ML16034A216

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Response to Request for Additional Information Regarding License Amendment Request to Adopt Dominion Core Design and Safety Analysis Methods and to Address the Issues Identified in Westinghouse Documents NSAL-09-5, Rev. 1
ML16034A216
Person / Time
Site: Millstone 
Issue date: 01/28/2016
From: Mark D. Sartain
Dominion Nuclear Connecticut
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
16-011, CAC MF6251
Download: ML16034A216 (15)
v • d • e


Text

S JDominion Dominion Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060 Web Address: www.dom.com U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 January 28, 2016 Serial No.

NLOSIWDC Docket No.

License No.16-011 R0 5 0-4 23 NPF-49 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST TO ADOPT DOMINION CORE DESIGN AND SAFETY ANALYSIS METHODS AND TO ADDRESS THE ISSUES IDENTIFIED IN WESTINGHOUSE DOCUMENTS NSAL-09-5. REV. 1. NSAL-15-1. AND 06-1C-03 (CAC NO. MF6251)

By letter dated May 8, 2015, Dominion Nuclear Connecticut, Inc. (DNC) submitted a license amendment request (LAR) for Millstone Power Station Unit 3 (MPS3).

The proposed amendment would revise the Technical Specifications (TS) to enable use of the Dominion nuclear safety and reload core design methods for MPS3 and address the issues identified in three Westinghouse communication documents.

In a letter dated January 8, 2016, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) to DNC related to the LAR.

The RAI contained eighteen questions. Attached is the DNC response to RAI Questions RAI-1 through RAI-8, RAI-13, and RAI-18.

DNC plans to submit the response to the remaining RA! questions, RAI-9 through RAI-12 and RAI-14 through RAI-17, by February 29, 2016.

The attachment to this letter provides DNC's response to the NRC's RAI.

If you have any questions regarding this submittal, please contact Wanda Craft at (804) 273-4687.

Sincerely, Mark D. Sartain Vice President - Nuclear Engineering COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO G

-;ARY DON M ILLER --

4Notary Public Commonwealth of Virginia

[

~~Reg. # 7629412 My Commission Expires August 31, 20.L13

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The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark D. Sartain, who is Vice President -

Nuclear Engineering of Dominion Nuclear Connecticut, Inc.

He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this 28,

  • 'day of~...
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2016.

My Commission Expires:

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Serial No.16-011 Docket No. 50-423 Page 2 of 2 Commitments made in this letter: None

Attachment:

Response to Request for Additional Information Regarding License Amendment Request to Adopt Dominion Core Design and Safety Analysis Methods and to Address the Issues Identified in Westinghouse~Documents NSAL-09-5, Rev. 1, NSAL-15-1, and 06-1C-03 (CAC No. MF6251) -

RAI Questions RAI-1 through RAI-8, RAI-13, and RAI-18 cc:

U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 R. V. Guzman Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08-C2 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director, Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No.16-011 Docket No. 50-423 ATTACHMENT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST TO ADOPT DOMINION CORE DESIGN AND SAFETY ANALYSIS METHODS AND TO ADDRESS THE ISSUES IDENTIFIED IN WESTINGHOUSE DOCUMENTS NSAL-09-5. REV. 1. NSAL 1, AND 06-1C-03 (CAC NO. MF6251) - RAI QUESTIONS RAI-1 THROUGH RAI-

8. RAI-13. AND RAI 18 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.16-011 Docket No. 50-423 Attachment, Page 1 of 12 By letter dated May 8, 2015, Dominion Nuclear Connecticut, Inc. (DNC) submitted a license amendment request (LAR) for Millstone Power Station Unit 3 (MPS3).

The proposed amendment would revise the Technical Specifications (TS) to enable use of the Dominion nuclear safety and reload core design methods for MPS3 and address the issues identified in three Westinghouse communication documents.

In a letter dated January 8, 2015, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) to DNC related to the LAR. The RAI contained eighteen questions. This attachment provides DNC's response to RAI Questions RAI-1 through RAI-8, RAI-13, and RAI-18.

DNC plans to submit the response to the remaining RAI questions by February 29, 2016.

RAI - 1 (SNPB):

Parameter Uncertainties Section 3.2 and Table 3.2-1 of Attachment 6 to the LAR list parameters and their uncertainties. Describe how these uncertainties were used to quantify the total uncertainty for each of these parameters.

