CP-200801686, Response to Request for Additional Information Regarding Spent Fuel Pool Criticality License Amendment Request

Response to Request for Additional Information Regarding Spent Fuel Pool Criticality License Amendment Request

ML083570151
Person / Time
Site:	Comanche Peak
Issue date:	12/11/2008
From:	Blevins M Luminant Generation Co, Luminant Power
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CP-200801686, TAC MD8417, TAC MD8418, TXX-08148
Download: ML083570151 (31)

Text

Mike Blevins Luminant Power Executive Vice President P 0 Box 1002 0____1&

'Chief Nuclear Officer 6322 North FM 56 Mike-Blevins@Lurninant.com Glen Rose, TX 76043 T 254 897 5209 C 817 559 9085 F 254 897 6652 CP-200801686 Ref. # 10 CFR 50.90 Log # TXX-08148 December 11, 2008 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION DOCKET NOS. 50-445 AND 50-446 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING SPENT FUEL POOL CRITICALITY LICENSE AMENDMENT REQUEST (TAC NOS. MD8417 AND MD8418)

REFERENCES:

1.

Letter logged TXX-07106 dated August 28, 2007 from Mike Blevins of Luminant Power to the NRC submitting License Amendment Request (LAR)07-004.

2.

Letter logged TXX-08087, dated June 30, 2008, from Mike Blevins of Luminant Power to the NRC submitting a supplement to the Spent Fuel Pool Criticality Analysis.

3.

Letter dated November 19, 2008, from Balwant Singal of NRR to Mr. Blevins.

Dear Sir or Madam:

Per Reference 1, Luminant Generation Company LLC (Luminant Power) requested changes to the Comanche Peak Steam Electric Station, herein referred to as Comanche Peak Nuclear Power Plant (CPNPP), Units 1 and 2 Operating Licenses and to Technical Specification 1.0, "USE AND APPLICATION" to revise rated thermal power from 3458 MWt to 3612 MWt. As part of the request to increase rated thermal power, Luminant Power requested to revise Technical Specifications 3.7.17, "Spent Fuel Assembly Storage," for the spent fuel pool criticality analysis CPNPP Units 1 and 2. In Reference 2,

'Luminant Power supplemented the information supporting the spent fuel pool criticality analysis.

On November 19, 2008, the NRC provided Luminant Power with a request for additional information (Reference 3) regarding the proposed changes to rated thermal power, arranged in two groups with staggered response dates.

The responses to the first set of questions are provided in the attachment to this letter.

In accordance with 10CFR50.91(b), Luminant Power is providing the State of Texas with a copy of this proposed amendment supplement.

AooS A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway. Comanche Peak

• Diablo Canyon

• Palo Verde - South Texas Project. Wolf Creek k-1 tzlzý,

U. S. Nuclear Regulatory Commission TXX-08148 Page 2 December 11, 2008 This communication contains no new or revised commitments. Should you have any questions, please contact Mr. J. D. Seawright at (254) 897-0140.

I state under the penalty of perjury that the foregoing is true and correct.

Executed on December 11, 2008.

Sincerely, Luminant Generation Company LLC Mike Blevins By:

ArKt.

Fred W. Madden Director, Oversight & Regulatory Affairs Attachment -

Response to NRC Request for Additional Information c - E. E. Collins, Region IV B. K. Singal, NRR Resident Inspectors, CPNPP Ms. Alice Rogers Environmental & Consumer Safety Section Texas Department of State Health Services 1100 West 49th Street Austin, Texas 78756-3189

Attachment to TXX-08148 Page 3 of 31 ATTACHMENT TO TXX-08148 Response to NRC Request for Additional Information

Attachment to TXX-08148 Page 4 of 31 Response to NRC Request for Additional Information NRC Question 1:

The current CPSES, Units 1 and 2 TS 4.3.1.1 requires "keff < 1.0 when fully flooded with unborated water which includes an allowance for uncertainties as described in Section 4.3 of the FSAR [Final Safety Analysis Report]" and "keff 5 0.95 if fully flooded with water borated to 800 ppm [parts per million], which includes an allowance for uncertainties as described in Section 4.3 of the FSAR." It appears that the proposed change to TS 3.7.17 (Reference 1) would create a technical discrepancy with TS 4.3.1. Please provide clarification and the revised TS pages to ensure that there is no discrepancy, if you agree.

CPNPP Response:

Luminant Power agrees with the NRC. A revised markup for TS 4.3.1 will be provided with the responses to the NRC's second group of requests for additional information.

NRC Question 2:

Deleted NRC Question 3:

The proposed change to TS 3.7.17 and WCAP-16828-P does not appear to uniquely consider the "4-out-of-4 with Axial Blankets". and "3-out-of-4 with Axial Blankets" storage configurations in the interface portion of the analysis. The analysis appears to assume that they are identical to the "4-out-of-4" and "3-out-of-4" storage configurations. Yet the reactivity at the interface of these storage configurations will be different than the '4-out-of-4" and "3-out-of-4" interface with the storage configurations. What is the reactivity effect of explicitly modeling the "4-out-of-4 with Axial Blankets" and "3-out-of-4 with Axial Blankets" storage configurations in interface situations?

CPNPP Response:

The analysis presented in WCAP-16827-P assumes that the reactivity of the blanketed assembly configurations will be bounded by that of the unblanketed assemblies. This assumption is based on the lower fissile material content and, specifically for configuration interfaces, the fact that the blankets will dramatically reduce the impact of the reactivity end-effect.

The reactivities of all configurations interfaced with the "4-out-of-4 with Axial Blankets" and "3-out-of-4 with Axial Blankets" storage configurations were examined and the results are provided below. The KENO full pool model was modified to contain an interface of two different configurations. The keff determined from this model is compared to the target keff of the more reactive configuration. The interface is acceptable if the reactivity of the more reactive configuration is not increased by the presence of the other configuration. The results show that the interface reactivity is less than the limiting target keff of the configurations on both sides of the interface.

Typically the maximum storable fresh enrichment for each configuration is used to model the interface condition. The maximum fresh enrichment is not considered in WCAP-16827-P for blanketed fuel because fresh fuel is not allowed to be stored in these configurations. The calculations performed here use the set of isotopics near the bumup limit to model the blanketed configuration. This minimizes the difference between the bumup limit and the burnup of the modeled assemblies. The assemblies modeled in these calculations contain 3 w/o enrichment

Attachment to TXX-08148, Page 5 of 31 with 40,000 MWd/MTU of bumup for the "4-out-of-4 with Axial Blankets" storage configuration and 5 w/o enrichment with 45,000 MWd/MTU of bumup for the "3-out-of-4 with Axial Blankets" storage configuration.

Results of Explicit Interface Simulations "4-out-of-4 with Axial Blankets" "3-.out-of-4 with Axial Blankets" Configuration keff Limiting keff Limiting Target keff Target keff "4-out-of-4" 0.96540 +/- 0.00025 0.97504 0.96871 +/- 0.00025 0.97560 "3-out-of-4" 0.96981 +/- 0.00037 0.97711 0.96934 +/- 0.00036 0.97711 "2-out-of-4" 0.97174+/- 0.00040 0.97538 0.96744 +/- 0.00055 0.97560 "4-out-of-4 with 0'.96669 +/- 0.00033 0.97504 0.96775 +/- 0.00028 0.97560 1 RCCA"

"'4-out-of-4 with 0.97225 +/- 0.00031 0.97778 0.97079 +/- 0.00036 0.97778 2 RCCAs" "4-out-of-4 with 2 RackSavers 0.96931 +/- 0.00035 0.97568 0.96656 +/- 0.00030 0.97568 and Axial Blankets" "4-out-of-4 with 3 RackSavers 0.97070 +/- 0.00035 0.97823 0.96985 +/- 0.00034 0.97823 and Axial Blankets" NRC Question 4:

Deleted NRC Question 5:

Deleted NRC Question 6:

Several of the figures in the proposed change to TS 3.7.17 (Reference 1) indicate a burnup well in excess of the current maximum licensed burnup. It appears that for some of the figures in the proposed change to TS 3.7.17, simulations with a burnup in excess of the current maximum licensed burnup were used to derive the figure (even though the figure itself does not indicate a burnup in excess of the current maximum licensed burnup). Please explain the following:

a)

The reason for its acceptability.

b) The basis and justification for using the computer codes well in excess of the current maximum licensed bumup.

Attachment to TXX-08148 Page 6 of 31 CPNPP Response-Part a)

The 62,000 MWd/MTU bumup limit is intended to ensure in-reactor fuel integrity and is driven by the performance of the materials used in the fuel assemblies. It is not related to the neutronic performance of the fuel itself. The large limits presented for some cases in TS 3.7.17 represent the burnup a fuel assembly must attain to be stored in that configuration, but are not meant to imply that such high burnup is acceptable for assemblies at Comanche Peak. These burnup limits, though not achievable, are still required to allow for interpolation to determine the burnup limits for intermediate enrichments.

Part b)

PHOENIX-P employs a standard depletion method which tracks nuclides individually via a series of coupled depletion equations. These equations are valid over a wide range of burnups, regardless of the material performance that limits allowable fuel assembly average burnup.

None of the inputs, correlations, models, or equations used in PHOENIX-P have any explicit maximum burnup. Furthermore, there is no burnup range specified in the SER associated with WCAP-11596-P-A, Reference 1. There is no evidence to date that PHOENIX-P either incorrectly calculates flux or contains faulty decay constants which would lead to an error in the depletion calculations in excess of that used in WCAP-16827-P. Further justification for the depletion uncertainty Will be provided in response to Question 28.

