ML20134F112

Provides Suppl Info Which Was Discussed During Telcon W/Util & NRC to Support NRC Review of License Amend & Exemption Request to Permit Use of Lead Test Assemblies in Plant, Units 1 & 2

ML20134F112
Person / Time
Site:	North Anna
Issue date:	02/03/1997
From:	Ohanlon J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20134F115	List: ML20134F112 ML20134F123
References
96-409A, NUDOCS 9702070284
Download: ML20134F112 (10)

Text

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VIHOINIA ELECTRIC ann POWER COMPANY nicnwoxu,vinorwrA cocos l February 3, 1997 l U. S. Nuclear Regulatory Commission Serial No. 96-409A Attention: Document Control Desk NO&LS/ETS Washington, D.C. 20555 Docket Nos. 50-338 50-339 l License Nos. NPF-4 l NPF-7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 SUPPLEMENTAL INFORMATION CONCERNING CORE DESIGN MARGINS AND FUEL MANAGEMENT FOR USE OF FCF LEAD TEST ASSEMBLIES l

This letter provides the supplemental information which was discussed in teleconference calls between Virginia Electric and Power Company and the NRC staff. This information is provided to support the NRC staff's review of the license amendment and exemption request to permit the use of lead test assemblies in North Anna Units 1 and 2. The license amendment and exemption request were submitted in our letter Serial No.96-409, dated September 4,1996. The discussions with the NRC staff provided data conceming margins which are inherent in the proposed core design containing the lead test assemblies. Specifically, on November 19,1996 and December 19,1996, staff from our Nuclear Analysis and Fuel Department provided numerical values of calculated margin in two key power distribution parameters, Enthalpy Rise Hot Channel Factor (FAH) and Heat Flux Hot Channel Factor (FO). Attachment 1 provides more complete calculated results which have been obtained from analysis of the proposed North Anna 1, Cycle 13 core design containing the lead test assemblies. Information pertaining to fuel l management considerations for the cores which will contain the lead test assemblies is l also provided.

In light of recent control rod insertion problems at some units, which may be related to i fuel assembly guide thimble conditions, the NPC requested the Company consider RCCA l control rod drag testing on the lead test assemblies. The Company has concluded that i this additional testing would contribute useful data which is fully consistent with the overall i program objectives. Therefore, Virginia Electric and Power Company will perform drag I[

testing on the FCF lead test assemblies as described in Attachment 1. Although it is intended to perform the testing as proposed, the scope and schedule of such testing may be modified if necessary to conform to overall refueling outage constraints.

9702070294 970203 PDR ADOCK 05000338 P PDR

i Also, during the telephone conference calls with the NRC Staff, altemative words were discussed for the proposed changes to Technical Specifications Section 5.3.1.

Attachment 2 provides revised words for the proposed change to Section 5.3.1 that more closely conform to the Standard Technical Specifications. The proposed changes in Attachment 2 supersede the respective changes contained in the initial September 4, 1996 submittal.

The proposed revision to Section 5.3.1 of the Technical Specifications does not affect the original basis for our determination that the changes do not involve a significant hazards consideration. The proposed Technical Specifications changes have been reviewed by the Station Nuclear Safety and Operating Committees and the Management Safety Review Committee.

The commitments made in this letter are summarized in Attachment 3. If you have any further questions or concems, please contact us.

Very truly yours, J. P. O'Hanlon Senior Vice President - Nuclear Attachments cc: U. S. Nuclear Regulatory Commission Region ll 101 Marietta Street, N. W.

Suite 2900 Atlanta, Georgia 30323 Mr. R. D. McWhorter NRC Senior Resident inspector North Anna Power Station Bureau of Radiological Herfth Room 104A 1500 East Main Street Richmond, Virginia 23219

Aurchm:nt 1 Fuel Management and inherent Core Design Margins Applicable to Use of FCF Lead Test Assemblies in North Anna 1 Cycle 13 General Description The FCF lead test assemblies are scheduled to be first irradiated in North Anna 1, Cycle 13. As described in the Company's license amendment request (Reference 1), the lead test assemblies l are functionally equivalent to the resident Westinghouse fuel and incorporate several advanced

! features. The intent of the program is to provide meaningful performance data regarding these features by irradiating the lead test assemblies in a manner which is representative of typical reload fuel. The description below provides additional discussion of fuel management plans, l core design margins and inspection and testing plans for the lead test assemblies.