DNC Response As stated in Section 3.2 of Attachment 6 to the May 8, 2015 LAR, the magnitudes and functional forms of the uncertainties for the statistically treated parameters were derived in a rigorous analysis of plant hardware and measurement/calibration procedures. Total uncertainties were quantified at the 20 level, corresponding to two-sided 95% probability for the normally distributed parameters. The standard deviations (a) were obtained by dividing the total uncertainty by 1.96, which is the z-value for the two-sided 95%

probability of a normal distribution.

The sampled parameters FAH engineering uncertainty (FAHE) and the bypass flow fraction are treated as having a uniform distribution, which is consistent with the approved methodology of VEP-NE-2-A. The magnitude of FARE was obtained from the fuel vendor, consistent with the approved methodology of DOM-NAF-2-P-A (Section 4.10). The uncertainty applied to the bypass flow fraction is the difference between the bypass flow fraction assumed in statistical DNB evaluations and that assumed in deterministic DNB evaluations.

Consistent with the approved methodology of VEP-NE-2-A, the total uncertainty used in the calculation of the Statistical Design Limit (SDL) is determined by a Monte Carlo assessment performed at each of the nominal statepoints by varying the parameters within their quantified uncertainty range.

Serial No.16-011 Docket No. 50-423 Attachment, Page 2 of 12 RAI -2 (SNPB):

Calculations of Critical Heat Flux Correlation Uncertainty Factor (a) Provide detailed calculations to obtain the CHF correlation uncertainty factor using the 95% upper confidence limit on the VIPRE-DAVVRB-2M and ABB-NV code correlation pair measured-to-predicted (MIP) CHF ratio and standard deviation.

(b)

Equation 3.2 of Section 3.3 of Attachment 6 is used to calculate the upper confidence limit and is a correction factor that gives one-sided 95% upper confidence limit on the estimated standard deviation of a given population. Demonstrate, by an example, how this equation is used in the statistical analysis for CHF correlations.

DNC Response The 95% upper confidence limit is calculated using Equation 3.2 of Section 3.3 of of the May 8, 2015 LAR. Equation 3.2 is taken directly from the approved methodology of VEP-NE-2-A. Equation 3.2 of Section 3.3 of Attachment 6 of the May 8, 2015 LAR is the same as Equation 2.4.5 of Section 2.4 of VEP-NE-2-A.

The M/P ratio and standard deviation for VIPRE-DIWRB-2M and VIPRE-D/ABB-NV were taken from Appendices C and D, respectively, of DOM-NAF-2-P-A.

The 95% upper confidence limit calculation for the WRB-2M CHF correlation is shown below. See the response to RAI-4 (SNPB) for further clarification on how this equation is used in development of the SDL.

K(95)=

22q

_(n-1) 2

(\\/2f 1.645) where n = 241 K(95) 2(21) 2 such that K(95) = 1.082

Serial No.16-011 Docket No. 50-423 Attachment, Page 3 of 12 RAI - 3 (SNPB):

Selection of 5-Percent Code Uncertainty Code uncertainty in the statistical design limit (SDL) analysis is said to account for any differences between the licensee's VIPRE-D and other thermal hydraulic codes in which the WRB-2M and ABB-NV data were correlated, and any effect due to the modeling of a full core with a correlation based on bundle test data. Provide justification for the selection of 5% code uncertainty as a conservative value under all types of thermal margin calculations; including full core, half-s ymmetiy, quarter-symmetry and 1/8 symmetry.

DNC Response Section 1.2 of VEP-NE-2-A states that the code uncertainty accounts for 1) the effect of analyzing a full reactor core with a CHF correlation based on bundle data and 2) the potential that the CHF correlation uncertainty is based on data reduction with a different thermal-hydraulic computer code. The full core here does not refer to the DNB model size but instead to the manner in which CHF testing is performed using bundles versus simulating a full reactor core in the CHF test. Simulating a full reactor core for a CHF test would be costly and unnecessary as CHF is a local phenomenon and not a core-wide phenomenon.

The CHF correlation uncertainty is based on data reduction performed using VIPRE-D (M/P standard deviation taken from the applicable Appendix of DOM-NAF-2-P-A) versus using the M/P standard deviation calculated by Westinghouse.

