-2012/00120 Page 29 of 46    Boraflex Degradation more Extensive A specific allowance beyond the current analysis assumptions are not proposed due to the extensive conservatisms already present in the current analysis. These include significant margins between the measured gap performance and the analysis assumptions, appropriate uncertainties in the Boraflex loss assumptions, margin to the criticality acceptance criteria, and the application of a strong monitoring program.
: d. Fuel handling activities are listed under "abnormal" conditions. Fuel handling is not an abnormal condition. The evaluation provided in Section 6.5.1 does not indicate how close the three evaluated assemblies may be to the spent fuel storage racks. Confirm that the three assemblies in and around the "upender" are far enough from fuel storage racks to preclude neutronic interaction.
Response The up-ender is located in the fuel transfer canal. The fuel transfer canal is separated from the spent fuel pool by 3 feet, 103/4 inches of concrete and a 1/4
-inch steel liner in both the canal and the spent fuel pool.
: c. Further, the analysis presented states that
[[
    ]]  From the analysis presented, it is not clear that some other spacing between
[[  ]] would not yield a higher keff value than the
[[ ]] spacing case. Confirm that the
[[ ]] case yields the peak keff value. Response [[ 
      ]] RAI 15 NEDC-33621P, Rev. 0, Table 21, [[ ]]. This parameter involves
[[  ]]  One expects the fuel reactivity to increase as the fuel temperature at which it was depleted increases. Confirm that the fuel temperatures used in both the nominal and very high temperature case were modeled correctly.
 
to GNRO-2012/00120 Page 30 of 46    Response The Region II analysis was updated to include a full checkerboard pattern (8 out of 16 empty cells), as described in the response to RAI 12. 
[[    ]] RAI 16 NEDC-33621P, Rev. 0, Table 22 presents results for fuel and rack manufacturing tolerances and uncertainties for the Region I racks. The following comments concern the information provided or not provided in this table:
: a. Provide justification for why the following were not evaluated:
(i) [[ (ii)  (iii)  (iv)  (v)  ]] b. The text on page 10 states
[[  ]], while the text on page 46 states
[[ ]]. Confirm that the correct value was used for Table
: 22. Respons e Additional cases were run for the Region I configuration. Updates in the treatment of validation uncertainties were made as described in the response to RAI 12. Updates to NEDC-33621P, Revision 0 Tables 18, 20, 22, 24, and 26 are included below.
 
to GNRO-2012/00120 Page 31 of 46    Table 18 - Spent Fuel Storage Rack In
-Rack K Results - Region I Configuration In-Rack k Error (1) [[                          ]]  [[  ]]  Table 20 - Spent Fuel Storage Rack Bias Summary
- Region I Term Description Keff Error (1) K K Uncertainty (2) [[                                                              ]] 
[[ ]]
to GNRO-2012/00120 Page 32 of 46    Table 22 - - Region I Term Description Keff Error (1) K K Uncertainty (2) [                                                                                      ]]  Table 24 - - Region I Term Description Value [[                          ]] [[  ]]
to GNRO-2012/00120 Page 33 of 46    Table 26 - Spent Fuel Storage Rack Results Summary
- Region I Term Value [[        ]] Individual responses to items 16a and 16b are provided below.
Response to RAI 16a (i) Cases were added to both the Region I analysis and the updated full checkerboard Region II analysis to evaluate the effect of
[[      ]] (ii) Cases were added to both the Region I analysis and the updated full checkerboard Region II analysis to evaluate the effect of
[[    ]] (iii) Cases were added to both the Region I analysis and the updated full checkerboard Region II analysis to evaluate the effect of
[[ 
  ]] (iv) [[  ]] (v) [[  ]] Response to RAI 16b The correct
[[ ]]. The tolerance cases for both Region I and the updated Region II configuration have been updated accordingly, as shown in the updated Tables 22 and 23, above.
 
to GNRO-2012/00120 Page 34 of 46    RAI 17 NEDC-33621P, Rev. 0, Table 23 presents results for fuel and rack manufacturing tolerances and uncertainties for the Region II racks. The following comments concern the information provided or not provided in this table:
: c. Provide justification for why the following were not evaluated:
(i) [[ (ii)  (iii)  ]] d. The text on page 10 states a tolerance of 7.5% on Gd enrichment, while the text on page 46 states a tolerance of 7%. Confirm that the correct value was used for Table 23.
Response See the response to RAI 16, above.
RAI 18 NEDC-33621P, Rev. 0, Table 24 includes [[  ]]. This term should be a bias rather than an uncertainty. See RAI 4.b above for additional discussion. Justify inclusion of this quantity as an uncertainty rather than a bias.
Response As discussed in the response to RAI 4.b, above, [[        ]] RAI 19 With respect to the validation work provided in NEDC
-33621P, Rev 0, provide the following information:
: a. Table 3 lists [[
    ]]  Clarify [[
]] in the validation benchmarks and justify any difference with the GGNS SFP.
 