Fuel Management Plans l

l The Company intends to irradiate the lead test assemblies in a manner typical of that employed in the low leakage cores designed for the North Anna reactors. This fuel management scheme typically involves irradiation in a relatively high power location during the first cycle; placement in a similar interior core location for the second cycle (sometimes under an RCCA); and use of the fuel near or on the periphery of the core in the third cycle, which provides a combination of shielding and most efficient use of the remaining fissionable material. The specific plans for use 1 of the lead test assemblies in accordance with this approach are provided below.

! In the first cycle of irradiation (Cycle 13), the proposed core loading pattern incorporates the lead

! test assemblies in the positions indicated in Figure 1 with 'LTA C13.' in these locations, the lead test assemblies will experience moderately severe duty typical of normal reload fuel. To ensure that the existing safety analyses based on the resident Westinghouse fuel remain applicable, calculation results for the proposed Cycle 13 loading pattem (see Tables 1a,1b, 2) have confirmed that the FCF assemblies are not in the highest fuel rod power density locations in the l core. In addition, the lead test assemblies do not have the highest cycle-averaged assembly average power. The lead test assemblies are not limiting in any key core operation parameters related to rod power density.

l This proposed loading pattem and its calculated results are subject to change in the event that a redesign is necessary for Cycle 13 operation. The Company willinform NRC if this should occur l so that the requirements for additional reporting of data for the redesigned core can be l established.

Figure 1 indicates by 'LTA C14A' and 'LTA C14B'some of the locations under consideration for i placement of the lead test assemblies during the second cycle of irradiation. These locations are typical positions for such once-bumed assemblies in North Anna core designs. It should be noted that locations denoted with 'LTA C14B' are under RCCAs. Placement under RCCAs

(which is typical for once-bumed assemblies) is being considered for the lead test assemblies. A final decision conceming the second cycle position for the lead test assemblies will be made during North Anna 1, Cycle 13 operation. This decision will consider both the advantage of I additional incore RCCA experience with this advanced fuel assembly design (beyond the data )

obtained from cont.ol rod drag testing discussed later) and the potential risks of such use in l

Page 1 of 8

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rodd:d cora locations. In th:ir third cycla, ths Isad t2st assembliss ara most likely to ba plac:d  ;

on the peripheral core locatior- c'mted with 'LTA C15' on Figure 1.

!. This description has been provided to indicate the proposed core management plans for use of l the FCF lead test assemblies. These plans are intended to provide the most meaningful data ,

, from the lead test assembly program, but are subject to change if required by overall core

management strategies.  !

i

Core Desian Marains 1

l In teleconferences between NRC and Company staff on November 19 and December 19,1996, j

calculated results from the North Anna 1, Cycle 13 core design were discussed. The results for two key core design parameters were presented: Enthalpy Rise Hot Channel Factor (FAH) and 1

Heat Flux Hot Channel Factor (FO). These discussions summarized the calculated values of 1 minimum margins available between the peak FAH and FO for the core and the FAH and FO Technical specification limits. Tables Ia,1b and 2 provide more complete calculated results which have been obtained from analysis of the Cycle 13 design containing the lead test

assemblies. i Table la provides calculated values and relevant margins for FAH as a function of bumup

! throughout Cycle 13, assuming the previous cycle (Cycle 12) shuts down at a bumup of 16100 MWD /MTU (referred to as low window). Data are presented for the peak calculated FAH of the Westinghouse fuel and the lead test assemblies. Percentage margin is presented for ,

comparison between: 1) Westinghouse fuel and the Technical Specifications limit; and 2) lead

' test assemblies and the Technical Specifications limit. Table 1b presents the same data assuming that Cycle 12 operates to a bumup of 17100 MWD /MTU (high window). The following

. summary is obtained from Tables la and 1b and corresponds to the values discussed during the November 19,1996 teleconference:  !