The effect of the size of the thermal-hydraulic model is accounted for in the model uncertainty term. As stated in the May 8, 2015 LAR, Dominion will only use a 1/8 core model for licensing calculations at MPS3 (Section 3.4 of Attachment 6 to the May 8, 2015 LAR) as this was the model used to calculate the SDLs.

In addition, both VEP-NE-2-A (Section 2.5, page 44) and DOM-NAF-2-P-A (Section 5.1) indicate that the 5% uncertainty is a bounding code uncertainty for thermal-hydraulic calculations performed in accordance with these methodologies.

RAI - 4 (SNPB):

Explanation of the Randomization Process Section 3.1 of Attachment 6 states that each minimum departure from nucleate boiling ratio (MDNBR) is randomized by a code/correlation uncertainty factor as described in Reference 1 using the upper 95% confidence limit on the VIPRE-DA/WRB-2M and VIPRE-D/ABB-NV code/correlation pair measured-to-predicted CHF ratio standard deviation. Explain the randomization process to obtain the randomized values that appear in Tables 3.6-3 and 3. 6-4.

Serial No.16-011 Docket No. 50-423 Attachment, Page 4 of 12 DNC Response The values listed in Tables 3.6-3 and 3.6-4 of Attachment 6 to the May 8, 2015 LAR are the standard deviation of the randomized DNBR results for each of the nominal statepoints. The Monte Carlo analysis consists of 2000 DNBR calculations performed around each of the nominal statepoints.

The calculated DNBR results are further randomized to account for the code/correlation pair uncertainty.

This is done by the following equation and is consistent with the methodology of VEP-NE-2-A.

Calculated DNBR Randomized DNBR = [1.0 + sM)"K(95)" Normal Random Number]j Where:

s(M/P) is the standard deviation of the code/correlation MIP database for the CHF correlation under study taken from the applicable Appendix of DOM-NAF-2-P-A.

K(95) is a sample correction factor that depends on the size of the experimental database supporting the correlation, and is calculated based on the equation given in Statistical DNBR Evaluation Methodology.

K(5)=

___[VEP-NE-2-A, page 37]

RAI -5 (SNPB):

Impact of NRC Information Notice (IN)-2014-01, on the DNBR SDL Determination Section 3.5 of Attachment 6 quantifies the code uncertainty that is used in the calculation of SDL of DNBR. However, in Section 3. 6.1, which discusses the impact of IN-2014-O1, the code uncertainty, Fc, is neglected. Explain why the code uncertainty is not used for the calculation shown in Section 3. 6.1.

DNC Response Section 3.6.1 of Attachment 6 of the May 8, 2015 LAR was provided to the NRC staff to demonstrate that the methodology of VEP-NE-2-A does not need to be modified to account for the issue raised in Information Notice IN-2014-01. The development of the SDLs in Section 3.6 of Attachment 6 of the May 8, 2015 LAR includes the code uncertainty. The terms for the code and model uncertainty were not included in Section

Serial No.16-011 Docket No. 50-423 Attachment, Page 5 of 12 3.6.1 to simplify the comparison of the final equation from Section 3.6.1 to Owen's factor.

RAI - 6 (SNPB):

Different Values for the Linear Regression Coefficient Table 3.8-1 and Table 3.8-2 of Attachment 6 list the linear regression coefficient, R2, for the verification of the nominal statepoints for the MPS3 1 7x1 7 RFA-2 fuel with VIPRE-D/WVRB-2M and VIPRE-D/ABB-NV correlations, respectively. There is considerable difference in R2 for the parameters in the two sets for the correlations. Explain the basis for the reliability of the selected nominal statepoints for the two correlations in the context of this significant difference.

DNC Response The difference in the R2 values between the two CHF correlations is expected since the correlations used to evaluate the behavior are based on different equation forms and experimental databases.

The nominal statepoints at which the two CHF correlations are evaluated are also different (Tables 3.6-1 and 3.6-2 of Attachment 6).

Use of different nominal statepoints is expected to contribute to differences in the linear regression analysis. The purpose of the check is to ensure that the selected nominal statepoints do not exhibit a trend that would bias the generation of the SDL. For a given correlation, the R2 values are similar and there is not strong dependence on any one single input condition. Therefore, the statepoints are deemed acceptable.