NRC Question 7:

241 The current TS Figure 3.7.17-1 indicates that a 5.0 w/o enriched fuel assembly with zero PU decay time requires approximately 63 gigawatt days per metric ton unit (GWD/MTU) of burnup to be stored in a "4-out-of-4" storage configuration in Region II. The proposed TS Figure 3.7.17-1 241 indicates that a 5.0 w/o enriched fuel assembly with zero PU decay time requires approximately 76 GWD/MTU of bumup to be stored in a "4-out-of-4" storage configuration in Region II. Please explain the reason for this large difference.

CPNPP Response:

An increase in burnup requirements is an expected outcome of this analysis. The Comanche Peak stretch power uprate (SPU) increases several parameters important to determining discharged fuel assembly reactivity, such as moderator and fuel temperatures. Also, more conservative methods, especially regarding the treatment of the assembly axial burnup profile, have been adopted since the last analysis was completed about seven years ago. The major contribution to the difference between the required burnup in the current TS and the proposed TS is the difference in the axial burnup profile assumptions used in the development of the burnup credit.

For the current TS, calculations were performed using an axial bumup profile from a representative Comanche Peak reactor core design. This axial bumup profile was developed conservatively by assuming the plant was operating with D bank inserted to 200 steps withdrawn throughout several cycles of depletion and the most limiting assembly at a given bumup (one with lowest burnup at the top) was selected for the burnup credit calculations.

For the proposed TS, calculations were performed using an axial bumup profile based on Profile 1 presented in Reference 2. This axial burnup profile is significantly more limiting than the one assumed in the current TS.

Attachment to TXX-08148 Page 7 of 31 The end effect reactivity difference at a burnup of 60 GWD/MTU in the "4-out-of-4" storage configuration in Region II in the current TS and in the "4-out-of-4" storage configuration in Region II in the proposed TS is about 0.065 Akeff. Based on the data presented in Table 4-17 of WCAP-16827-P, the reactivity difference in the two axial burnup profiles causes a burnup requirement difference of approximately 10 - 15 GWD/MTU.

The proposed TS also integrates the burnup measurement uncertainty into the sum of biases and uncertainties at unborated conditions. The worth of this uncertainty at 5 w/o is 0.01412 Akeff, which after convolution with the other uncertainties increases the total sum of biases and uncertainties 0.00785 Akeff. The impact of this additional uncertainty is an increase of approximately 1 to 2 GWd/MTU.

There are several other differences in the assumptions made during the criticality analyses for the current TS and the proposed TS. Each assumption listed in the table below contributes to the difference in the required burnup but is less significant than the axial burnup profile.

Assumption Treatment in Current Treatment in Proposed Analysis Analysis Maximum Kff 1.0 0.995 (0.005 administrative margin)

U02 density/tolerance 95.5/+ 2%

97.5/+0%

Pellet Dishing Convoluted in total uncert.

No dishing Grid/Sleeve credit Yes No Reactor Power for Non-uprate Uprate depletion Boron letdown for Typical reload core Constant 1000 ppm depletion Rack cell ID tolerance 0.025 inch 0.05 inch NRC Question 8:

The current and proposed TS figures for the "3-out-of-4" storage configuration in Region I1 show an increase in required burnup. Please explain the reason for this large difference.

CPNPP Response:

As discussed in the response to Question 7, an increase in burnup limits was expected as an outcome of this analysis. The use of the more limiting axial burnup profile also impacts the "3-out-of-4" configuration.

241 The currentTS Figure 3.7.17-2 indicates that a5.0 w/o enriched fuel assembly with zero Pu decay time requires approximately 43 GWD/MTU of burnup to be stored in a "3-out-of-4" storage configuration in Region I1. The proposed TS figures for the "3-out-of-4" storage configuration indicate that a 5.0 w/o enriched fuel assembly with zero 241Pu decay time requires approximately 47 GWD/MTU of burnup to be stored in a "3-out-of-4" storage configuration in Region II. Therefore, the difference in the required burnup is about 4 GWD/MTU.

The major contribution to the difference between the required burnup in the current TS and the proposed TS for the "3-out-of-4" storage configuration in Region II is the difference in the axial burnup profile treatment, as described in response to Question 7. The difference in end effect reactivity at 45 GWD/MTU in the "3-out-of-4" storage configuration between the current and proposed TS is approximately 0.026 Akeff. Based on the data presented in Table 4-23 of WCAP-

Attachment to TXX-08148 Page 8 of 31 16827-P, the reactivity difference in the two axial burnup profiles causes a bumup requirement difference of less than 5 GWD/MTU.

The proposed TS also integratesthe burnup measurement uncertainty into the sum of biases and uncertainties at unborated conditions. The worth of this uncertainty at 5 w/o is 0.01022 Akeff, which after convolution with the other uncertainties increases the total sum of biases and uncertainties 0.00465 Akeff. The impact of this additional uncertainty is an increase of approximately 1 GWd/MTU.

There are other differences in the assumptions for the criticality analyses for the current TS and the proposed TS for the "3-out-of-4" storage configuration in Region II, as discussed in the response to Question 7. Each of the different assumptions contributes to the difference in the required burnup but is less significant than the axial burnup profile.

NRC Question 9:

The current and proposed TS figures for the "2-out-of-4" storage configuration in Region II show a significant decrease in required burnup. Please explain what has changed to warrant this large difference.

CPNPP Response:

A decrease in burnup limits is expected for the "2-out-of-4" storage configuration based on the use of soluble boron credit in the proposed TS. The current TS do not credit the presence of soluble boron, and must therefore meet a lower keff limit through the use of burnup credit alone.

Both the current and proposed TS use the same uniform axial burnup treatment.

The current TS Figure 3.7.17-3 indicates that a 5.0 w/o enriched fuel assembly requires approximately i7 GWD/MTU of bumup to be stored in a "2-out-of-4" storage configuration in Region II. The proposed TS figure for the "2-out-of-4" storage configuration indicates that a 5.0 w/o enriched fuel assembly requires approximately 7 GWD/MTU of bumup to be stored in a "2-out-of-4" storage configuration in Region II.

First, there is no end effect for fuel assemblies with low burnup (less than about 10 GWd/MTU).

There is, therefore, no difference in the axial burnup profile treatment between the current and proposed TS. Furthermore, there is little interaction between fuel assemblies in the "2-out-of-4" storage configuration due to the presence of empty cells between assemblies.

No boron credit was used in the development of the current "2-out-of-4" storage TS. The calculations were performed with a maximum kff of 0.95. Soluble boron credit is assumed in the development of the proposed "2-out-of-4" storage TS; therefore the calculations are performed with a maximum keff of 1.0. This is the major contribution to the difference between the required burnup in the current TS and the proposed TS. Based on the data presented in Table 4-25 of WCAP-16827-P, 0.05 Akeff causes a burnup requirement difference of approximately 6 GWd/MTU. Note that for fuel assemblies with no burnup, the maximum allowable enrichment for the current TS is 2.90 w/o compared to the maximum allowable enrichment for the proposed TS of 3.67 w/o. At no burnup, the difference in enrichment is almost 0.80 w/o and is due mainly to the difference in the maximum keff mentioned above.

There are other differences in the assumptions for the criticality analyses for the current TS and the proposed TS for the "2-out-of-4" storage configuration in Region II, as discussed in the response to Question 7. Each of the different assumptions contributes to the difference in the required burnup but is less significant than the different keff criteria. Because of the low burnup requirement in the new analysis, the impacts of the different assumptions identified in Question 7 are less than for the "4-out-of-4" storage configuration.

Attachment to TXX-08148 Page 9 of 31 NRC Question 10:

Response due in January 2009.

NRC Question 11:

WCAP-16827-P uses four nodes to represent the axial burnup profile in most storage configurations. The storage configurations that contain an RCCA are modeled with seven nodes.

Note this review does not consider the storage configurations with RCCAs, as they are not part of the licensee's request. NUREG/CR-6801 does not specify the number of nodes to be used, but indicates 10 is too few and more than 18 is not necessary. However, NUREG/CR-6801 uses nodes of uniform size Whereas WCAP-16827-P uses three small nodes to represent the top of the fuel assembly and one large node to represent the rest. During the April 24, 2008, teleconference, the licensee indicated the supplement would provide information to justify the four-node model.

WCAP-16827, Addendum I does contain a discussion on nodalization, however, it is uncertain how this discussion justifies the four-node model. Therefore, the staff requests the following information.

a) WCAP-16827, Addendum 1, Figures 3-10 through 3-18 appear to be comparisons of CPSES, Units 1 and 2 core simulator axial bumup profiles and a WCAP-16827-P profile.

Please explain the background behind the core simulator axial burnup profiles. Please also explain why they are appropriate for comparison and what they represent.

b) Please explain why the figures in WCAP-16827, Addendum 1, while attempting to justify the four-node model, are actually compared to a seven-node model.

C) Please explain how the comparisons in those figures show that the four-node model is able to model the reactivity with sufficient precision and how this comparison would change with different axial bumup profiles.

The staff requests the licensee provide quantitative evidence that the four-zone nodalization adequately captures the "end effect" vis-a-vis a more detailed nodalization.

CPNPP Response:

Part a)

The core simulator axial bumup profiles are end of cycle bumup profiles for each unique location in the core. These data represent the actual discharged burnup shapes generated at COMANCHE PEAK Units 1 and 2 for the past several cycles. These profiles are important for several reasons.

As discussed in Reference 2, the reactivity "end-effect" is driven by the reduced burnup near the ends of the assembly. The magnitude of the end-effect is therefore inversely correlated to the burnup profile near the ends of the assembly: The most limiting profile can be found by determining the axial bumup profile with the lowest bumup and corresponding highest end-effect.