l Westinghouse fuel minimum margin to FAH limit = 2.5% at 11000 MWD /MTU i LTA minimum margin to FAH limit = 7.7% at 15000 MWD /MTU i Table 2 presents calculated parameter values and relevant margins for FO. These data were  !

l calculated with the approved Relaxed Power Distribution Control methodology (Reference 2)  !

l which involves calculations at three bumup values. Results are presented for both the low end '

and high end of the of Cycle 12 burnup window. Data are presented for the peak calculated FO t

of the Westinghouse fuel and the lead test assemblies. Percentage margin is presented for comparison between: 1) Westinghouse fuel and the Technical Specifications limit; and 2) lead test assemblies and the Technical Specifications limit. The following summary of the Table 2  !

data corresponds to the valaes discussed during the December 19,1996 teleconference:  ;

]

Westinghouse fuel minimum margin to FO limit (at the peak FO location) = j 6.5% at MOC bumup - High Window

LTA minimum margin to FO limit (at the peak FO location) =

10.6% at EOC bumup - High Window 4

The data in Tables 1a,1b and 2 are calculated for Cycle 13 operation. Calculated margins for future cycles in which the lead test assemblies are used will differ from the data presented for Page 2 of 8 i

Cycle 13. As indicated in both teleconferences, FAH and FO margin to the limit is generally greater.in the second and third cycles of irradiation.

l LTA Inspection and Testing During fabrication, the lead test assemblies were characterized to determine baseline values of dimensions and other features for later comparison with post-irradiation examination results.

l This characterization included, but was not necessarily limited to: certification of the composition

)

and material properties of the fuel rod cladding and guide thimble materials; dimensional characterization of the fuel rods, including cladding diameters and wall thicknesses, plus the lengths of all peripheral rods; dimensional characterization of the guide thimbles, including lengths and diameters; dimensional characterization of the fuel assemblies, including assembly lengths, fuel rod to top nozzle gaps, grid envelopes, grid elevations, and water channel measurements; and measurement of the force (rotation torque) required for the top nozzle quick disconnect mechanism.

Post irradiation examinations of the lead test assemblies will be performed during the program as permitted by the North Anna refueling schedule. In addition to fulllength visual examinations,

, we currently anticipate these examinations will include: measurement of fuel assembly length and bow, holddown spring compression testing, functional testing of the quick disconnect locking mechanism, oxide measurements on fuel rods and guide thimbles, measurements of fuel rod diameter, and shoulder gap measurements (fuel rod length determination).

The North Anna lead test assembly program has been structured from the outset to provide the most meaningful irradiation experience and relevant data with which to characterize performance  !

of the advanced fuel design features. This approach has resulted in a significant amount of )

analysis and evaluation to validate the design for use in the North Anna cores. In light of recent i control rod insertion problems at some units which may be related to fuel assembly guide thimble conditions, the NRC staff discussed our plans to perform RCCA control rod drag testing on the lead test assemblies to provide additional performance data for this advanced fuel design.

l Virginia Electric and Power Company has concluded that this additional testing would contribute

useful data which is fully consistent with the overall program objectives. Therefore, Virginia i Electric and Power Company will perform drag testing on the FCF lead test assemblies. The proposed approach would involve testing analogous to that which we have performed for the resident Westinghouse fuel assemblies during recent refueling outages. While it is intended to 1

obtain as much data as is reasonable, this testing (and other planned post-irradiation lead test I

assembly testing) will be subject to refueling outage schedule constraints. The current test plan l involves performing control rod drag tests on all four lead test assemblies at the end of each cycle of irradiation. An initial drag test of the lead test assemblies may also be performed prior to irradiation to establish a benchmark result for comparison of post-irradiation test results. l 1

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Refercnces

1. Letter, James P. O'Hanlon to U.S. Nuclear ihtulatory Commission, ,7'rginia Electric and Power Company; Nodh Anna Power Station Units No.1 and 2 - Notification of Intention to Use Lead Fuel Assemblies with Advanced Cladding Materials," Serial No.96-409, September 4,1996.

2. K. L. Basehore, et al., " Virginia Power Relaxed Power Distribution Control Methodology and Associated FO Surveillance Technical Specifications," Topical Report VEP-NE-1-A, March 1986.