RAI - 7 (SNPB): Nuclear Safety Advisory Letter (NSAL)-09-5, Revision 1 NSAL-09-5 Revision 1, "Relaxed Axial Offset Control EQ Technical Specifications Actions," identified a technical concern which is applicable to MPS3 SR 4.2.2. 1.2.e.

which may not provide assurance that the non-equilibrium FQ(Z) limiting condition for operation (LCO) limit will not be met between the performance of the required 31 Effective Full Power Day (EFPD) core power distribution surveillances. Westinghouse has determined in NSAL-09-5 that if FQw (Z) is not within the LCO limit following a surveillance performed at >- 75% RTP, the following actions should be administratively implemented with NRC Administrative Letter 9 8-10, in addition to the current Required Actions contained in the plant specific FQ Technical Specifications:

1. Reduce the maximum allowable power by 3% for each 1% FQW(Z) exceeds the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
2. Reduce the power range neutron flux - high trip setpoints >- 1% for each 1% that the maximum allowable power level is reduced within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Serial No.16-011 Docket No. 50-423 Attachment, Page 6 of 12

3. Reduce the overpower A T trip setpoints by >- 1% for each 1% that the maximum allowable power level is reduced within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
4. Perform SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER above the limit of action 1. Note that this action must be completed whenever the FQW(z) limit is not met following a surveillance performed at > 75% RTP.

The licensee has listed the actions in LAR Attachment 2 that shall be taken with Specification 4.2.2. 1.2.c not being satisfied as INSERT 'A". Insert "A", item a, states:

"Within 15 minutes, control the AFD [axial flux difference] to within the new reduced AED limits specified in the COLR [core operating limits report] that restores FQ(Z) to within its limits." Also, item b states: "Reduce the THERMAL POWER by the amount specified in the COLR that restores FQ(Z) to within its limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,"

(1) What is the reason for specifying the reduction in AFD limits and the reduction in the amount of thermal power specified in the COLR rather than in the TSs?

(2) Specify the amounts for the AFD limit reduction and THERMAL POWER reduction that are proposed. Specify the reason for the amounts of reduction if they deviate from the NSAL-09-5 Revision 1 actions listed above.

DNC Response

1) Dominion performed a four cycle analysis to determine if the NSAL-09-5, Revision 1, power reduction and AFD reduction actions cited above were always sufficient to recover the intended FQw(Z) margin at all axial levels and cycle burn ups.

Dominion concluded from this analysis that:

a) In some cases, a greater than 3% reduction in power, for each 1% that FQw(Z) exceeds the limit, is necessary to recapture the desired margin.

b) In some cases, a greater than 1% AFD reduction, for each 1% FQw(Z) margin shortfall, is necessary to recapture the desired margin.

Dominion observed that variations in cycle-specific core design (especially axial design of fuel and burnable absorber arrangement) can impact the size of the core power and AFD reductions necessary to accomplish the desired increase in margin.

As such, Dominion concluded that a reload-specific value for the reduction in both core power and AFD is more appropriate than a fixed value.

Therefore, these values should be stated in the COLR (along with other cycle specific parameters) for each reload cycle.

Serial No.16-011 Docket No. 50-423 Attachment, Page 7 of 12

2) Representative values for AFD and thermal power reduction are provided in Table 3 (Attachment 1, page 11 of 20) of the May 8, 2015 LAR. These values are representative only. Actual values will be calculated on a cycle-specific basis and listed in the COLR.

The amounts of AFD and power reduction may be different than those specified in the NSAL-09-5, Revision 1, actions for the reasons stated in the above response to Part (1).

RAI - 8 (SRXB): Acceptability of WCAP-14441 On page 10 of Attachment 4, the LAR indicated that the Westinghouse BORDER code and the constituent equations discussed in WCAP-14441 are used to verify the cycle-specific boration requirements in the reload safety analysis. Justify the use of the WCAP-14441 methods for the stated purpose in the reload safety analysis.

DNC Response As stated in MPS3 ESAR Section 4.3.2.1, the constituent equations discussed in WCAP-14441 are currently used to verify the cycle-specific boration requirements in the reload safety analysis.