A series of calculations were performed to demonstrate this correlation. Three seven zone profiles were generated to represent a very steep burnup gradient, a moderate bumup gradient, and a shallow bumup gradient. The profiles used are given in the table below. These seven zone profiles were then depleted at Comanche Peak Units 1 and 2 bounding stretch power uprate (SPU) conditions using the PARAGON lattice physics code. The discharged assembly keff is calculated at a range of burnups, as presented in the table below, and quite clearly shows the relationship between low burnup at the end of the assembly and high assembly. reactivity.

Attachment to TXX-08148 Page 10 of 31 A review of the core simulator burnup profiles also indicates that the axial burnup profiles are similar for all assemblies and that the profiles are smooth and well behaved. No outlying profiles are noted for the past several cycles of COMANCHE PEAK operation. The profiles are also largely similar across both units for all recent cycles. This observation provides confidence that the burnup profiles generated in the future will be similar to those generated in the recent past.

The final important characteristic to be derived from the figures presented is that the burnup gradient at the top end of the assembly has been captured by the profile assumed. In all cases, the figures show that the top burnup gradient starts at approximately 300 cm from the bottom of the assembly. The four zone model uses eight-inch zones at the top of the assembly, and the gradient is therefore modeled starting at 304.8 cm from the bottom of the assembly. The figures therefore demonstrate that additional nodes near the top of the assembly would not increase the calculated reactivity.

Hypothetical Burnup Profiles Zone Midpoint Shape I Shape 2 Shape 3 (in)

(Steep Gradient)

(Moderate Gradient)

(Shallow Gradient)

Relative Burnup 140 0.22 0.45 0.7 132 0.575 0.74 0.9 124 0.87 0.955 1.04 72 1.2225 1.1425 1.06 20 0.87 0.955 1.04 12 0.575 0.74 0.9 4

0.22 0.45 0.7 Comparison of Reactivities of Hypothetical Bumup Profiles Burnup (MWd/MTU)

Shape 1 Shape 2 Shape 3 (Steep Gradient)

(Moderate Gradient)

(Shallow Gradient) 55,000 1.16778 +/- 0.00017 1.09995 +/- 0.00018 1.04154 +/- 0.00018 65,000 1.14089 +/- 0.00016 1.06068 +/- 0.00015 0.99105 +/- 0.00018 75,000 1.11640 +/- 0.00020 1.02442 +/- 0.00015 0.94580 +/- 0.00017 Part b)

The seven zone model is used to capture the burnup gradient at the bottom of the core. The burnup will always be higher at the bottom of the core than the top due to higher moderator

• density induced by the moderator temperature gradient. The lower moderator density at the top of the fuel assembly will also cause a harder neutron spectrum which in turn causes more plutonium production. The combination of lower burnup and more plutonium at the top of the assembly will cause a greater end-effect at the top. The quantitative evidence of this is provided in the response to Part c).

Part c)

A series of calculations were performed to demonstrate that the four zone model is sufficient for accurate reactivity predictions. The three hypothetical assembly burnup profiles from the response to part a) were collapsed to four zones. The three bottom nodes were combined with the middle node and the depletion of this node was used to normalize the assembly burnup. The

Attachment to TXX-08148 Page 11 of 31 results of these calculations are given below. The results confirm that the reactivity is driven by the top end of the assembly, even when a symmetric bumup profile is assumed. The difference would be larger for a real burnup profile with higher bumup at the bottom end.

Comparison of Four Zone and Seven Zone Burnup Profiles Bumup 4-zone profile 7-zone profile Reactivity Difference Shape (MWd/MTU)

(keff - a)

(keff +

)

(Akeff - O) 55,000 1.16836 +/- 0.00017 1.16778 +/- 0.00017 0.00058 +/- 0.00024 (Steep) 65,000 1.14130 +/- 0.00017 1.14089 +/- 0.000i6 0.00041 +/- 0.00023 75,000 1.11655 +/- 0.00018 1.11640 +/- 0.00020 0.00015 +/- 0.00027 2

55,000 1.10076 +/- 0.00018 1.09995 +/- 0.00018 0.00081 +/- 0.00025 65,000 1:06112 +/- 0.00018 1.06068 +/- 0.00015 0.00044 +/- 0.00023 75,000 1.02446 +/- 0.00016 1.02442 +/- 0.00015 0.00004 +/- 0.00022 55,000 1.04266 +/- 0.00017 1.04154 +/- 0.00018 0.00112 +/- 0.00025 (Shallow) 65,000 0.99158 +/- 0.00020 0.99105 +/- 0.00018 0.00053 +/- 0.00027 75,000 0.94611 +/- 0.00019 0.94580 +/- 0.00017 0.00031 +/- 0.00025 NRC Question 12:

Response due in January 2009.

NRC Question 13:

WCAP-16827-P characterized the core operating parameters it used as "...conservative temperature profiles for Comanche Peak Units 1 and 2 at uprated conditions. The use of uprated conditions for depletion calculations with increased power, moderator temperatures and fuel temperatures -lead to increased reactivity determinations at any given burnup relative to fuel irradiated in the core prior to the uprate. The fuel temperatures for each axial zone are calculated based on a representative fuel temperature correlation while the moderator temperatures are based on a linear relationship with axial position." The staff finds this analysis insufficient for the following reasons:

a)

When the staff compared the core exit temperature used in WCAP-16827-P with the core exit temperature information contained in the COMANCHE PEAK, Units 1 and 2 stretch power uprate (SPU) information in WCAP-16840-P Table 2.8.3-1 (Enclosure 1 to Reference 1), it appears that WCAP-16827-P was using the nominal post-SPU core exit temperature rather than a conservative post-SPU core exit temperature. A conservative temperature profile would require the use of the maximum core exit temperature as indicated by NUREG/CR-6665, "Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel," (Reference 7). While WCAP-16840-P Table 2.8.3-1 indicates the post-SPU nominal core exit temperature exceeds the pre-SPU nominal core exit temperature, there is no indication that the post-SPU nominal temperature exceeds the pre-SPU maximum core exit temperature. During the April 24, 2008, teleconference, the staff stated that use of the nominal post-SPU core exit temperature was not considered conservative. WCAP-16827, Addendum 1 contains a section on core depletion. However, the new information provided is not significantly different from the information in WCAP-16827-P. NUREG/CR-6665 estimates the reactivity effect of the depletion moderator temperature to be 90 pcm/degree Kelvin. A 10 degrees Fahrenheit (fF) difference between the nominal core exit temperature and maximum core exit

Attachment to TXX-08148 Page 12 of 31 temperature is likely to result in an approximate increase in keff of 0.0045, essentially enough to eliminate the 0.005 Llkeff analytical margin reserved in WCAP-16827-P.

Therefore, the staff finds the potential magnitude of the non-conservatism to be such that it would preclude a reasonable assurance conclusion that the licensee meets the regulatory requirements in 10 CFR 50.68.

b) The use of the uprated power is not conservative with respect to pre-uprate assemblies.

According to NUREG/CR-6665, "Calculations with both actinide and fission product credit show a trend for conservative prediction of fuel reactivity worth when fuel is burned at lower specific power for a longer period of time for a given burnup. The magnitude of the conservatism increases with increasing burnup." Therefore, the use of the higher specific power associated with the uprate would be non-conservative with respect to the effect power history has on the assembly's Final reactivity. Therefore, the staff requests the licensee determine the effect of using appropriate power, moderator/fuel temperature on all storage configurations, and that the burnup/enrichment loading curves be adjusted accordingly.

CPNPP Response:

Part a)

The core outlet temperature used in WCAP-16827-P, 623.8 'F, is the same as that used in the uprate licensing report, WCAP-16840-P (Reference 3). This outlet temperature is based on calculations performed as described in Section 1.1 of WCAP-16840-P. The analyses document the use of a range of vessel average moderator temperatures forming the licensing basis for plant operation at uprated (3612 MWt) conditions. In addition, the reactor coolant thermal design flow of 95,700 gpm/loop is used. The thermal design flow is less than the minimum measured flow required for power operation. The use of thislower flow conservatively increases the temperature rise across the core, and thus assures the determination of a maximum core outlet temperature. This bounding outlet temperature is used as the analysis temperature, as noted in WCAP-16840-P Table 2.8.3-1, because it is the maximum outlet temperature that the plant can achieve with thermal design flow from the highest allowed vessel average moderator temperature (e.g., T-Avg). In actual plant operations, the reactor coolant flow rate will be higher than the thermal design flow, yielding a smaller temperature rise and a lower-outlet temperature.

It should also be noted from WCAP-16840-P, Table 2.8.3-i that the difference in the outlet temperature.at uprated conditions is only 1.3 'FR The corresponding increase in discharged assembly reactivity based on sensitivities provided in NUREG-6665 (Reference 4) is approximately 0.00025 to 0.00065 Akeff. This increase has been accounted for explicitly by the use of the higher outlet temperatures in the analysis presented in WCAP-16827-P.

Part b)

The depletion calculations performed as part of the analysis presented in WCAP-16827-P will conservatively predict the reactivity of pre-SPU fuel. While it is likely true that the use of uprated power is slightly non-conservative, this impact will be small and is compensated for by

• the higher fuel temperatures, and correspondingly increased discharge assembly reactivity. This, when combined with the higher moderator temperatures discussed above, provides ample margin for pre-uprate fuel considered at uprated conditions. The specific power in the top two nodes of the distributed burnup profile, which drive overall assembly reactivity, are approximately 23 and 37 MW/MTU. These specific power levels are close to the peak specific power indicated in Reference 4, indicating that increased power level will have little effect.

