ML18149A150

Forwards Responses to 860304 Request for Addl Info Re 850919 Rev 1 to VEP-FRD-42, Reload Nuclear Design Methodology, Per 860404 Meeting.Methods Used to Evaluate Parameters for Reload Safety Analysis Consistent w/WCAP-9272

ML18149A150
Person / Time
Site:	Surry, North Anna, 05000000
Issue date:	05/02/1986
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Harold Denton, Rossi C Office of Nuclear Reactor Regulation
References
86-152, NUDOCS 8605120122
Download: ML18149A150 (34)

Text

e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 W. L. STEWART VxcE PRESIDENT NucLEAR OPERATIONS May 2, 1986 Mr. Harold R. Denton, Director Serial No.86-152 Office of Nuclear Reactor Regulation NO/EJL:vlh Attn: Mr. C. E. Rossi, Assistant Director Docket Nos. 50-280 Division of PWR Licensing-A 50-281 U.S. Nuclear Regulatory Commission 50-338 Washington, D.C. 20555 50-339 License Nos. DPR-32 DPR-37 NPF-4 NPF-7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION NORTH ANNA POWER STATION RELOAD DESIGN METHODOLOGY-ADDITIONAL INFORMATION On September 19, 1985 we submitted Revision 1 of the topical report VEP-FRD-42, "Reload Nuclear Design Methodology"_, to you for review and approval. This report describes the current methodology and code models that we use to perform nuclear reload design and safety an'alyses. In a letter dated March 4, 1986 from Mr. Herbert N. Berkow, we received a request for additional information on VEP-FRD-42, Revision 1. On April 4, 1986 we met with members of your staff to review the information that was requested. Enclosed are our reponses to the information request.

Very truly yours, Enclosure

1. Responses to the Information Request on VEP-FRD-42, Revision 1.

e VIRGINIA ELECTRIC AND POWER COMPANY TO Mr. Harold R. Denton cc: Dr. J. Nelson Grace Re~ional Administrator NRG Region II Mr. Lester S. Rubenstein, Director PWR Project Directorate No. 2 Division of PWR Licensing-A Mr. Leon B. Engle NRG North Anna Project Manager PWR Project Directorate No. 2 Division of PWR Licensing-A Mr. Ghandu P. Patel NRG Surry Project Manager PWR Project Directorate No. 2 Division of PWR Licensing-A Mr. Larry King NRG Resident Inspector North Anna Power Station NRG Senior Resident Inspector Surry Power Station Mr. Marv Dunenfeld Senior Nuclear Engineer Reactor Systems Branch Division of PWR Licensing-A

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Enclosure Responses to the Information Request On VEP-FRD-42, Revision 1

e e RESPON.SE TO INFORMATION REQUEST ON VEP-FRD-42 REVISION 1 Question 1: Please indicate and discuss the main areas of difference, if any, between VEP-FRD-42 and the Westinghouse Reload Safety Evaluation Methodology (WCAP-9272). Is the Virginia Power methodology intended to apply as well to reactors and fuei other than those manufactured by Westinghouse?

Response: As stated in VEP-FRD-42, Rev. 1, the methods we use to determine and evaluate the key parameters for a reload safety analysis are consistent with those outlined in WCAP-9272, and Reference 1.

The main area of difference between WCAP-9272 and VEP-FRD-42 Rev. 1 is in the calculation and verification of axial flux difference bands for reactor operation. WCAP-9272 describes the process by which the constant axial offset control method is verified. Our analogous method for verifying a constant axial offset band is described in Reference 2. However as indicated in VEP-FRD-42 Rev. 1, we have developed a relaxed power distribution control (RPDC) mode of operation for our reactors as an alternative to axial offset control. The method of shape verification for RPDC is described in Reference 1.

The methodology described in VEP-FRD-42 Rev. 1 is currently being used for our nuclear units utilizing Westinghouse fuel. The methodology described in VEP-FRD-42 Rev. 1 would be applicable to other reactors of Westinghouse design. However, use of this methodology for non-Westinghouse reactors would have to be justified based on the specific reactor type. The basis for this justification would be submitted for NRC review prior to application. The methodology described in VEP-FRD-42 Rev. 1 would be 1

... /'

e applicable to fuel from other vendors as long as it was determined that the design basis accidents and their associated key parameters remained the same. In the event that we utilize fuel from other vendors, the appropriate documentation will be submitted for NRC review.

Question 2: Are the codes indicated in Section 2. 1 the only codes used by Virginia Power in their reload safety analyses? Have all codes used in the reload safety analyses been reviewed and approved by the N RC?

Response: In general, all of our reload analyses are performed with the codes listed in Section 2.1 of VEP-FRD-42 Rev. 1, which have been reviewed and approved by the NRC. The exceptions to this are as follows:

a) Data handling routines, designed to summarize and/or transfer data between approved codes are not submitted for specific NRC approv~l. They are, however, subject to the same general verification and configuration control requirements as the approved analysis codes.

b) Current evaluations of the dropped rod transient for North Anna (a negative rate trip protected plant) are based on the Westinghouse interim methodology discussed in Reference 3 and as such make use of the Westinghouse LOFTRAN code. LOFTRAN is an NRG-approved code and is routinely used for transient safety analysis of most Westinghouse units. We have a new dropped rod methodology under development which will use the Section 2.1 codes. Qualification documentation for this methodology will be submitted for NRC review.