RAI -

13 (SRXB):

Main Steam Line Break RETRAN Benchmark Analysis - Split Vessel Nodal Scheme The licensee proposed to use the reactor vessel nodal scheme in Figure 2-2 in the analysis of the main steam line break (SLB), which is an asymmetric response transient with lower temperature in the core next to the ruptured steam generator (SG) and higher temperature in other side of the core. Figure 2-2 represented the nodal scheme change by adding volume to create a second parallel flow path through the active core from the lower plenum to the upper plenum. The licensee indicated that the nodal diagram can represent RCS loop temperature asymmetries.

Discuss the model of the mixing flow between the cold-and-hot sides of the core and address the adequacy of the flow mixing model for the SLB analysis, when use the nodal scheme in Figure 2-2.

DNC Response The split reactor vessel model described in Attachment 5 of the May 8, 2015 LAR is consistent with Dominion's approved reactor system transient analysis methodology using RETRAN as documented in VEP-FRD-41-P-A. VEP-FRD-41-P-A describes the split reactor vessel model in Appendix 1 Sections 3.3, 5.2.3.4 and 5.2.3.5 and in

Serial No.16-011 Docket No. 50-423 Attachment, Page 8 of 12 Appendix 7 on page A-7-42. The split reactor vessel model and the specification of mixing flow fractions allow the model to simulate conditions from perfect (complete) to imperfect (incomplete) flow mixing. The mixing flow fractions are based on scale model mixing tests performed by Westinghouse (WVCAP-7909).

Consistent with the methodology of VEP-FRD-41-P-A, the assumption of imperfect mixing combined with appropriate azimuthal weighting factors applied to the reactivity

feedback, conservatively models the core kinetics response to a main SLB transient.

RAI -

18 (SRXB):

TS 6.9.1.6.b Reference List -

Addition of NRC-Approved Dominion Methodologies On page 14 of Attachment I to the LAR, the licensee indicated that it proposed to apply the Dominion's analysis methods to MPS3 in the reload design and safety analysis for licensing applications. The methods are documented in the topical reports (TR) as follows:

TR-1I: VEP-FRD-42-A, "Reload Nuclear Design Methodology."

TR-2: VEP-NE-1-A, "Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications."

TR-3: VEP-NE-2-A, "Statistical DNBR Evaluation Methodology."

TR-4: DOM-NAF-2-P-A, "Reactor Core Thermal-Hydraulics Using the V/PRE-D Computer Code."

TR-5: DOM-NAF-1-P-A, "Qualification of the Studsvik Core Management System Reactor Physics Methods for Application to North Anna and Surry Power Stations."

TR-6: VEP-FRD-4 1-P-A, "VEPCO Reactor System Transient Analyses Using the RETRAN Computer Code."

The licensee proposed to add TR-1 through TR-4 to TS 6.9. 1.6.b and leave TR-5 and TR-6 out of the TSs. Each of the applicable referenced TRs in the TSs will be included in the COLR reference list with the complete identification (i.e., report number, title, revision, date, and any supplements) for a specific reload cycle. The lcensee clarified on page 16 that in accordance with guidance of Generic Letter (GL) 88-16, "a methodology in the COLR reference list is intended to satisfy these criteria: (1) it is used to determine core operating limits, and (2) it has been previously approved by the NRC."

Discuss the purposes of the methods documented in TR-5 and TR-6 for use in the reload design and safety analysis and justify the adequacy of not including TR-5 and TR-6 in TS 6.9. 1.6.b (and thus being able to not reference TR-5 and TR-6 in the COLR).

As part of the discussion, address compliance with the above discussed two criteria established in meeting the GL 88-16 guidance for the case without inclusion of TR-5 and TR-6 in TS 6.9. 1.6.b.

Serial No.16-011 Docket No. 50-423 Attachment, Page 9 of 12 DNC Response This response is provided in two parts: 1) a discussion of the adequacy of not including TR-5 and TR-6 to the TS list of Core Operating Limits Report (COLR) references, and

2) clarification of the licensee process for maintaining and submitting analysis methods.

As described in Insert F of Attachment 2 of the May 8, 2015 LAR, TR-1 through TR-4 are used to establish specific core operating limits for parameters contained in the COLR.

In contrast, DOM-NAF-1-P-A and VEP-FRD-41-P-A do not establish core operating limits listed in the COLR.