Attachment to TXX-08148 Page 13 of 31 NRC Question 14:

What are the CPSES, Units 1 and 2 cycle-average soluble boron concentrations?

CPNPP Response; Based on design data for current operating cycles:

UIC14 760 ppm U2Cll 818 ppm Based on the average of measured daily samples for past 2 cycles on each unit:

UIC13: 817 ppm*

U1C12: 733 ppm U2C1O: 841 ppm U2C09: 825 ppm NRC Question 15:

In addition to power, moderator/fuel temperature, and soluble boron concentration, the licensee is also requested to address the other core depletion parameters indicated in NUREG/CR-6665.

CPNPP Response:

NUREG/CR-6665, Reference 4, identifies several depletion parameters which are important to a conservative determination of discharged fuel assembly reactivity. The three depletion parameters of importance which are not excluded for this question are: operating history, specific power, and fixed/integral burnable absorbers. Each of these parameters has been considered in the selection of the depletion parameters used in WCAP-16827-P such that a conservative determination of discharge assembly reactivity is made.

Reference 4 indicates both specific power and operating history effects as weakly correlated to increased reactivity for discharged fuel assemblies. The maximum impact noted in Reference 4 is approximately 0.00200 Ak~ff. This result is caused by reduced power operation near the end of assembly depletion. The depletion calculations supporting the analysis presented in WCAP-16827-P do not include part power operation. Instead, the soluble boron concentration is maintained at a constant value above the cycle average value for the entire depletion. The spectral hardening from the presence of boron, especially at the end of the cycle when the concentration is several hundred ppm above physical values, provides additional margin to account for this potential impact. The cycle average boron concentration for Unit I Cycle 14 is 760 ppm (see response to Question 14). The constant value used in the depletion calculations supporting the analyses reported in WCAP-16827-P was 1000 ppm. The reactivity sensitivity for soluble boron concentration reported in Reference 4 is 3 - 3.5 pcm/ppm, so the conservatism of more than 200 ppm is adequate to provide more than 0.00200 Akeff of additional margin. The use of additional margin is the approach suggested in Reference 4 for accounting for the potential for operating history effects.

Comanche Peak fuel management uses both fixed wet annular burnable absorbers (WABA) and integral fuel burnable absorbers (IFBA), so the investigation involved the use of A120 3-B4C WABA and ZrB2 IFBA. WABA rodlets are fabricated into assemblies and inserted into the fuel assembly prior to the fuel assembly being loaded into the core. The WABA rodlets are located inside the fuel assembly guide tubes during core operation. The IFBA is applied to the outer surface of the fuel pellets prior to fuel rod loading.

Attachment to TXX-08148 Page 14 of 31 The largest number of WABA rodlets typically used in a fuel assembly in Comanche Peak fuel management is 16. This number was conservatively maximized at 24 for this investigation to increase the spectral hardening effects caused by the WABA. Because there are 24 guide tubes in a 17x17 fuel assembly, 24 WABA is the maximum number that can be used. The WABA rodlets are only included in the depletion calculations for the bottom three depletion regions because the WABA rodlets extend from 12" to 132" of core height. The top node, which is located above 136" core height, is above the top of the rodlets. The WABA are removed from the fuel assembly after the first cycle the assembly is in the core because Comanche Peak fuel management does not use WABA in previously depleted assemblies. The largest number of IFBA rods typically used in a fuel assembly in Comanche Peak fuel management is 128. This number was conservatively increased to 156 for this investigation as it maximizes the spectral hardening effects of the presence of IFBA. A typical 156 IFBA pattern was used both in fuel depletion calculations and in KENO spent fuel models. A nominal 1.5X IFBA loading was used as it is representative of uprated Comanche Peak operation, and also maximizes the amount of BA in the assembly. The residual IFBA material present as a function of burnup is included in the KENO model. The 4-zone distributed burnup model was used with the same relative burnups as documented in WCAP-16827-P. The depletion calculations were performed at Comanche Peak uprated conditions for power level and moderator temperatures (the same conditions as in WCAP-16827).

The soluble boron concentration present during core operation was modified from a constant 1000 ppm to a more accurate, yet still conservative, model. The cycle average boron concentrations in the response to Question 14 were considered and a conservatively high value of 865 ppm was chosen. The predicted boron concentration as a function of core burnup is shown below. The cycle length was increased to 25,000 MWd/MTU by increasing the burnup of each reported value by the ratio of 25,000/23,148, which is the estimated length of Unit 1 Cycle 14.

This unit and cycle were chosen because they represent the first cycle at Comanche Peak to implement the stretch power uprate. These points were then fit to determine a cubic relationship between boron concentration and cycle burnup. The boron concentration is then conservatively increased by 50 ppm. The average boron concentration in each axial zone varies based on the actual depletion steps used, and ranges from 906 to 940 ppm. This preserves the use of margin in the soluble boron concentration to account.for operating history effects. The conservatism of these average soluble boron concentrations is bolstered by the fact that the core soluble boron concentration would be lowered if the conservatively increased burnable absorber patterns were actually used during operation.

An assembly was assumed to operate at an average power of 1.2 for two cycles and then at a relative power of 0.6 for a third cycle. The assembly thus accrued 30,000 MWd/MTU depletion in each of the first two cycles and 15,000 MWd/MTU depletion in the third cycle for a total of 75,000 MWd/MTU burnup. For each depletion step in each zone, the boron concentration was determined based on the cycle burnup corresponding to the depletion of that zone. This accounted for the relative power of each node relative to the assembly average and the assembly power relative to the core average. The boron concentration used in the depletion calculations is shown below as a function of fuel assembly average burnup.

For computational efficiency, these fuel depletions and KENO calculations were performed using the PARAGON code and SCALE Version 5.1. These computer codes perform the same functions as the PHOENIX-P and SCALE Version 4.4 codes used in WCAP-16827-P. The depletion calculations used 97.5% of theoretical density.

Two sets of KENO calculations were performed. The first reproduces the depletion calculations performed in WCAP-16827-P and the second is the depletion calculations including IFBA, WABA, and the boron letdown modeling described above. This allows for a direct comparison of assembly reactivity as a function of burnup. The conservatism in the results presented in WCAP-

Attachment to TXX-08148 Page 15 of 31 16827-P is calculated by subtracting the keff from the depletion including burnable absorbers from the keff from the first set of depletion calculations. The results shown below indicate that neglecting the presence of burnable absorbers in fuel assembly depletion is a conservative practice.

The nonlinear behavior of the margin identified is a function of the combination of the two effects providing this additional margin, namely the residual IFBA boron and the soluble boron concentration modeling. The margin provided by residual IFBA is initially large and drops quickly. The margin provided by the less conservative soluble boron modeling is small at low depletion but increases with burnup. The point of minimum margin from these two sources is approximately 45,000 MWd/MTU assembly burnup.

Soluble Boron Concentration Based on Comanche Peak Unit 1 Cycle 14 Unit 1 Cycle 14 Extended Cycle Boron Burnup (MWd/MTU)

Burnup (MWd/MTU)

Concentration (ppm) 150 162 1038 1000 1080 1074 2000 2160 1147 3000 3240 1200 4000 4320 1230 5000 5400 1243 6000 6480 1242 7000 7560 1227 8000 8640 1202 9000 9720 1166 10000 10800 1122 11000 11880 1070 12000 12960 1010 14000 15120 869 16000 17280 706 18000 19440 526 20000 21600 333 22000 23760 132 23148 25000 13

Attachment to TXX-08148 Page 16 of 31 Soluble Boron Concentration Versus Assembly Average Burnup 1400 1200 1000*

800*

0a P600 0

400' 200 0

1. 0 4

4 S

4 0

0 10000 20000 30000 40000 50000 Assembly Average Burnup (MWdIMTU) 60000 70000 Conservatism of WCAP-16827-P Depletion Parameters Assembly Average Burnup (MWd/MTU)

Akefi +/- 0 25,000 0.008 18 +/- 0.00021 35,000 0.00483 +/- 0.00025 45,000 0.00225 +/- 0.00028 55,000 0.00312 +/- 0.00028 65,000 0.00637 +/- 0.00026 75,000 0.00663 +/- 0.00023 NRC Question 16:

WCAP-16827-P states, "The design parameters of the Westinghouse and Siemens 17x17 STD and OFA fuel assembly types are summarized in Table 3-5. Illustrations of these designs are contained in Figure 3-4 and Figure 3-5. Simulations are performed for each storage configuration in this analysis to determine the fuel assembly combinations that produce the highest reactivity."

Please clarify if the determination was made before or after all biases and uncertainties were applied.

Attachment to TXX-08148 Page 17 of 31 CPNPP Response:

The design basis fuel type was determined by a best estimate reactivity ranking of the two major fuel types currently being stored in the Comanche Peak units I and 2 pools. The two basic fuel types present are the standard (STD) and optimized (OFA) fuel assemblies. In this context, standard refers to any fuel assembly with a fuel rod outer diameter of 0.374" (Westinghouse) or 0.376" (Siemens) and optimized refers to any fuel assembly with a fuel rod outer diameter of 0.360". Many different names have been applied to these two basic assembly types over the years to denote differences in grids or other fuel assembly features which are conservatively neglected in this analysis.

A comparison of Westinghouse to Siemens STD fuel and Westinghouse to Siemens OFA fuel found that Westinghouse fuel is slightly more reactive than the comparable Siemens fuel. This is due to the larger pellet diameters in Westinghouse fuel. To determine the design basis fuel assembly for the analysis Westinghouse STD fuel was then compared to Westinghouse OFA fuel.