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c) Our Loss of Coolant Accident (LOCA) capability is also based on NRG-approved Westinghouse codes. For further discussion, see the response to Question 18.

d) For those transients for which we have received NRC approval to apply the Westinghouse statistical DNB methodology (Improved Thermal Design Procedure-ITDP), the Westinghouse THING code, an NRG-approved.core thermal-hydraulics code, is being used for DNB calculations. We have submitted, in Reference 4, a topical report describing an in-house statistical DNBR evalua-tion methodology. Upon approval of this methodology, we plan to use our approved COBRA model for these transients.

Question 3: Have all the initial current limits (as defined in Section 3.3.2) been verified using the Virginia Power methodology? If not, justify the validity of comparing values of key analysis parameters determined by using two distinct methodologies.

Response: As indicated in Section 3.3.2 of VEP-FRD-42 Rev. 1, the significance of the initial current limits lies not in the codes which generated the physics parameter values, but in the fact that they were assumed in the transient analysis. Thus for the transient analysis to remain bounding for a given reload cycle, it must be demonstrated that parameter values for that reload fall within the limits established by the safety analysis. The list of key safety parameters is a "living document",

i. e., whenever a key parameter is violated and a transient reanalysis is required, the list is updated to reflect the new key safety parameter values assumed in the analysis.

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e e The governing consideration for each reload, therefore, is to ensure that the key safety parameter values predicted by the physics analysis are 1) less limiting than the current limits (otherwise a transient reanalysis may be required) and 2) bound the range of values expected for the cycle.

The former constraint is met simply by comparing the cycle specific results to the limits, which in turn are set by the transient analysis assumptions, and not by physics calculations. The second constraint is met by qualifying the physics models against data as described in References 2,5,6, and 7 and by applying the appropriate design calculational uncertainty factors as discussed in the response to Question 7. Thus direct comparison of cycle-specific core physics results to previous results*, whether generated by the same or a different methodology, is not required. For further discussion on comparing analyses generated with different transient models, see the response to Question 5.

Question 4: How are the results of the safety evaluations of reload safety analysis (as described in Section 3 .4) incorporated in the limiting conditions of operation, limiting safety system setpoints, and technical specifications for a reload cycle? For a typical reload which are the limiting conditions of operation, limiting safety system setpoints, and technical specifications that are expected to change?

Response: During the design and safety analysis initialization process for each reload cycle, the Technical Specifications are reviewed to determine whether changes to the limiting conditions of operation and limiting safety system setpoints have occurred since the previous reload which could impact the results of the safety analysis of the reload core.

Similarly, if changes to the limiting conditions for operation or limiting safety system setpoints are found to be required for operation of the 4

reload core, the required changes and the associated technical justification are incorporated into the Reload Safety Evaluation (RSE),

which is reviewed and approved by the our Safety Evaluation and Control staff and the Station Nuclear Safety and Operating Committee. The RSE and associated Technical Specifications changes are then submitted to the NRC for approval prior to startup.

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The reload design and analysis philosophy for our units has been such that changes to the limiting safety system settings have not been requested to support operation of specific reload cores. Those changes which have been requested (e.g., changes to the. core thermal limits and associated overtemperature/overpower delta-T setpoints following steam generator replacement, change of the part-power multiplier on the local enthalpy rise hot channel factor (F-delta-H) limit equation) have been outside the reload licensing process.

Historically, the sections of the Technical Specifications for our nuclear units which have been most likely to change from cycle-to-cycle have been:

a. The control rod bank insertion limits. These limits influence the shutdown reactivity margin, ejected rod worths and peaking factors and normal operation radial peaking. In the past, the normal operation peaking at lower powers has typically driven requests for revisions to the insertion lim.its. Subsequent to these requests, we obtained NRC approval for increasing the normal operation power peaking factor (F-delta-H) limits at part power. As a result reload-related requests to change the rod insertion limits are not anticipated in the future.

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b. The axially-dependent radial peaking factor (Fxy) limits.

For North Anna, current reloads involve reporting of cycle-specific Fxy limits as part of a core surveillance report. Once the Relaxed Power Distribution Control (RPDC) methodology is implemented, the core surveillance report will include cycle-specific axial flux difference (AFD) limits and cycle-specific values for the nonequilibrium power peaking factor discussed in detail in Reference 1.

c. The minimum power for which frequent local power peaking (FQ) surveillance is required (the FQ surveillance threshold power level). In some previous Surry cycles these changes occurred when the reload calculations performed under the ECCS Final Acceptance Criteria showed transient local power peaking factors tFQ) in excess of the limit established by the LOCA/ECCS analyses. This situation was aggravated by the relatively low FQ limits which were established for the units when high degrees of steam generator tube plugging (25% or more) were present.