to GNRO-2012/00120 Page 35 of 46    Response The Gd form for each experiment set is shown in the table below.  [[
    ]]  All critical benchmark experiments that contained gadolinium had UO 2-Gd rods in the pin lattice which is consistent with BWR fuel geometry. Summary of the Critical Benchmark Experiments Experiment Experiment s Year Where Gadolinium Form
 
[[                                                  ]]  b. The validation set does not describe how many of the critical experiments have Gd. Considering the importance of Gd in the calculation of peak reactivity, Gd should be significantly represented in the validation set. Identify the Gd experiments used in the validation set and justify their adequacy for validating Gd.
 
to GNRO-2012/00120 Page 36 of 46    Response The critical experiments containing Gd are identified in response to RAI 19a, above. A significant portion of the validation set [[
]] contain UO 2-Gd rods similar to those used in BWR fuel bundles and are, thus, adequate for validating Gd.
: c. No trending analysis of benchmark results is provided. Trends should have been evaluated as a function of EALF, enrichment, Pu content (e.g., g Pu / (g Pu + g U), Gd worth, pin pitch, etc. The safety analysis models should be compared to the ranges of these trends. Credit for reduced bias and bias uncertainty should be taken only where statistically significant trends exist. Evaluate the bias and bias uncertainty as a function of trend parameter.
Response To determine if any trend is evident in the pool of critical benchmark experiments, the parameters listed in the following table were considered as independent variables.
Trending Parameters Energy of the Average Lethargy Causing Fission (EALF)
Uranium Enrichment (wt% U235)
Plutonium Content (wt% Pu239)
Atom Ratio of Hydrogen to Fissile Material (H/X)
Each parameter was plotted against the kcalc results independently for each case that was analyzed. These plots are provided in the following figures. The scatter plots of data were first analyzed by visual inspection to determine if any trends were readily apparent in the data. No clear evidence of a trend, linear or otherwise, was observed from this inspection.
 
to GNRO-2012/00120 Page 37 of 46    [[     
 
]] Scatterplot of EALF versus knorm [[       
 
]] Scatterplot of wt% U235 versus knorm to GNRO-2012/00120 Page 38 of 46      [[         
  ]] Scatterplot of wt% Pu239 versus knorm to GNRO-2012/00120 Page 39 of 46    [[ 
 
]] Scatterplot of H/X versus knorm 22 is an indicator of the agreement between the observed (calculated) and expected (fitted) values for some variable. A more convenient way to report this result is the reduced chi squared value, which is denoted as "" and is defined by the following equation, where "d" is the degrees of freedom for the evaluation.
If a value of 1 or less is obtained for this equation, then the expected (fitted) distribution is reasonable; however, if the value is much larger than 1, the expected distribution is unlikely to be a good fit. Results for each trending parameter are summarized in the table below. To further check for trends in the data, a linear regression was performed. The fitted lines are included in the above figures. The linear correlation coefficient, r 2, is a quantitative measure of the degree to which a linear relation exists between two variables. Results from this linear regression evaluation are summarized in the table below.
 
to GNRO-2012/00120 Page 40 of 46    Trending Results Summary Trend Parameter Intercept Slope r 2  Valid Trend H/X [[    U-235 wt%      EALF      Pu-239 wt%    ]]  [[  ]] d. Page 4 in Section 3.3 contains the following text:
[[    ]] While this may be small, it is probably not "insignificant."  Evaluate the impact of the variation of cross sections with temperature.
Response [[   
  ]] e. Section 3.3 identifies a [[  ]]. The text does not say [[  ]]. The calculation of [[
  ]] presented in Table 21 shows that [[
    ]]  Describe which actinides were retained and which were removed in this determination.
Response An updated MCNP validation has been performed [[
    ]]  The new results indicate no change in the applied bias and a slight reduction in the bias uncertainty as shown below.
 