The fuel type with the highest eigenvalue was then used to calculate biases and uncertainties.

The table below shows a reactivity comparison of the two different fuel types currently being stored in the pool near the minimum burnup analyzed for 20 years of decay time. This burnup was used as a conservative lower bound on the reactivity differences as the standard assembly type, for a given enrichment, becomes increasingly more reactive than the optimized assembly with increasing burnup. When evaluated at or near the bumup limits, standard fuel is found to be more reactive. The reactivity difference ranges from nearly 1% to almost 2% Akeff. The sum of all biases and uncertainties vary from approximately 1.5% to 2.5% Akeff among the configurations reported in WCAP-168278-P. Any variations in the sum of biases and uncertainties between the fuel types would be small because the same tolerances apply to both assembly types. Therefore the choice of design basis fuel assembly need not consider the impact of biases and uncertainties.

Enrichment BU of analysis OFA keff STD keff Akeff 2.0 w/o 20,000 MWd/MTU 0.99285 1.01125 0.01840 3.0 w/o 35,000 MWd/MTU 1.00593 1.02076 0.01483 4.0 w/o 50,000 MWd/MTU 1.00918 1.02171 0.01253 5.0 w/o 60,000 MWd/MTU 1.03177 1.04038 0.00861 NRC Question 17:

Response due in January 2009.

NRC Question 18:

WCAP-16827-P makes an assumption not to model the fuel assembly spacer grids stating, "No credit is taken for spacer grids or spacer sleeves." The staff requested the licensee to justify that this assumption remained conservative in modem analysis during April 24t" teleconference. The licensee agreed that the supplement would contain to justify this information. WCAP-16827, Addendum 1 provides information for two scenarios in the "4-out-of-4" storage configuration.

One is for very low enriched fuel with no burnup (1.02 w/o 235U with 0 bumup), both with and without grids. The other is maximum enrichment with very high burnup (5.0 w/o 235U with 75 GWD/MTU burnup), both with and without grids. WCAP-16827, Addendum 1 claims these scenarios are bounding for all other storage configurations and burnup/enrichment

Attachment to TXX-08148 Page 18 of 31 combinations. With regard to this analysis the staff requests the following additional information:

a)

Provide additional details regarding the parameters used in the simulations that were performed.

b) The staff has noted a difference between the zero soluble boron starting point on WCAP-16827, Addendum 1 Figure 3-7 for, the 5.0 w/o 235U with 75 GWD/MTU burnup without grids and the WCAP-16827-P Table 4-17 value for '5.0 w/o 235U with 75 GWD/MTU burmup and zero 24 1Pu decay (also without grids) of approximately 5000 pcm. Please explain the differences between the simulations in WCAP-16827-P and WCAP-16827, Addendum 1 that result in this large difference. Explain how those differences affect the conclusions reached in WCAP-16827-P and WCAP-16827-P, Addendum 1. Please also explain why the difference does not manifest itself for the 1.02 w/o 235U with 0 bumup scenario.

c)

WCAP-16827, Addendum 1 indicates that the'l.02 w/o 235U with 0 bumup scenario in the "4-out-of-4" storage configuration results in the largest non-conservatism. While the "4-out-of-4" storage configuration may be bounding with respect to the other storage configurations; that does not mean the result for those storage configurations will be zero. It appears that WCAP-16827, Addendum 1 made no effort to determine the effect on those storage configurations. As WCAP-16827-P and WCAP-16827, Addendum 1 indicate CPSES, Units 1 and 2 require in excess of 1600 PPM of soluble boron to meet the regulatory requirements the staff considers the determination of this bias to be important.

In order to ensure compliance with the regulation the licensee should develop and apply an appropriate bias for each storage configuration. The licensee is requested to provide the information used to develop. and apply the bias. CPSES, Units 1 and 2 have used several different fuel designs, please indicate, which fuel design was used as the basis for the grids and how is this fuel design limiting?

CPNPP Response:

It has long been assumed in criticality safety analyses that neglecting the presence of grids is a conservative assumption for reactivity determination. Over the years different materials have been used for these grids including stainless steel, Zircaloy, and ZIRLO. The assumption of conservatism may date back.to the use of stainless steel, so a reconfirmation of the assumption is warranted. The calculations presented below confirm the continuing validity of the assumption that neglecting the presence of grids is a conservative assumption for criticality safety analyses.

The uniform burnup profile was analyzed in the WCAP-16827-P Addendum 1 in order to simplify the models used in the analysis. This simplification is the reason for the difference in the zero soluble boron concentration starting point between the original WCAP and the addendum.

Reactivity differences associated with uniform and distributed profiles are provided in the table below. The reactivity margin associated with the distributed profile results in a higher margin, therefore proving, that neglecting the, presence of grids is conservative for highly depleted fuel and that the conclusion is not affected by the use of the uniform burmup model. Comparison of the results of both analyses show that while there is a difference in reactivity margin between the distributed and uniform burnup profile, the Akeff associated with the distributed burnup profile is greater and therefore more conservative. Thus, the conclusions presented in WCAP-16827-P Addendum 1 still-remain valid.

Attachment to TXX-08148 Page 19 of 31 Reactivity Margin as a Function of Power Shape and Boron Concentration Boron Reactivity Margin (10-5 Akeff)

Concentration Fresh Fuel Uniform Burnup Distributed (ppm)

Profile Bumup Profile 0

41 164 224 100 16 119 250 200

-17 92 200 300

-9 110 146 400

-15 99 195 500

-54 79 153 600

-27 103 139 The models used in the analysis to examine the reactivity impact of grids over a range of soluble boron concentrations from 0 ppm to 600 ppm are consistent with those used in WCAP-16827-P, Addendum 1. These models consisted of eight spacer grids equally spaced across the active height of the fuel. The grids were modeled using Zircaloy, the least absorptive grid material currently used. Zircaloy has a lower absorption cross section than ZIRLO because it lacks niobium as a material component. Niobium is known to be a neutron absorber, thus its presence in a grid would lower reactivity. Both zirconium alloys have significantly lower cross sections than the stainless steel used in grids in some older fuel assembly designs. To remain conservative, a water/Zircaloy smear was incorporated outside the fuel rods to model a physical grid as accurately as possible. Using the water smear reduces the volume fraction of the Zircaloy material and conserves the actual number of Zircaloy atoms in each grid across the as-modeled volume.

Addendum I analyzed the effect of grids with boronconcentrations ranging from 0-2000 ppm.

This analysis focused on boron concentrations of 0-600 ppm because it was found that the kff resulting from use of the distributed burnup shape at boron concentrations of 600 ppm and greater are less than 0.90. Beyond this point sufficient margin exists due to the presence of soluble boron to account for any potential non-conservatism caused by neglecting the presence of grids.

Assumptions for fresh 1.02 w/o 235U fuel are potentially non-conservative in that neglecting the grids can result in a less reactive assembly model than when grids are taken into account.

However, potential non-conservatisms are overwhelmed by the first order effect of soluble boron lowering reactivity. In order to prove that fresh fuel still meets the requirements of 10CFR50.68, a very conservative determination of absolute reactivity will be made to show that keff is less than the 0.95 requirement. First, the sum of biases and uncertainties are increased by more than 20%

to account for the presence of soluble boron. Both recent (Reference 5) and historical (Reference

6) evaluations have shown that soluble boron does not have a large positive impact on the sum of biases and uncertainties, so this assumption is extremely conservative. Second, the worst potential non-conservatism caused by neglecting grids reported in WCAP-168277P, Addendum 1 is doubled. The boron concentration for this largest potential non-conservatism is 1800 ppm.

These two terms can be added to the highest kff with identified potential non-conservatism, as shown below in Equation 1. This occurs at 200 ppm of soluble boron, and the corresponding keff is 0.90862 for fresh 1.02 w/o fuel. This results in a conservative estimate of keff equal to 0.94186.

Attachment to TXX-08148 Page 20 of 31 This result still satisfies the required limit of 0.95.

k c7,.s = k I + Ak BU + Akg,.is Eqn (1)

Where: kco°s is a conservative estimate of the absolute reactivity kcaic is the KENO calculated kff equal to 0.90862 AkB&u is the conservatively increased sum of biases and uncertainties, equal to 0.03 Akeff Akgrids is the conservatively doubled magnitude of potential non-conservatism caused by not modeling grids, equal to 0.00324 NRC Question 19:

WCAP-16827-P makes an assumption not to model Boraflex wrapper material present in the CPSES, Unit 2 Region II SFP. Although the Boraflex was never installed, the wrapper material is present. WCAP-16827, Addendum 1 provides information to validate that assumption. WCAP-16827, Addendum 1 provides information for~two scenarios in the "4-out-of-4" storage configuration. One is for very low enriched fuel with no burnup (1.02 w/o 235U with 0 burnup).

The other is maximum enrichment with very high burnup (5.0 w/o 235U with 75 GWD/MTU burnup). WCAP-16827, Addendum 1 claims these scenarios are bounding for all other storage configurations and burnup /enrichment combinations. Information provided in WCAP-16827, Addendum 1 indicates there is a large amount of conservatism associated with this assumption at zero soluble boron concentrations, and that the conservatism decreases with increasing soluble boron concentration. The staff has noted a difference between the zero soluble boron starting point on WCAP-16827, Addendum I Figure 3-9 for the 5.0 w/o 235U with 75 GWD/MTU burnup without grids and the WCAP-16827-P Table 4-17 value for 5.0 w/o 235U with 75 GWD/MTU burnup and zero 241Pu decay (also without grids) of approximately 5000 pcm. Please explain the differences between the simulations in WCAP-16827-P and WCAP-16827, Addendum 1 that result in this large difference. Please also explain how those differences affect the conclusions reached in WCAP-16827-P and WCAP-16827-P, Addendum 1 and why the difference does not manifest itself for the 1.02 w/o 235U with 0 burnup scenario.