Following replacement of the steam generators, LOCA reanalyses resulted in an FQ limit which is high enough that frequent FQ surveillance is not anticipated for future reloads.

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e Question 5: Please provide a set of key analysis parameters of each accident included in Table 1. For each accident, and each key analysis parameter, indicate the direction of the variation that would increase the severity of the accident. Specify the criteria that are used to determine whether a given variable is a key analysis parameter for a given accident.

Response: For a complete correlation of key safety parameters with the various accidents as well as limiting directions, the reviewer is referred to WCAP-9272, which has been submitted to the NRC for review.

In order to ensure that the transient analysis results conservatively bound the expected response, the various accidents must be correlated with those parameters which have a significant impact on them and the limiting direction of variation for those parameters. This is done in an integrated manner for each accident, drawing primarily upon physical reasoning and past analytical experience.

Physical reasoning, although not adequate in and of itself to establish the limiting direction for all parameters, provides an important starting point for assessing the impact of various parameters. For example, for cooldown accidents the limiting DNBR is expected to occur for the most negative moderator temperature coefficient, since this maximizes core power. Accidents which reach their most severe condition very rapidly, such as rod ejection, are expected* to be much less sensitive to the moderator coefficient, etc.

Sources of past analytical experience for the various accidents are also drawn from. For many transients, the identification and limiting direct.ion*

of key safety parameters are presented in the units' UFSARs. A more formal 7

e presentation of key safety parameters for*the various accidents along with indication of limiting directions i_s given in WCAP-9272. We have confirmed the applicability of these key parameters and limiting directions for a broad selection of the UFSAR transients using the models and methods referred to in Section 2. 0 of VEP-FRD-42 Rev. 1. The fact that this information is valid for both the Vendor's methodology and our methodology is not coincidence, but reflective of the fact that the basic physical characteristics of the accidents being modeled remain fixed.

This process of confirming the applicability of the WCAP-9272 conclusions to our models and methods involved the execution of numerous sensitivity studies using the RETRAN computer code. As a result, a considerable base of experience has been compiled over several years. Some of the results of these studies have been submitted to the NRG. Reference 8, for example, presented the results of sensitivity studies for the rod ejection event using our methodology. These studies covered variations in eight different neutronics parameters as well *as various thermal hydraulics and power-distribution related factors.

As part of the NRG review of our RETRAN transient analysis topical report (Reference 9), we submitted sensitivity studies (Reference 10) which covered a spectrum of parameters for transients involving reactivity addition, changes in primary to secondary heat transfer (both increase and decrease) and decrease in RCS flow rate. In summarizing those studies, it was concluded that our RETRAN models showed the same general sensitivities as discussed in the Surry and North Anna UFSAR's.

In addition to these externally published sensitivity studies, we have performed and documented internally numerous other studies which have 8

e confirmed the limiting directions of key parameters for those accidents which have been shown historically to be most subject to reanalysis as a result of core reloads.

Question 6: Discuss the scope of the review of design basis information (Section 3.2.1). Indicate the steps that are taken (comparison to checklists, e.g.,) to ensure that the review is complete.

Response: Prior to initiation of the reload design safety analysis, we perform a review of the pertinent design basis information applicable to the reload. Th.is review is governed by our internal procedures, which provide for the review of the following:

1. Procedures controlling the performance of, review of, approval of, and documentation of all production calculations.

2. The unit's previous cycle documentation including;

a. Safety Analysis,

b. Nuclear Design Analysis,

c. Operating History,

d. Design or Technical Specification changes made since startup.

3. Applicable correspondence with vendor and NRG.

4. Applicable sections of the UFSAR.

5. Current Technical Specifications.

6. Technical Specifications changes expected to be made before or during the cycle.

7. Current Safety Related NSSS Parameter Checklist.

8. Current Key Safety Parameters Checklist.

9. Current Fuel Assembly and Insert Component Restricted List.

10. Current Fuel Management Scheme.

To ensure that all applicable information is current and being properly utilized a-series of meetings between the design and safety engineers is held. A design initialization checklist provides the schedule for these meetings and indicates the information reviewed, a summary of any design requirements or restrictions unique to the cycle, and design assumptions to be used for the cycle resulting from operational or licensing 9

e uncertainties. In addition, this checklist provides the necessary operating and energy requirements for the cycle. The checklist is reviewed and approved by management.

The NSSS Parameter Checklist contains the current NSSS parameters (fluid volumes, syst:em flow rates, safety and relief valve capacities, etc.)

necessary for the accident analys,es. This checklist is reviewed prior to use by the safety analysis group, power station personnel, the nuclear operations support group, and other appropriate personnel to ensure the accident analysis input assumptions are consistent with the current system parameters. Note that this is a double check since all approved design changes which potenti~lly impact the safety analyses will have been previously reviewed by Nuclear Engineering in accordance with the appropriate procedures for the design change approval process.