to GNRO-2012/00120 Page 41 of 46    Recommended Bias and Bias Uncertainty in Criticality Analyses for MCNP with ENDF/B
-VII Bias [[ Bias Uncertainty (95/95 Confidence Level)
]]  f. In NEDC-33621P, REV 0, Page 3, Table 2
- Summary of the critical benchmark experiments: [[
]]. According to ICSBEP (September 2010), both of them were identified with borated stainless steel plate. However, based on the GE14 and GNF2 fuel description in Section 4.0 Fuel Design Basis, no borated stainless steel plate is found in the fuel assembly or the GGNS SFP storage racks. Please provide appropriate justification how the selected criticality experiments validate the criticality analysis on the cross section data in the AOA (Area of Applicability) of spent fuel assembly.
Response The experiments used in the validation were selected to cover a number of moderator
-to-fuel ratios and poison materials that represent material and geometric properties similar to that of BWR fuel lattices both in and out of fuel storage racks. 
[[      ]] the main neutron absorbing material, boron, is the same and the critical experiments chosen for the validation adequately validate the boron cross
-sections stated in the AOA.
References 19-1. F. Femex, "Programme HTC
- Phase 1: R~seaux de crayons dans l'eau pure (Wate r-moderated and reflected simple arrays) R6valuation des experiences,"
DSU/SEC/T/2005
-33/D.R., Institut de Radioprotection et de Siiret6 Nucl~aire, 2008 19-2. F. Femex, "Programme HTC
- Phase 2: R~seaux simples en eau empoisonn~e (bore et gadolinium) (Reflected simple arrays moderated by poisoned water with gadolinium or boron) R6valuation des experiences," DSU/SEC/T/2005
-38/D.R., Institut de Radioprotection et de Siiret6 Nucl~aire, 2008 19-3. F. Fernex, "Programme HTC
- Phase 3: Configurations "stockage en piscine" (Pool storage)R6valuation des experiences," DSU/SEC/T/2005
-37/D.R., Institut de Radioprotection et de Siiret6 Nucl~aire, 2008 19-4. F. Fernex, "Programme HTC
- Phase 4: Configurations "chateaux de transport" (Shipping cask)
-R6valuation des experiences," DSU/SEC/T/2005
-36/D.R., Institut de Radioprotection et de Siiret6 Nucl~aire, 2008
 