CPNPP Response:

The information provided in WCAP-16827-P, Addendum 1 is less reactive because the uniform burnup profile was used for the analysis. This decision was made to simplify the construction of the computational model. When the distributed burnup profile is used, results show greater margin relative to the presence of the wrapper material. Therefore, the uniform profile provides a conservative estimate of the reactivity associated with the presence of the wrapper. Both burnup profiles confirm the conservatism of the assumption to neglect the wrapper in this analysis.

The table below shows the reactivity margin associated with neglecting the presence of the wrapper material when modeling fresh, uniform burnup and distributed burnup fuel assemblies.

The fresh and uniform burnup profile results are the results reported in WCAP-16827-P, Addendum 1. The depleted fuel has an initial enrichment of 5.0 w/o and a burnup of 75,000 MWd/MTU. The results are presented through a soluble boron concentration of 600 ppm because the calculated kff of the distributed burnup profile case drops below 0.9 at higher concentrations.

Attachment to TXX-08148 Page 21 of 31 Boron Reactivity Margin (10-5 Akeff)

Concentration Fresh Fuel Uniform Burnup Distributed (ppm)

Profile Burnup Profile 0

909 598 661 100 801 570 603 200 744 517 546 300 688 489 510 400 622 442 494 500 535 403 460 600 519 371 395 NRC Question 20:

With regard to the oversized inspection cell, WCAP-16827-P states, "An empty row of storage cells is included in all adjacent locations, including diagonal cells. The surrounding storage locations in the model contain STD fuel assemblies at the maximum permissible enrichment for the "4-out-of-4 storage configuration." This appears to require the oversized inspection cells to be bordered by a complete row of empty cells. This requirement is not captured in the proposed CPSES, Units 1 and 2 TS. Please provide a revised TS pages to capture this requirement.

CPNPP Response:

As required by 10CFR50.68, Comanche Peak has fuel handling procedures that ensure the spent fuel storage requirements of the Technical Specifications are met and prohibit the handling, at any one time, of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water. Procedural controls are used for movement of fuel that is outside of the spent fuel storage racks, including fuel located in the transfer canal upender, in the new fuel elevator, as well as the oversized inspection cell.

The oversized inspection cell is used for inspection of spent fuel and not an approved as a spent fuel storage location. By procedural controls, fuel outside of a storage location may not be left unattended and must be controlled consistent with design requirements. Procedural controls have been in place for a number of years which prohibit the use of the oversized inspection cell while adjacent fuel storage locations are occupied, which is consistent with both the current design requirements and the analysis supporting the proposed license amendment. Since the oversized inspection cell is not an approved storage location, procedural requirements ensure that fuel is not left unattended during use of the cell for fuel inspection. The oversized inspection cell is not typically used, since typical fuel inspections occur in the transfer canal or wet cask pit area. In the case of utilizing the area for fuel inspections, the fuel assembly would not be unlatched from the handling tool, and total time in the inspection cell would likely be well under one hour.

Previous license amendments regarding the use of the spent fuel storage racks have not identified the need to include a Technical Specification requirement for fuel handling procedures regarding the oversized inspection cell. No changes in the design or the requirement to have an empty row of storage cells in all adjacent locations to the oversized inspection cell while the cell is in use are proposed. Therefore, Luminant Power believes that no additional Technical Specification requirements regarding fuel handling procedures are warranted.

Attachment to TXX-08148 Page 22 of 31 NRC Question 21:

WCAP-16827-P doesn't establish uncertainties for the U0 2 density manufacturing tolerances or dishing and chamfering. Rather WCAP-16827-P models a set density and all fuel pellets are modeled as full right circular cylinders. WCAP-16827, Addendum 1 provides additional information that indicates the set density is above the nominal density plus tolerances. WCAP-16827, Addendum I provides additional information that indicates the nominal dishing and chamfering for CPSES, Units I and 2 fuel. WCAP-16827,;Addendum 1 claims these assumptions are conservative due to the additional fissile material they provide. However, that rationale ignores the potential for self-shielding. A review of past submittals, which determined a U0 2 density uncertainty (References 11, 12, and 13), indicate the Akeff between the nominal case and perturbed case with increased U0 2 density can be less than the KENO case uncertainty, and when the KENO case uncertainty is applied to determine a margin that margin may be negative.

Please state how is the CPSES, Units 1 and 2 analyses crediting the 'margin' associated with U0 2 density and dishing and chamfering.

CPNPP Response:

Throughout the Comanche Peak Units I and 2 analyses presented in WCAP-16827-P and Addendum 1, no credit is taken for assumptions related to fuel pellet theoretical density or dishing and chamfering. Should any of this margin be used to respond to any requests for additional information, the reactivity impact will be explicitly determined. The direct calculation of additional margin will, if needed, demonstrate that an appropriate amount of reactivity margin can be regained by modifications of the pellet theoretical density or dishing and chamfering assumptions.

NRC Question 22:

WCAP-16827-P does consider uncertainties for the fuel assembly manufacturing tolerances associated with pellet diameter, cladding thickness, and enrichment, but does not establish uncertainties for fuel rod pitch or the guide tubes. Please state why reactivity uncertainties were not considered for these manufacturing tolerances.

CPNPP Response:

The tolerance on guide tubes is not treated directly because there will be no statistically significant reactivity impact as a result of the manufacturing tolerance on guide tubes. The manufacturing controls on guide tubes are similar to those applied to fuel cladding, so the tolerances are similar. The reactivity sensitivity to the cladding tolerance has been shown to be small, that is a best estimate reactivity change of less than 0.00100 Akeff in each of the "2-out-of-4",

"3-out-of-4", and "4-out-of-4" configurations. The changes in metal and water volumes associated with the guide tube tolerances will be reduced compared to those caused by the cladding tolerances because only 25 guide and instrument tubes are present in each assembly compared to 264 fuel rods. The combination of a small sensitivity and a small change result in a negligible reactivity effect caused by manufacturing tolerances on guide and instrument tubes.

Direct simulations were performed to investigate the impact of a bounding fuel rod pitch uncertainty. Relevant manufacturing tolerances for the design basis fuel assembly for each configuration were obtained and statistically combined to estimate a fuel rod pitch tolerance. The resulting estimated value, approximately 0.0023", was then applied to all the rods to increase the fuel rod pitch once it was determined that the design basis fuel assembly is under moderated in all three configurations. While manufacturing such an assembly is not credible, this treatment provides a bounding estimate of the worst case reactivity uncertainty caused by variability in the pin pitch due to manufacturing. The overall assembly size after the pitch is increased is

Attachment to TXX-08148 Page 23 of 31 approximately 8.472", which is larger than the assembly pitch in the core of 8.466", confirming the bounding nature of the pitch tolerance determined here. The fuel rod pitch reactivity uncertainty was then statistically combined with the other uncertainties for each configuration to determine the reactivity impact on the sum of biases and uncertainties. The results are listed in the table below. The largest magnitude of potential non-conservatism identified is less than 0.00200 Akeff. This increase in uncertainty is less than the 0.00500 Akeff of administrative margin.

Only the unblanketed configurations were examined. The blanketed configurations will be less sensitive to the additional tolerance because the burnup uncertainty is larger in every case for blanketed version of a configuration.

Fuel Rod Change in Sum of Configuration Pitch Uncertainty (Akeff)

Biases and Uncertainties (Akeff)

"2-out-of-4" 0.00666 0.00195 "3-out-of-4" 0.00333 0.00046 "4-out-of-4" 0.00209 0.00016 NRC Question 23:

WCAP-16827-P also establishes an uncertainty for eccentric positioning of fuel assemblies in the

" storage cells. That uncertainty is established by simulations "...performed to investigate the effect of off-center position of the fuel assemblies for each of the fuel assembly storage configurations.

These simulations positioned the assemblies as close as possible in four adjacent storage cells and at intermediate positions in between."

a)

Moving the assemblies closer together in the center of the model moves them further away from the boundary. In determining this uncertainty each storage configuration is modeled as infinitely repeating 2x2 arrays with periodic boundary conditions. Therefore, as the distance to the modeled 2x2 array boundary increases, the distance to the next fuel assembly is being doubled effectively opening the intra-array gap and producing a potentially negative reactivity effect that may not actually be present. Please state how the results of the eccentric position uncertainty would change if fuel assemblies surrounding the 2x2 array were held in the nominal position.

b)

No eccentric positioning uncertainty is listed for the "4-out-of-4," "4-out-of-4 with Axial Blankets," "3-out-of-4," or "3-out-of-4 with Axial Blankets," storage configurations, thus indicating the analysis did not find a statically significant resultant. Please state if the simulations performed in the analysis considered both the presence and absence of axial blankets in reaching the conclusion that there was not a statistically significant resultant.

CPNPP Response:

The treatment of the eccentric position uncertainty in WCAP-16827-P was based on the 2 x 2 infinite array models. A range of off-nominal positions were considered for each storage configuration and significant negative reactivity trends were noted in the "3-out-of-4" and "4-out-of-4" storage configurations. The sensitivity of the models to assembly position indicated that a nominal location would always be most reactive, but direct calculations considering the specific scenario indicated in Part a) are warranted. The reactivity increase noted in the "2-out-of-4" configuration should bound all less extreme scenarios.