During the reload safety analysis phase the design engineer calculates the parameters identified by the Key Safety Parameters Checklist. This checklist contains the limiting nuclear and thermal hydraulic parameters and the values used in the current safety analyses. If a reload parameter is not bounded by the current limit 'the safety engineer is notified. At the end of the analysis phase for the design engineer, the calculations are documented and the results are formally summarized by memorandum to the safety engineer. This memorandum takes the form of a comparison between the calculated reload values and the limits specified by the Key Safety Parameters Checklist. Prior to the finalization of the design engineer's calculation of the key parameters, the design engineer and safety engineer will have formally reviewed the Key Safety Parameters Checklist at least twice: once prior to the initiation of the calculations to review the limits and the second time at the conclusion of the design 10

calculations to ensure the reload values are bounded by the current limits and/or to identify areas where accident reanalysis or reevaluation is required.

Question 7: Provide the calculational uncertainty and bias in all key analysis parameters, and indicate how these uncertainties are conservatively accounted for in the reload analysis.

Response: As indicated in Table 2 of VEP-FRD-42 Rev. 1 the key analysis parameters to be determined for each reload cycle may be divided into three categories. These are non-specific, specific, and fuel performance and thermal hydraulic related parameters.

a) Non-specific Key Parameters The non-specific* parameters are generated to allow evaluation of the genera*l core characteristics of the reload cycle. These parameters are calculated for conditions bounding those expected to occur during the reload cycle to ensure the limiting values of the parameter are determined.

These conditions include conservative assumptions for such parameters as xenon distributions, power level, control rod position, operating history, and burnup. Table 1 provides a list of the non-specific key parameters generated in the reload safety analysis along with the associated conservatisms used in calculating the parameters.

b) Specific Key Parameters Specific key parameters are generated by statically simulating an accident with the physics models. These parameters are (or are directly related to) rod worth, reactivity insertion rate, or peaking factors. The static conditions for the accident are the most conservative conditions for the 11

accident and include conservative assumptions for variations in such parameters as power level, control. rod position, xenon distribution, previous cycle burnup, and current cycle burnup. Table 2 provides a list of the specific key parameters generated in the reload safety analysis and the conservatisms applied in calculating the parameters.

c) Fuel Performance and Thermal/Hydraulic Parameters The core thermal limits are generated assuming the design uncertainties discussed in Reference 11. In that report, detailed discussions are presented on hot channel factors, including radial and local power factors, channel geometry, local flow starvation, etc. DNB correlation uncertainties are also discussed.

Fuel performance related information, including initial fuel temperatures (both core average and hot spot) the effects of fuel densification on core average power and on local power, fuel rod internal gas pressure, fuel stored energy and decay heat are currently generated by the fuel vendor and transmitted to us for each reload core for use in performing the reload evaluations. The vendor calculations are performed using the latest approved version of the Westinghouse PAD code,* documented in References 12, 13, and 14. All of these documents have been reviewed and approved by the NRC.

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e e Question 8: The concept of bounding analysis proposed in Section 3.1 assumes the existence of a reference cycle safety analysis (e.g. the FSAR) to which the safety analysis of later cycles is compared. Is the methodology of VEP-FRD-42 intended to produce such a reference safety analysis? If so, justify the methods Virginia Power intends using to perform neutronic,

• thermal-hydraulic, fuel performance, transient and accident analysis calculations for the reference safety analysis of a mixed core containing non-Westinghouse fuel.

Response: The methodology presented in VEP-FRD-42, Rev. 1 and its referenced documents is our integrated methodology for ensuring that existing or "reference!' safety analyses bound the conditions realised for reload cores and for producing new reference analyses as required. It should be pointed out that in the development and maintenance of a list of key safety parameters for a given plant, the reference safety analyses performed for various UFSAR chapter 15 accidents may have been originally performed at different points in the history of the unit as various cycle designs have required the reanalysis of different accidents. The key safety parameters list is therefore a "living document" which reflects the most recent analyses of the various events.

In general, we expect th& methods presented in VEP-FRD-42 Rev. 1 to be valid for both Westinghouse/Non-Westinghouse fuel mixes as well as cores designed by other vendors for use in Westinghouse designed plants. The codes that we use are currently being used by other design organizations to analyze a variety of fuel and NSSS types. However, we do not have models at the present time which reflect non-Westinghouse fuel. In the event that we utilize fuel from other vendors, the appropriate documentation will be submitted for NRC review.

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Question 9: The concept of bounding analysis assumes (1) a monotonic dependence of the accident consequence on each key analysis parameter (i.e., the limiting direction of ~he key analysis parameter does not change over the entire range of the key analysis parameter and (2) the effects of the different key analysis parameters on the accident consequence are.

not coupled (i.e., the effect of simultaneous changes in key analysis parameters may be determined by varying those parameters separately).

Since these assumptions are not valid in general, provide the range of validity for each key analysis parameter over which bounding analysis is applicable.