The methodologies of DOM-NAF-1-P-A and VEP-FRD-41-P-A document and qualify computer code models that are used in the reload analysis process. The reload methodology, VEP-FRD-42-A, documents how the computer codes are used to establish core operating limits. VEP-FRD-42-A references the analysis tools documented in DOM-NAF-1-P-A and VEP-FRD-41-P-A.

The methodologies documented in DOM-NAF-1-P-A and VEP-FRD-41-P-A do not satisfy Criteria 1 of GL 88-16 and are not required to be listed in the COLR reference list. The methodology that establishes the core operating limits is VEP-FRD-42-A, which is proposed to be included in the COLR reference list.

The inclusion of only TR-1 through TR-4 to the MPS3 COLR is consistent with the application of Dominion's approved reload methods to North Anna (TS 5.6.5.b) and Surry (TS 6.2.C).

Additionally, the MPS3 COLR currently does not contain the Westinghouse equivalent references to TR-5 and TR-6.

DOM-NAF-1-P-A documents the use of the Studsvik Core Management System (CMS) core modeling code package, consisting primarily of CASMO-4 and SIMULATE-3 computer codes. DOM-NAF-1-P-A demonstrates the validity and accuracy of the CMS package used at Dominion for core reload design, core follow, and calculation of key core parameters for reload safety analysis.

DOM-NAF-1-P-A documents a

methodology for calculating nuclear reliability factors (NRF) that account for model predictive bias and uncertainty when calculating reload core physics parameters. The methodology for calculating reload core physics parameters resides in Section 2.2 of VEP-FRD-42-A.

The methodology of DOM-NAF-1-P-A was approved by the NRC in a letter dated March 12, 2003.

VEP-FRD-41-P-A documents the use of the general purpose thermal hydraulics RETRAN computer code for transient analyses of complex fluid flow systems.

The RETRAN computer code calculates 1) general system parameters as a function of time, and 2) boundar'1 conditions for input into more detailed calculations of DNBR or other thermal and fuel performance margins. VEP-FRD-41-P-A describes Dominion's reactor system transient analysis models for use with the RETRAN computer code.

The models have been qualified for FSAR Chapter 15 Non-LOCA transient analyses. The

Serial No.16-011 Docket No. 50-423 Attachment, Page 10 of 12 various reactor system component models are described and qualified for their intended applications. Comparisons to plant data and alternate calculations are provided. VEP-FRD-41-P-A also addresses restrictions and limitations and conditions of use imposed by the NRC's generic safety evaluation reports for the RETRAN computer code. The methodology for how RETRAN is used to establish core operating limits is documented in Sections 2.1.3 and 3.3 of VEP-FRD-42-A.

The methodology of VEP-FRD-41-P-A was approved by the NRC in a letter dated April 11, 1985. VEP-FRD-41-P-A has been modified twice since its original approval. The modifications to VEP-FRD-41-P-A were performed in accordance with Dominion's approved process that is described in Section 2.3 of VEP-FRD-42-A and is further discussed in the response to RAIs contained in Appendix B to VEP-FRD-42-A.

The modifications to VEP-FRD-41-P-A were performed in accordance with the provisions of 10 CFR 50.59.

The change management process used by Dominion is discussed below.

Section 2.3 of VEP-FRD-42-A, "Analytical Model and Method Approval Processes,"

indicates several acceptable means by which either analytical models or methods may achieve approved status for use in Dominion's reload methodology.

The following discussion describes Dominion's process for performing maintenance and modifications of NRC-approved methodologies.

This process is applied to the models and methodologies that are used in Dominion's reload design methodology.

The determination of the requirement to submit methodology changes to the NRC for approval prior to application is based on published NRC guidance. Specifically:

  • 10 CFR 50.59, and in particular 10 CFR 50.59c(2)(viii): (2) A licensee shall obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would, (viii) Result in a departure from a method of evaluation described in the ESAR (as updated) used in establishing the design bases or in the safety analyses.
  • NEI 96-07, Revision 1, "Guidelines for 10 CFR 50.59 Evaluations"
  • Generic Letter 83-11, Supplement 1, "Licensee Qualifications for Performing Safety Analyses" Relevant sections of the documents used in the determination process are as follows:

Serial No.16-011 Docket No. 50-423 Attachment, Page 11 of 12

1. Generic Letter 88-16 establishes the concept of reload cycle-dependent operating limits in the TS.

"Generally, the methodology for determining cycle-specific parameter limits is documented in an NRC-approved Topical Report or in a plant-specific submittal.