Attachment to TXX-08148 Page 24 of 31 The effect of eccentric positioning. on blanketed fuel assemblies was assumed to be bounded by that for unblanketed assemblies. The reduced fissile mass present in the assembly should have a slight dampening effect on the assemblies as they are misaligned.

Part a)

The uncertainty calculations were redone with a model including two rows of nominally positioned assemblies arrayed outside of the eccentrically positioned 2 x 2 array. Meaning a 6 x 6 array was modeled with',the position of the assemblies Within a cell only changing in the center 2 x 2 array of storage cells. The results of both the original analysis and the reanalysis are reported below. For each of the unique rack geometries mentioned above the reactivity of the 6 x 6 array of assemblies was closer to nominal than the original 2 x 2 array analysis. In other words, the eccentric positioning of the "4-out-of-4" and "3-out-of-4" configurations had a less negative reactivity impact and the eccentric positioning of the "2-out-of-4" configuration showed a less positive reactivity insertion compared to the original analysis. It is concluded that the eccentric positioning uncertainty treatment included in the WCAP-16827-P analysis is appropriate.

Positioning Eigenvalue +

Best Estimate Uncertainty Reactivity Change "4-out-of-4" Nominal 0.96624 +/- 0.00025 Half Displacement 0.96325 +/- 0.00023

-0.00299 Maximum Displacement 0.95594 +/- 0.00024

-0.01030 Maximum Displacement w/ nominal ring 0.96527 +/- 0.00023

-0.00097

."3-out-of-4" Nominal 0.97670 +/- 0.00030 Half Displacement 0.97472 +/- 0.00030

-0.00198 Maximum Displacement 0.96921 +/- 0.00028

-0.00749 Maximum Displacement w/ nominal ring 0.97670 +/- 0.00029 0.00000 "2-out-of-4" Nominal 0.97528 +/- 0.00039 Half Displacement 0.97650 +/- 0.00039 0.00122 Maximum Displacement 0.98068 +/- 0.00041 0.00540 Maximum Displacement w/ nominal ring 0.97589 +/- 0.00043 0.00061 Part b)

Simulations were performed with isotopics near the burnup limit on a configuration specific basis for both the "4-out-of-4 with Axial Blankets" and the "3-out-of-4 with Axial Blankets" storage configurations. Models were constructed with the uniform burnup profile consistent with the axial blanket modeling procedure used throughout the analysis. The results listed in the tables below show that the reactivity effect associated with the blanketed configurations is also negative and is appropriately taken as zero in the analysis presented in WCAP-16827-P.

Intermediate eccentric positioning was not considered for the blanketed configurations as it was always bounded by either the maximum displacement or maximum displacement with nominal ring configurations for the unblanketed configurations.

Attachment to TXX-08148 Page 25 of 31 "4-out-of-4 with Axial Blankets"...

Nominal 0.96478 +/- 0.00024 Maximum Displacement 0.95301 +/- 0.00023

-0.01177 Maximum Displacement w/ nominal ring 0.96369 +/- 0.00021

-0.00109 "3-out-of-4 with Axial Blankets" Nominal 0.95942 +/- 0.00029 T_

Maximum Displacement 0.95121 +/- 0.00029

-0.00821 Maximum Displacement w/ nominal ring 0.95845 +/- 0.00027

-0:00097 NRC Question 24:

The reactivity uncertainties established in WCAP-16827-P are based on perturbations of the minimum allowed fresh fuel in a given storage configuration. In some storage configurations there is a large range between the minimum allowed fresh fuel and the maximum allowed enrichment/burnup combination. For example the minimum allowed fresh fuel in the "4-out-of-4" storage configuration' is 1.02 w/o 235U while the maximum allowed enrichment/burnup combination is 5.0 w/o 235U with 75,729 MWd/MTU burnup.

a)

What assurance is there that the uncertainties do not change over the range covered by the storage configurations?

b) If burnup is considered, how might decay time affect the uncertainties?

CPNPP Response:

The calculation of biases and uncertainties in WCAP-16827-P has selectively considered fresh fuel in some cases and depleted fuel in others to maximize the calculated reactivity effect. The uncertainty calculations for material and dimensional tolerances are performed with fresh fuel because the fresh fuel is expected to be more responsive to these types of changes. Depleted fuel is used in the determination-of the temperature bias because the depleted fuel should.have a somewhat hardened neutron spectrum which will be more sensitive than fresh fuel to the moderation change. The presence of plutonium, especially 241Pu, will also increase the magnitude of the temperature bias. Because 241Pu is a significant factor in the determination of the temperature bias, it is expected that this bias will decrease as a function of decay time.

Biases and uncertainties were calculated at high burnup and decay time conditions for the parameters listed in the tables below to ensure the most conservative reactivity uncertainties were reported. The total of the reactivity biases and uncertainties was then determined for.

depleted fuel, and depleted fuel with credited decay time and compared to the bias and uncertainty results for fresh fuel reported in WCAP-16827-P. Each uncertainty related to dimensional tolerances was analyzed for highly depleted fuel (75,000 MWd/MTU burnup) and

• for highly depleted, decaying fuel (70,000 MWd/MTU burnup, 20 year decay period). All cases examined use~d fuel with an initial enrichment of 5.0 w/o 235U. The enrichment and burnup uncertainties were used as presented in Table 4-8 of WCAP-16827-P.

The cases using depleted fuel resulted in generally lower sensitivities, but increased uncertainties are noted for cladding thickness and cell inner dimension. While the maximum pellet outer diameter uncertainty was analyzed, the value (kff= 0.97780 +/- 0.00012) was omitted because the sensitivity was found to be less than zero. Therefore, there is no reactivity increase associated with the pellet diameter tolerance. The temperature bias reported below was reported in WCAP-16827-P based on the use of 5.0 w/o fuel depleted to 80,000 MWd/MTU. The total of all

Attachment to TXX-08148 Page 26 of 31 reactivity biases and uncertainties was found to be 0.1996 Akeff, which is 0.00036 Akeff lower than originally stated in WCAP-16827-P.

Biases and Uncertainties for 75,000 MWd/MTU Burnup Fuel Case Description keff Akeff Nominal Case 0.97781 +/- 0.00014 Minimum Cell Pitch 0.97967 +/- 0.00026 0.00226 Minimum Cladding Thickness 0.97920 +/- 0.00028 0.00181 Minimum Cell Wall Thickness 0.97930 +/- 0.00030 0.00193 Minimum Storage Cell Inner Dimension 0.97823 +/- 0.00014 0.00070 Burnup Uncertainty 0.01412 Enrichment Uncertainty 0.00171 Methodology Uncertainty 0.00705 Methodology Bias 0.00283 Pool Temperature Bias 0.00086 Total Biases and Uncertainties 0.01996 The effects of cooling time credit on reactivity biases and uncertainties were analyzed using 70,000 MWd/MTU burnup fuel at 5.0 w/o 235U enrichments after a 20 year decay period. The cases resulted in generally lower sensitivities, but increased uncertainties are noted for cladding thickness and cell inner dimension. Pellet outer diameter uncertainty was also omitted for the decay cases because it was found to be statistically insignificant, with a kff of 0.93644 +/- 0.00012. If this uncertainty were taken into account to add conservatism, the resulting total reactivity bias and uncertainty would still be less than originally stated in WCAP-16827-P. The enrichment and burnup uncertainties were again used directly from Table 4-8 of WCAP-16827-P. Analysis found that temperature has no impact as the reference conditions of 68°F and 1 g/cm3 water density are the most reactive in the normal operating range. Re-evaluation of all parameters resulted in a total sum of reactivity biases and uncertainties of 0.01900 Akeff, which is less than the originally reported value of 0.02032 Akeff.

Biases and Uncertainties for 20 year Decay Period, 70,000 MWdIMTU Burnup Fuel Case Description keff Akeff Nominal Case 0.93640 +/- 0.00014 Minimum Cell Pitch 0.93815 +/- 0.00027 0.00216 Minimum Cladding Thickness 0.93739 +/- 0.00029 0.00142 Minimum Cell Wall Thickness 0.93756 +/- 0.00028 0.00158 Minimum Storage Cell Inner Dimension 0.93677 +/- 0.00015 0.00066 Burup Uncertainty 0.01412 Enrichment Uncertainty 0.00171 Methodology Uncertainty 0.00704 Methodology Bias 0.00283 Pool Temperature Bias 0.00000 Total Biases and 0.01900 Uncertainties The temperature bias was recalculated for fresh fuel at 1.01 w/o 235U to provide consistent data for fresh fuel. The temperature bias reported in WCAP-16827-P was for 5.0 w/o highly depleted fuel. The new temperature bias calculated was less than the temperature bias stated in Table 4-1 of WCAP-16827-P. Applying the new temperature bias lowers the total reactivity bias and uncertainty by 0.00033 Akeff. This demonstrates that the evaluation of uncertainties with1 fresh

Attachment to TXX-08148 Page 27 of 31 fuel and the temperature bias with depleted fuel yields a conservative determination of the applicable biases and uncertainties. Furthermore, this analysis supports the conclusion that if burnup dependent biases and uncertainties were calculated, their sum would be lower than that reported in Table 4-1 of WCAP-16827-P for each combination of enrichment, burnup and decay time.