Response: The assumption of monotonic dependence of accident consequences on each key anal~sis parameter (assumption (1) above) can be established for most parameters for most events by the application of physical reasoning. This intuitive approach is backed up by numerous sensitivity studies performed by us and our fuel vendor which confirm the trends for the various key parameters. Expressed in terms of a response surface, this combination of physical reasoning plus sensitivity studies is used to demonstrate with a high level of confidence that for every dependent/independent variable plane, there are no inflection points in the response curve. For those combinations of parameters and events where inflection points have been identified (non-monotonic dependence), a spectrum of values of the parameter in question is assumed in performing the analysis. A case in point is the analysis of the uncontrolled control rod bank withdrawal at power, where a range of reactivity insertion rates, corresponding to varying rod speeds and differential rod worths, is analyzed.

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The second condition (no coupling between the key analysis parameters) is not required for the bounding analysis concept to be valid as long as the first condition (monotonic dependence in every plane) is rigorously satisfied. In other words, even if the effect of simultaneous changes in key parameters cannot be determined by varying the parameters separately, as long as there are no inflection points in the surface and all parameters are within their respective limits, then the analysis results are bounded by the reference analysis. The only thing that cannot be determined is the exact degree of margin between the cycle-specific results and the reference analysis results.

The range of validity for each key parameter is therefore established by the parameter limits themselves. As accidents are reanalyzed to envelop new reload cycles, the limits and therefore the ranges of validity are extended.

Question 10; Describe in detail the simple quantitative evaluation that may be made instead of an actual reanalysis (Section 4.0). If the simple quantitative evaluation makes use of available sensitivities to key analysis parameters, indicate how they are obtained and whether these sensitivities are generic or plant specific. Specify quantitative criteria that need to be satisfied if a simple quantitative evaluation is to be made instead of a complete reanalysis of the accident.

Response: Quantitative evaluation of a small violation of parameter limits may be made in one of several ways. First, if the interplay between the various key safety parameters in determining accident response is well defined, margin in one parameter may be used to offset a small violation 15 I

in another parameter. This process is best defined by presenting some examples:

1) Studies performed by us and others have shown that a key para-meter in determining the severity of the core power response to a rod ejection event is the ejected rod worth in units of dollars (delta k/k ejected rod worth/delayed neutron fraction).

For the case of a cycle-specific violation of the minimum delayed neutron fraction, the safety analyst can take advantage of avail-able cycle-specific margin in ejected rod worth by showing that the ejected worth in dollars is less than the worth assumed in the safety analysis.

2) For small violations in the normalized trip reactivity shape curve, the safety analyst may use the minimum total trip reactivity calculated for the cycle to show that at the points on the curve where the violations occur, the actual integral trip reactivity is greater than that assumed in the safety analysis.

A second method of quantitative evaluation involves using tradeoffs of known sensitivities which have been generated either specifically for our plants using our methods or, in one case, using generic three-loop results generated by Westinghouse using approved methods. The specific case involves generic sensitivities published in Reference 15 which show the trade-off between allowable post-ejection peaking factor (FQ) and ejected rod worths. For some reload cycles where small violations (a few percent) of the FQ limits occur, these studies can be used to show that when there is appreciable margin to the ejected rod worth limit, the reference safety analysis remains bounding. The validity of performing this type of 16

e e evaluation is based on the use of available sensitivities which have been generated with the same transient methodology as. used in performing the reference analysis.

The general philosophy followed in performing an accident evaluation as opposed to a reanalysis is that the analyst must be able to clearly demonstrate that the results of an analysis performed with cycle-specific input would be less severe than the results of the reference analysis.

In other words, in performing the evaluation, no credit is taken for margin between the reference analysis results and the design basis criteria, even though this margin may be substantial. In some cases the analyst and/or reviewer may determine that a cycle specific transient analysis should be performed to verify that the reference analysis remains bounding. No specific quantitative criteria have been established for making this determination, but every instance in which an evaluation (as opposed to a reanalysis) of a key parameter violation is performed must be documented.*

In the documentation the analyst presents *the exact numerical values pertaining to the violation and a detailed discussion of the reasoning and approach used in reaching a conclusion regarding the parameter in question.

This documentation is subject to peer and management review and approval.

The results of these cycle specific evaluations are summarized in the Reload Safety Evaluation.

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e Question 11 : Describe the Virginia Power methodology for determining fuel performance key analysis parameters such .as densification power spike, axial fuel rod shrinkage, fuel rod internal gas pressure, fuel stored energy and decay heat.

Response: Fuel performance key analysis parameters for our reloads are determined by the fuel vendor using their fuel rod performance analysis methodology. Fuel region specific and fuel rod specific input for this analys_is is based on information provided by our reload design models for each reload design. This specific input includes estimates of the achievable fuel region burnups for the reload cycle and the associated individual fuel rod power histories, fluxes and fluences. Results of the fuel rod performance analysis are evaluated by the vendor to insure that fuel performance design criteria will not be violated for the reload design. The results are also provided to us for comparison with fuel performance related key safety parameters.

Question 12: How is a fuel census curve determined for the single RCCA withdrawal accident (Section 3 .3 .4 .2)? What is the conservative shape assumed for the fuel rod census curve?

Response: The single RCCA withdrawal accident is classified as an ANS

• Condition III event. For Condition III events a small amount of fuel failures are allowed. The fuel rod census is performed to determine whether the percentage of rods which may enter into DNB and exceed the thermal limits of the fuel is greater than the amount assumed in the safety analysis.