As a consequence, the NRC review of proposed changes to TS for these limits is primarily limited to confirmation that the updated limits are calculated using an NRC-approved methodology and consistent with all applicable limits of the safety analysis. These changes also allow the NRC staff to trend the values of these limits relative to past experience. This alternative allows continued trending of these limits without the necessity of prior NRC review and approval."

2. NEI 96-07, Rev. 1, as endorsed by RG 1.187, provides guidance for evaluating changes to methods under the provisions of 10 CFR 50.59.

For example, Paragraph 4.3.8.1, states:

4.3.8.1, Guidance for Changing One or More Elements of a Method of Evaluation "The definition of departure... provides licensees with the flexibility to make changes under 10 CFR 50.59 to methods of evaluation whose results are

'conservative' or that are not important with respect to the demonstrations of performance that the analyses provide.

Changes to elements of analysis methods that yield conservative results, or results that are essentially the same, would not be departures from approved methods."

3. Generic Letter 83-11, Supplement 1, provides a method for licensee qualification of analysis methodologies, including those used to establish core operating limits, without formal NRC review and approval:

"The U.S. Nuclear Regulatory Commission (NRC) is issuing this supplement to Generic Letter (GL) 83-11 to notify licensees and applicants of modifications to the Office of Nuclear Reactor Regulation (NRR) practice regarding licensee qualification for performing their own safety analyses. This includes the analytical areas of reload physics design, core thermal-hydraulic analysis, fuel mechanical analysis, transient analysis (non-LOCA), dose analysis, setpoint analysis, containment response analysis, criticality analysis, statistical analysis, and Core Operating Limit Report (COLR) parameter generation. It is expected that recipients will review the information for applicability to their facilities. However, suggestions contained in this supplement to the generic letter are not NRC requirements; therefore, no specific action or written response is required."

"To help shorten the lengthy review and approval process, the NRC has adopted a generic set of guidelines which, if met, would eliminate the need to submit detailed topical reports for NRC review before a licensee could use approved codes and methods. These guidelines are presented in the Attachment to this

Serial No.16-011 Docket No. 50-423 Attachment, Page 12 of 12 Generic Letter. Using this approach, which is consistent with the regulatory basis provided by Criteria II and Ill of Appendix B to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50), the licensee would institute a program (such as training, procedures, and benchmarking) that follows the guidelines, and would notify NRC by letter that it has done this and that the documentation is available for NRC audit."

Dominion's process for maintaining and modifying approved methodologies is consistent with NRC and industry guidance and includes the following elements:

  • Under the provisions of 10 CFR 50.59(c)(2)(viii) and the guidance of NEI 96-07, Rev. 1, Dominion may change NRC-approved codes and methodologies used to establish core operating limits without additional NRC review and approval.
  • Under the provisions of 10 CFR 50.59(c)(2)(viii) and the guidance of Generic Letter 83-11, Supplement 1, Dominion may implement or substitute NRC-approved codes and methodologies for use in establishing core operating limits without additional NRC review and approval of these methods.
  • Dominion concludes that, in updating the list of approved methodologies for establishing core operating limits in the Technical Specifications, licensee affirmation that the changes to the methodologies have been prepared as described by either of the above is adequate to retain the "approved" status for these methods.

Changes to approved methodologies that may not be performed under the provisions of 10 CFR 50.59(c)(2)(viii) would be submitted to the NRC for review and approval prior to their use in Dominion's reload process.

Dominion applies a similar process to the one outlined above when evaluating changes to Dominion's reload methodology for application at the plant. The proposed change to the station design bases is reviewed under the provision of 10 CFR 50.59 using the guidance of NEI 96-07, Rev. 1, as endorsed by Reg. Guide 1.187. If the provisions of 10 CFR 50.59 are satisfied, the change is implemented.

If the provisions of 10 CFR 50.59 are not satisfied, a license amendment request pursuant to 10 CFR 50.90 would be submitted to the NRC for review and approval prior to implementing the proposed change.

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