to GNRO-2012/00120 Page 42 of 46    RAI 20 In NEDC-33621P, Rev. 0, Table 16 it is unclear what is meant by the Error column. Describe what that column means, how it is used, and its justification.
Res ponse The Error column in Table 16 of NEDC
-33621P, Revision 0 is the estimated standard deviation reported in the MCNP output for each of the gap configuration cases. 
[[ 
  ]] RAI 21 As the Region I cells are converted to Region II cells the capacity of the SFP decreases. The SFP capacity is listed in TS Spec 4.3.3. According to the response to RAI 35 in the licensee's April 21, 2011, letter the RACKFILE code prediction shows significant increase in number of Region II storage cells by Jan 1, 2015. How does the licensee plan to maintain compliance with TS 4.3.3 as the SFP capacity continues to decrease?
Response There are a number of options available to meet the reduction in storage locations associated with increases in Region II storage configuration.
These include additional dry cask storage; re-rack project and use of poison inserts. GGNS has not yet selected their preferred long
-term option.
RAI 22 The list of assumptions and conservatisms listed in Section 3.6 includes a statement that [[
  ]]. This is not a conservatism. Provide a list of the nuclides credited in the analysis. Justify the use of any nuclide whose half
-life less than the life of the plant. Response This issue has been previously studied generically in a storage rack analysis for 10x10 fuel.
Pin-specific isotopic modeling as a function of exposure is performed based on the NRC
-approved lattice physics code TGBLA06A.  [
[      ]]  As noted in NEDC
-33621P, Revision 0, Section 3.4, [[ 
    ]]
to GNRO-2012/00120 Page 43 of 46    The in-rack fuel model used in this analysis is described in detail in NEDC
-33621P, Revision 0, Section 4.3. 
[[            ]] Table 22-1 TGBLA06A Days of Decay Post- Shutdown Standard Cold Core Geometry k inf [[              ]]  RAI 23 The analysis incorrectly identifies some depletion parameter variations, and expected operations, such as [[
]]. This is not correct. For example, [[
  ]]. For most cases, this has little impact because the reactivity effect is added as a bias. Ensure that all variations of the appropriate normal conditions are considered when evaluating the impact of tolerances and uncertainties and the impact of truly abnormal and accident conditions.
Response There are some cases under Section 6.5, "Accident/Abnormal Configuration Analysis," that are not accident or abnormal cases, such as the depletion and fuel handling cases.
[[
to GNRO-2012/00120 Page 44 of 46        ]] RAI 24 The NRC staff has noted several typographical errors that should be addressed.
: a. Reference to Table 2 on page 4 should be to Table 3.
: b. Text at the end of the first paragraph on page 5 says [[
  ]]  It should be covering the [[
  ]]. c. Text at the end of the next bullet on page 5 says [[
  ]]  It should be covering the [[
]]. Response GNF has agreed to address these typos in a future revision of NEDC
-33621. RAI 25 Verify that the fission source has converged for all MCNP cases. For any MCNP case for which the fission source has not converged, determine the effect on the estimation of keff.
Response All MCNP cases had a converged fission source except the following five cases
:  [[          ]] [[  ]]  Although the fission source of the above cases did not converge, the impact to the estimation of keff is negligible. Confidence in the results is maintained due to the general agreement between similar runs that did and did not converge and the consistency across different run times for those cases which did not converge. Furthermore, even if there were a statistically significant increase in reactivity of one of the above cases due to the fission source not being adequately converged, [[  ]]
to GNRO-2012/00120 Page 45 of 46    RAI 26 Attachment 3 Figure 29
-1 shows the event tree that models different avenues to correctly and incorrectly load fuel into the spent fuel pool. Based on the event tree analysis, the licensee states that the frequency of misloads is approximately 2.70 E
-7/year. The staff has sufficient indication that the industry misload frequency is approximately two per year or 1.92 E
-02/year per unit. The licensee asserts that their practices and procedures regarding loading assemblies into the spent fuel pool are one hundred thousand times better than historical industry average. The staff finds no reason to believe this assertion.
Furthermore, the staff finds the event tree calculation conducted by the licensee misleading. The licensee indicates two fuel movement plan errors with no dependency; whereas, the same code and organization are used for both plans indicating a clear dependence. The possibility of having a misload based solely on an incorrect fuel plan is not modeled; although, a large majority of actual misloads in industry are due primarily to an incorrect fuel plan.
Standard Review Plan 2.2.3, "Evaluation of Potential Accidents" Section II provides criteria to determine credible scenarios:
"The identification of design
-basis events resulting from the presence of hazardous materials or activities in the vicinity of the plant or plants of specified type is acceptable if all postulated types of accidents are included for which the expected rate of occurrence of potential exposures resulting from radiological dose in excess of the 10 CFR 50.34(a)(1) as it relates to the requirements of 10 CFR Part 100 is estimated to exceed the NRC staff objective of an order of magnitude of 1E
-7 per year."
For those initiating events greater than 1E
-7 per year, the staff finds the scenario credible for review. The initiating event for this scenario is the incorrect fuel plan as all additional steps are categorized as mitigating actions. The staff does not find a mis
-load as an initiating event. The incorrect fuel plan (FMP ERROR) is 1.00E
-3; and for an additional error in the fuel plan (FMP ERR2), the total continues to stay above 1E
-7 so this event is credible.
The staff also recognizes that in June 2008, Grand Gulf inadvertently loaded thirty four fuel assemblies into four casks (NUREG/CR
-6998[7] Event Notification Report #44306 June 20, 2008). The mis
-load was preliminary attributed to an error in the Cask Loader Database and was discovered during a data update of the database. Clearly, the organizational practices and procedures in place for cask loading are not representative of 1E
-7/year. The NRC staff finds that this event is credible because the historical industry record indicate large frequency of occurrence and the occurrence probability of the initiating event is greater than 1E-7. Please provide the criticality analysis for the limiting misload accident at GGNS.
Response As stated in the response to RAI 12, above, Region II has been redefined to require a full checkerboard pattern (8 empty out of 16 cells). 
[[ 
 
to GNRO-2012/00120 Page 46 of 46        ]]
ATTACHMENT 4 GRAND GULF NUCLEAR STATION GNRO-2012/00 120 LIST OF REGULATORY COMMITMENTS to GNRO-2011/00091 Page 1 of 1 LIST OF REGULATORY COMMITMENTS
 
The following table identifies those actions committed to by Entergy in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.
COMMITMENT TYPE (Check one)
SCHEDULED COMPLETION DATE  (If Required)
ONE-TIME ACTION CONTINUING COMPLIANCE The CSA LAR submitted via Entergy letter GNRO
-2011/00076 (ADAMS Accession #ML1125321287) proposed a change to TS 4.3.1 that specified six of 16 cells in a Region II 4x4 cell array will be blocked. In order to reflect that eight rather than six cells will be blocked as determined in the criticality safety analysis, Entergy is processing a change to proposed TS 4.3.1.
This revision will be submitted under separate letter on or before October 22, 2012.
10/22/2012 While the contribution to the calculated panel dose from diagonally
-adjacent fuel assemblies is relatively small, an adjustment to the dose calculations will be augmented to include this effect for empty cells.
The zero-do s e panels in cells ZQ14, ZQ16, and Zr15 will be included in the upcoming BADGER test campaign. 12/31/2012}}

Latest revision as of 03:56, 12 April 2019

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