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The fuel rod census for the single rod withdrawal accident is performed using a 1D-2D synthesis technique to determine the* F~delta-H for each rod in the core. Full core 2-D pin by pin radial power distributions are calculated for both all rods out and D-bank in less a single RCCA. The 1-D model is then used to determine the axial power sharings for the rodded and unrodded planes associated with D-bank to the insertion limits less a single RCCA. Using the power sharings from the 1-D calculations, the pin by pin radial power distributions from the 2-D calculations are.synthesized to obtain F-delta-H's for each pin which are then tabulated to determine the percentage of rods with an F-delta-H greater than the value assumed to produce DNB. This analysis is performed at various times in life to account for cycle burnup effects on the power distribution. In addition axial xenon variations which may occur are accounted for in the 1-D model used for generation of power sharings by artificially skewing the axial power distributions in a conservative manner.

Question 13: Are the delayed neutron fraction, prompt neutron lifetime and Doppler weighting function included in the list of key analysis parameters for the rod ejection accident (3.3.4.3)? If not, justify their exclusion.

Response: The delayed neutron fraction, post ejected rod peaking factor, ejected rod worth and Doppler defect are included in the list of key analysis parameters for the rod ejection accident. The prompt neutron lifetime is not included (see response to Question 14). The Doppler weighting function is not included because it is a generic parameter coupled to the post-ejection maximum power peaking factor. Additional discussion of our methodology for analyzing the rod ejection accident may be found in Reference 8.

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e Question 14: Since the use of a maximum prompt neutron lifetime is not conservative for the analysis of the rod ejection accident, justify the use of a maximum value of the prompt neutron lifetime as the key analysis parameter for transients (Section 3 .3 .3 .5).

Response*: Instead of the maximum value of the prompt neutron lifetime, a generic value is used as described in Reference 8 which describes our rod ejection accident methodology. Analyses described in Reference 8 showed a low sensitivity of the methodology to this parameter and were used to justify the use of a generic value.

Question 15: Discuss the relationship of a conservative trip reactivity shape (Section 3.3.3.3) to axial flux distributions realizable under CAOC and RPDC.

Response: A conservative trip reactivity shape is produced in the reload safety analysis by minimizing the initial worth of the tripped rods through the use of the most bottom peaked axial power distribution realizable under CAOC or RPDC operation. As indicated in VEP-FRD-42 Rev. 1, the trip reactivity shape is produced using thi 1-D NOMAD code (Reference 2). At the most limiting cycle burnup the axial power distribution is artificially skewed to produce an axial flux difference consistent with the negative side of the operating . band (CAOC or RPDC). The trip shape is then generated by determining the reactivity versus position for the trip banks using the above core conditions.

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e Question 16: How does Virginia Power determine the six biases and constants kl through kG, the seven time constants, tl through t7, and the trip reset function f(delta-1) associated with the overpower and overtemperature delta-T trip for a reload cycle? Has Virginia Power established the adequacy of the presently used f(delta-1) function under RPDC?

Response: The coefficients of the steady state overtemperature and overpower delta-T protection equations (kl,k2,k3,k4,k6) as well as the f(delta-I) function are developed in acc0rdance with the methodology of Reference 16. The basis for the current protection equations is verified on a reload basis and would normally remain unchanged. Previous changes to the protection equations have been the result of a modification to the thermal design flowrate, core power uprate, or the implementation of a F-delta-H part power multiplier of 0.3.

The constants in the dynamic term of the overpower delta-T equation (kS and t3) are generic Westinghouse values chosen based on studies documented in Reference 16. The time constants tl,t2 and t4-t7 are selected during the initial plant design phase and are governed by such considerations as compensation for thermal and transport delays in the RCS and in the temperature sensors, and by minimizing the frequency of spurious noise-related trips. These considerations will not change from reload to reload and the time constants are therefore expected to remain fixed. The time constants are reflected in our transient analysis models.

As noted in Section 3.4 of Reference 1, the adequacy of the F(delta-I) function is demonstrated on a cycle-specific basis for RPDC. Evaluations to date have shown that ample margin exists in the existing £(delta-I) 21

function to accomodate the wider range of axial power shapes inherent in RPDC.

Question 17: Provide a discussion of how Virginia Power intends to account for the effects of fuel rod bowing, power spiking and fission gas release in their reload safety analysis.

Response: The effects of fuel rod bow are quantified in Reference 17.

Appropriate penalties derived from this analysis are fully compensated for in our thermal hydraulic analysis methodology by the generic retained DNBR margin for each type of fuel as documented in References 18 and 19.

Power spiking effects are accounted for by the application of a conservative power spiking factor to applicable safety analyses. This factor is calculated by the fuel vendor using their fuel rod performance methodology, (see the response to Question 11 above). The effects of fission gas release is likewise considered by the fuel vendor in their analysis of the fuel rod design for the reload cycle as described in the response to Question 11 above.