Total Biases and Uncertainties of Fresh 1.01 w/o 235U Fuel Case Description keff Akeff Nominal Case 0.96624 +/- 0.00025 Minimum Cell Pitch 0.96997 +/- 0.00025 0.00423 Minimum Cladding Thickness 0.96638 +/- 0.00023 0.00062 Minimum Cell Wall Thickness 0.96882 +/- 0.00026 0.00249 Enrichment Uncertainty 0.01916 Methodology Uncertainty 0.00704 Methodology Bias 0.00283 Pool Temperature Bias 0.00054 Total Biases and 0.02437 Uncertainties NRC Question 25:

The reactivity uncertainties established in WCAP-16827-P are based on perturbations of the minimum allowed fresh fuel in a given storage configuration. The "4-out-of-4" and "4-out-of-4 with Axial Blankets" storage configurations, with the exception of the burnup uncertainty, use the same 'rackup' of biases and uncertainties. Similarly the "3-out-of-4" and "3-out-of-4 with Axial Blankets" storage configurations, with the exception of the burnup uncertainty, share the same 'rackup' of biases and uncertainties.

a) Please explain why the presence of axial blankets does not affect the enrichment uncertainty.

b) What assurance is there that the other uncertainties do not change with the presence axial blankets?

CPNPP Response:

Part a)

The presence of axial blankets does not change the enrichment uncertainty because the overall reactivity of the fuel assembly is driven by the section of the fuel with the highest reactivity. In cases with axial blankets, this section of fuel is the fully enriched, non-blanketed portion of the assembly. The case originally analyzed without axial blankets serves as a bounding analysis because the presence of axial blankets lowers overall reactivity of the fuel rod.

Enrichment uncertainty was analyzed to account for axial blankets. The results confirmed the above statement. The calculated enrichment uncertainty was found to be lower than the value reported in Table 4-9 of WCAP-16827-P. The reported enrichment uncertainty in Tables 4-8 and 4-9 is 0.00522 Akeff for 3 w/o assemblies. The calculated enrichment uncertainty for a 3 w/o assembly with axial blankets of 2 w/o is 0.00449 Akeff.

Attachment to TXX-08148 Page 28 of 31 Part b)

Other calculated uncertainties do not change as a result of the presence of axial blankets because the uncertainties in reactivity are driven by the high enriched fuel center, not the low enriched axial blankets. Therefore, uncertainties calculated using fully enriched fuel bound cases containing axial blankets.

NRC Question 26:

Response due in January 2009.

NRC Question 27:

WCAP-16827-P assumed the "4-out-of-4" storage configuration as bounding for all other storage configurations in its soluble boron crediting methodology. The assumption was proven invalid in WCAP-16827-P, Addendum 1. WCAP-16827-P, Addendum I repeatedly used the "4-out-of-4" storage configuration as bounding for all other storage configurations. The staff requests quantitative evidence that the "4-out-of-4" storage configuration is indeed bounding for all other storage configurations were it is used as such.

CPNPP Response:

The "4-out-of-4" storage configuration is usually presented as bounding because the fuel assembly burnup is maximized. This is a direct influence on parameters such as the depletion uncertainty. In other cases, the "4-out-of-4" storage configuration is chosen because all locations contain a fuel assembly, thus maximizing a particular perturbation. This is the case for determining the impact of modeling grids, as the larger number of grids present in four assemblies will likely bound the effect of a lesser number of grids. Other effects, such as burnup and neutron energy spectrum, can also impact the sensitivity of a particular storage configuration. The calculations below demonstrate that the "4-out-of-4" storage configuration is a very reliable indicator of the reactivity impact for the situations in WCAP-16827-P which have identified it as limiting.

The depletion uncertainty reported in Figures 3.1-3.4 of WCAP-16827-P, Addendum I was calculated using the "4-out-of-4" storage configuration. This uncertainty has been recalculated for the "2-out-of-4" and "3-out-of-4"' storage configurations near the burnup limits. The results, displayed in the table below, show that the "4-out-of-4" configuration is limiting, and therefore bounding for depletion uncertainty.

Depletion Uncertainty for 2-out-of-4, 3-out-of-4, and 4-out-of-4 Storage Configurations at Burnup Limits Storage Burnup Depletion Configuration (MWd/MTU)

Uncertainty Akeff) 2-out-of-4 10,000 174 3-out-of-4 45,000 485 4-out-of-4 75,000 664 The effect of grids over a range of boron concentrations was also investigated for the three configurations used in analysis. The effect of grids over the range of boron concentrations analyzed varies among the three configurations. The "3-out-of-4" storage configuration shows

Attachment to TXX-08148 Page 29 of 31 less sensitivity to the presence of grids, but these differences still demonstrate that neglecting grids is a conservative approach. The "2-out-of-4" storage configuration is demonstrated to be more conservative with respect to neglecting grids than the "4-out-of-4" storage configuration.

The effect of grids over a range of soluble boron concentrations is shown in the table below. All three configurations exhibit conservatism in this analysis.

Effect of Grids and Boron Concentration on Reactivity for 2-out-of-4, 3-out-of-4, and 4-out-of-4 Storage Configurations Boron 2-out-of-4 3-out-of-4 4-out-of-4 Concentration (ppm) 10,000 MWd/MTU 45,000 MWd/MTU 75,000-MWd/MTU 0

278 213 224 100 284 152 250 200 235 205 200 300 213 160 146 400 242 114 195 500 231 111 153 600 195 137 139 NRC Question 28:

Response due in January 2009.

NRC Question 29:

Please state how the "1-out-of-4" storage configuration currently in the CPSES, Units 1 and 2, TS is affected by the proposed changes.

CPNPP Response:

The current licensing basis for the "l-out-of-4" storage configuration does not credit fuel bumup to demonstrate compliance with the limits contained in 10CFR50.68. The proposed changes are intended to account for differences in fuel assembly reactivity caused by the stretch power uprate at Comanche Peak Units 1 and 2. The configuration is therefore unaffected and remains unchanged in the proposed technical specifications.

NRC Question 30:

Describe'the process used to determine that fuel assemblies have attained proper bumup for storage in the burnup dependent racks.

CPNPP Response:

The source of Fuel Assembly Burnup and enrichment values used in the TS 3.7.17 determinations is controlled procedurally to ensure the source is valid. There are two approved sources for the information.

Much of the historical data is extracted from the TracWorks software program, which is utilized throughout the industry for Special Nuclear Material tracking and reporting. Updating TracWorks is controlled procedurally, and the data used to input into TracWorks comes from approved Burnup determinations. For current cycles, BEACON is the source of fuel assembly burnup and isotopics. For previous cycles, the CONFORM software package was utilized. Both

Attachment to TXX-08148 Page 30 of 31 of these sources utilize measured power distributions to iteratively determine the accumulation of assembly brmup.

Because TracWorks is not a real-time database (for example, updates to TracWorks may not be performed prior to offloading the core during a refueling outage), the source for TS 3.7.17 determinations may also come directly from BEACON or CONFORM. When this method is used, the information is typically more conservative, since the burnup values are determined prior to shutdown (and additional burnup on the assemblies occurs prior to offloading the core).

The burnup and enrichment information is obtained and input into a validated software program which performs the interpolation of TS 3.7.17 limits for each assembly, to determine which configuration the assembly may be stored in. Each step of the proccess, including power distribution measurements, accumulated assembly burnup reports, TracWorks updates, and TS Limit Determinations, requires the performer and independent reviewer to be a Qualified Reactor Engineer. Work is performed in accordance with procedures controlled through the.

CPNPP Quality Assurance program.

NRC Question 31:

Describe the process used to control movement of items within the SFP.

CPNPP Response:

Any fuel or fuel components requires movement tracking paperwork. This paperwork is generated by Core Performance Engineering and approved by the Core Performance Supervisor and a Fuel Handling qualified Senior Reactor Operator. The software used to generate fuel moves is administratively controlled to ensure the most recent configurations are used.

A separate quality assurance controlled software program is used to determine each fuel assembly's allowed TS configuration. This information is captured under a TS 3.7.17 surveillance work order, prior to generation of any fuel movement paperwork, which is referenced for each fuel movement within the Region II racks. Maps are created from the TS computations to review acceptability of movement within the Region II racks.

Following a fuel move campaign, the configuration is checked against maps generated from the fuel movement tracking software. This provides additional assurance that the pool was properly configured per the fuel movement plans.

CPNPP experienced a misloaded fuel assembly in the Region II racks in 2004 and personnel are aware of industry issues related to fuel storage errors. In the specific case of the CPNPP violation, a root cause investigation determined that the primary cause was an error in the database which.stored fuel assembly burnups. The error was introduced during a conversion of the database to support implementation of new software. Since then, a 100% verification has been performed on the database, and additional procedural controls have been put in place to ensure the validity of the information used to determine storage configurations in the future.

NRC Question 32:

Describe how this LAR affects CPSES, Units 1 and 2's B.5.b commitments.

CPNPP Response:

There is no affect on B.5.b commitments related to this LAR. CPNPP will continue to store the fuel in a manner which meets the requirements of reactivity related Technical Specification Limits as well as thermal storage requirements related to B.5.b commitments.

Attachment to TXX-08148 Page 31 of 31 References

1. "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," WCAP-11596-P-A, June, 1988.

2.

J.C. Wagner, et al, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses," NUREG/CR-6801, March, 2003.

3.

"Comanche Peak Nuclear Power Plant Stretch Power Uprate Licensing Report," WCAP-16840-P, August, 2007.

4.

Parks, C. V. et. al., "Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel," NUREG/CR-6665, February, 2000.

5.

L. Meyer (Point Beach) Letter to USNRC, "Supplement to License Amendment Request Number 247, Spent Fuel Pool Storage Criticality Control," dated September 19, 2008, ADAMS Accession Number: ML082630114.

6.

J.D. O'Hare, et. al., "Comanche Peak High Density Spent Fuel Rack Criticality Analysis Using Soluble Boron Credit/And No Outer Wrapper Plates," dated April 2000, ADAMS Accession Number: ML003760081.