Question 18: How does Virginia Power intend re-analyzing the Loss of Coolant Accident (LOCA) for reload cores?

Response: The LOCA key safety parameters for each reload are compared to the current limiting values assumed in the most recent plant specific LOCA analysis to confirm that the limits remain bounding. As discussed previously in the response to Question 4, we attempt to perform any accident reanalysis on a schedule that would allow NRC review and approval well in advance of the Reload Safety Evaluation. For LOCA reanalysis, we 22

e use the Westinghouse LOCA-ECCS evaluation models which have been developed and approved for use with 10CFRSO Appendix K applications. This was discussed with the NRC staff during a May 16, 1984 NRC audit (Reference 20). The NRC staff has previously reviewed and approved several LOCA analyses (References 21 and 22) performed by our staff with Westinghouse codes and methods.

Question 19: How would a set of conservative initial conditions be determined when a reanalysis of a specific accident is necessary?

Response: Most accidents exhibit some small sensitivity .to the initial conditions assumed. For accident evaluation, the initial conditions are obtained by adding or subtracting, as appropriate, maximum steady-state errors to or from rated values. Steady-state errors which are applied are:

a) Core power +2 percent allowance for calorimetric error b) Average RCS +/- 4°F allowance for deadband and measurement temperature error c) Pressurizer +/- 30 psi allowance for operational pressure fluctuations and measurement error In general, errors are chosen in the directions which minimize core margins or margins to other plant design criteria (e.g., overpressure) and are therefore dictated by the type of analysis being performed. Similar to the application of uncertainties to the key analysis. parameters, the limiting directions for the application of errors to the initial conditions are determined based on a) physical reasoning and a conceptual understanding of the type of accident being analyzed; b) vendor experience and insight as documented in the accident analysis chapters of the FSARs and c) in some cases, sensitivity studies performed with our transient analysis models and methods.

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For transients involving increase or decrease in secondary heat removal, the initial steam gene~ator mass has some influence on transient results also. We use a 10% uncertainty about the best estimate initial steam generator mass.

The initial conditions for transient analyses are chosen and justified on a case by case basis. These assumptions are identified in the licensing submittal for each reanalysis.

A discussion of the statistical treatment of errors for certain proposed DNB applications is presented in Reference 4 which is currently awaiting NRG review and approval.

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References:

1. K. L. Basehore, et al., "Vepco Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications", VEP-NE-1-A (March 1986).

2. S. M. Bowman, "The Vepco NOMAD Code and Model",

VEP-NFE-1-A (May 1985).

3. NRC Meeting Report, "Summary of Nov. 19, 1979 Meeting with Westinghouse and Licensees Regarding Dropped Rod Protection,"

January 18, 1980.

4. Letter from W. L. Stewart (Vepco) to H. R. Denton (NRC),

"Statistical DNBR Evaluation Methodology", Serial No.85-688, October 8, 1985.

5. M. L. Smith, "The PDQ07 Discrete Model", VEP-FRD-19A, (July 1981).

6. J. R. Rhodes, "The PDQ07 One Zone Model", VEP-FRD-20A, (July 1981).

7. W. C. Beck, "The Vepco FLAME Model", VEP-FRD-24A, (July 1981).

8. J. G. Miller, J. 0. Erb, "Vepco Evaluation of the Control Rod Ejection Transient", VEP-NFE-2A, (December, 1984).

9. N. A. Smith, "Vepco Reactor System Transient Analysis using the RETRAN Computer Code", VEP-FRD-41A, (May, 1985).

10. Letter from W. L. Stewart (Vepco) to H. R. Denton (NRC),

Serial No. 376, July 12, 1976.

11. F. W. Sliz, "Vepco Reactor Core Thermal -Hydraulic Analysis using the COBRA IIIC/MIT Computer Code", VEP-FRD-33-A, (October, 1983).

12. J. V. Miller, et al., "Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations", WCAP 8720, (October 1976).

13. W. J. Leech, "Improved Analytical Models Used in Westinghouse Westinghouse Fuel Rod Design Computations Application for Transient Analysis", WCAP 8720 Addendum 1, (September 1979).

14. W. J. Leech, et al., "Revised PAD Code Thermal Safety Model",

WCAP 8720 Addendum 2", (October 1982).

15. D. H. Risher, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Reactors using Spatial Kinetics Methods", WCAP 7588 Rev. 1-A, (January, 1975).

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,16. S. L. Ellenberger, et al., "Design Bases for the Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Functions", WCAP 8745, (March, 1977).

17. J. Skarita, et al., "Fuel Rod Bow Evaluation", WCAP 8691 Rev. 1, (July, 1979).

18. Letter from W. L. Stewart (Vepco) to H. R. Denton (NRG),

"Amendment to Operating Licenses NPF-4 and NPF-7 North Anna Power Station Unit Nos. 1 and 2 Proposed Technical Specification Change", Serial No. 731, February 14, 1985.

19. Letter from W. L. Stewart (Vepco) to H. R. Denton (NRG),

"Virginia Power Reduction in Rod Bow DNBR Penalty for Surry Power Station Unit Nos. 1 and 2", Serial No.85-064, March 21, 1985.

20. Letter from J.R. Miller (NRC) to W.L Stewart (Vepco), "NRG Audit for VEPCO Utilization of Westinghouse Computer Codes - Surry 1&2 and North _Anna 1&2", June 19, 1984.

21. Letter from L.B. Engle (NRC) to R.H. Leasburg (Vepco), April 13, 1982.

22. Letter from L.B. Engle (NRC) to R.H. Leasburg (Vepco), January 27, 1983.

23. J. G. Miller, "Vepco Nuclear Design Reliability Factors",

VEP-FRD-45A,- (October, 1982).

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TABLE 1 - NON-SPECIFIC KEY PARAMETERS PARAMETER CONSERVATISM APPLIED C0"4ENTS

1. Moderator Temp. Coef.

a) HZP Surry 2.6 pcm/°F The HZP conservatisms are based North Anna 1.8 pcm/°F on the maximum difference in the b) HFP Surry North Anna 3.0 pcm/°F, 2.2 pcm/°F conservative direction between the measured and predicted values plus a e 1 pcm/°F.bias. The HFP North Anna conservatism is based on the maximum difference in the conservative direction between the measured and predicted values. The most limiting positive and negative coefficients for the cycle are calculated.

2. Doppler Temp. Coef. Calculated for range of conservative

3. Doppler Power Coef. operating conditions including HZP-HFP, rodded conditions, cycle burnups. Range of values used in safety analyses.

4. Delayed Neutron Frac. 1.05 Calculated at conservative cycle burnups and operating conditions to determine limiting values for reload.

TABLE 1 - NON-SPECIFIC KEY PARAMETERS PARAMETER CONSERVATISM APPLIED COMMENTS

5.

• Prbmpt Neutron Lifetime 1.05 Calculated for cycle burnup which provides limiting value. For further discussion see question 14.

6. Boron Worth Accident analysis assumes value of e 16 pcm/ppm. Reload values calculated are in 7-8 pcm/ppm range.

7.. Control Bank Differential 1.10 Conservative xenon and power Worth distributions used in calculation.

8. Shutdown Margin Assumes most reactive rod stuck.

Calculated rod worth is reduced by 10%. Impact of rod insertion limits is accounted for.

Calculation performed for range of cycle burnups to determine most limiting value~

TABLE 1 - NON-SPECIFIC KEY PARAMETERS PARAMETER CONSERVATISM APPLIED COMMENTS

9. Boron Concentration

.Surry 36 ppm Conservatisms applied are based North Anna 18 ppm on maximum difference in the conservative direction between measured and predicted ~alues for HZP. Values calculated for cycle burnup which provides most limiting value.

10. Trip Shape Calculated using cycle burnup, power distribution, and xenon distribution which minimizes initial reactivity insertion.

11. Trip Magnitude Calculated assuming most reactive rod stuck. Calculated rod worth is reduced by 10%. Impact of rod insertion limits for near full power operation is accounted for.

12. Fuel Power Census a) Normal Operation 1.08 Calculated for range of cycle burnups to determine most limiting value. Control bank insertion limits are accounted for.

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TABLE 2 - SPECIFIC KEY PARAMETERS PARAMETER CONSERVATISM APPLIED I. Integral Rod Worth

1. Dropped Rod 1.10 1.10 1. Conservative cycle burnup e

2. Conservative axial power distribution

2. Ejected Rod 1.10 1. No feedback

2. Conservative axial burnup distribution

3. Conservative ~xial xenon distribution II.* Differential Rod Worth 2pcm/step

1. Reactivity Insertion 1.10 1. Applied to maximum value calculated Rate due to Rod 2. Conservative axial xenon distribution Withdrawal 3. 10% translates to 3-10 pcm/step (HFP-HZP)

III. Radial Peaking Factor 1.05

1. Normal Operation 1.08 1. Conservative cycle burnup

2. Conservative versus power level

3. Conservative axial xenon distribution

2. Rod-Misalignment 1.11 1. Conservative cycle burnup

2. Conservative axial xenon distribution

. - ' 1' TABLE 2 - SPECIFIC KEY PARAMETERS PARAMETER CONSERVATISM APPLIED IV. Power Peaking Factor 1.075 2D/3D synthesis

1. 0815 2 lD/2D/3D synthesis

1. FQxP vs Core Height (CAOC) 1. 0815 3 1. Base load operati~n

2. Load follow operation

3. Impact of previous cycle burnup (RPDC) 1. 08153 1. Limiting cycle burnups

2. Conservative axial xenon distributions

3. Large number of axial power distributions analyzed

4. Axial power distributions bound expected operating range

2. Ejected Rod Hot Channel 1.10 1. No.feedback Factor

• 2. Conservative axial xenon distribution

3. Conservative axial burnup distribution .

3. Overpower Peak Kw/ft 1. Conservative axial xenon distribution ~

2. Conservative axial burnup distribution

1. Reference 23

2. Reference 1

3. An engineering uncertainty is also applied

4. An engineering uncertainty and densification spike factor are also applied