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Table 1a Calculated Values of Normal Operation Enthalpy Rise Hot Channel Factor (FAH)'

For North Anna 1 Cycle 13 Core Design Containing FCF Lead Test Assemblies Low Window End of Cycle 12 Burnup (16100 MWD /MTU)

N1C13 W Fuel LTA Peak Margin (%)*' Margin (%)"

Burnup Peak FAH FAH (W to Limit) (LTA to Limit)

(MWD /MTU) 0 1.359 1.305 5.3 9.1 150 1.343 1.298 6.4 9.5 1000 1.326 1.300 7.6 9.4 3000 1.350 1.304 5.9 9.1 5000 1.366 1.304 4.8 9.1 7000 1.381 1.305 3.8 9.1 9000 1.393 1.306 2.9 9.0 11000 1.394 1.308 2.9 8.8 13000 1.383 1.313 3.6 8.5 15000 1.370 1.321 4.5 7.9 17000 1.356 1.322 5.5 7.9 18600 (EOR) 1.343 1.319 6.4 8.1

' Unrodded Hot Full Power Results - LTAs will exhibit more margin with control rods at the rod insertion limits

• Limit value for FAH is 1.435

• Margin (%) to the limit is calculated as follows:

[Value - Limit]

Margin (%) = x 100 Limit Page 5 of 8

j l

l Table 1b

]

j Calculated Values of Normal Operation Enthalpy Rise Hot Channel Factor (FAH)'

For North Anna 1 Cycle 13 Core Design Containing FCF Lead Test Assemblies i High Window End of Cycle 12 Bumup (17100 MWD /MTU)

N1C13 W Fuel LTA Peak Margin (%)" Margin (%)

Bumup Peak FAH FAH (W to Limit) (LTA to Limit)

(MWD /MTU) 0 1.349 1.314 6.0 8.4 150 1.337 1.302 6.8 9.3 1000 1.337 1.303 6.8 9.2 3000 1.353 1.306 5.7 9.0

., 5000 1.370 1.306 4.5 9.0

7000 1.386 1.307 3.4 8.9 9000 1.398 1.308 2.6 8.8 1

11000 1.399 1.310 2.5 8.7 13000 1.389 1.317 3.2 8.2

15000 1.376 1.324 4.1 7.7 l 17000 1.362 1.325 5.1 7.7

18100 (EOR) 1.353 1.324 5.7 7.7

' Unrodded Hot Full Power Results - LTAs will exhibit more margin with control rods at the rod insertion  ;

limits

• Limit value for FAH is 1.435

• Margin (%) to the limit is calculated as follows:

[Value - Limit]

Margin (%) = - x 100

Limit i

Page 6 of 8

Table 2 Calculated Values of Heat Flux Hot Channel Factor (FO)

For North Anna 1 Cycle 13 Core Design Containing FCF Lead Test Assemblies (Values Presented are Normal Operation FO)'

N1C13 Burnup W Fuel LTA Margin (%)*' Margin (%)

(Time in Cycle- Peak FO Peak FQ (W to Limit) (LTA to Limit)

Window)

BOC - Low 1.9156 1.8645 12.5 14.9 MOC - Low 2.0406 < 1.9434 6.8 > 11.3 EOC - Low 1.9787 1.9494 9.6 11.0 BOC - High 1.9461 1.8680 11.1 14.7 MOC - High 2.0466 < 1.9491 6.5 > 11.0 EOC - High 1.9975 1.9573 8.8 10.6

' Includes all Relaxed Power Distribution Control power shapes within assumed Al operational limits

• Limit value for FO is 2.19

• Margin (%) to the limit is calculated at the point of peak FO as follows:

[Value - Limit] j Margin (%) = x 100 '

Limit l

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NORTH ANNA UNIT 1 Figure 1 -

R P N M L K J H G F E D C B A LTA C 15 1

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LTA c ia 3

LTA

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.- . .;.y 7 LTA

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LTA LTA 1LTA{ hLTAJ.. LTA LTA LTA C 15

• C 13 C 14A C 148 iC 14B:: C 14A C 13 ' M '- C 15 8 u_ ,

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12 LTA C ta ' '

13 14

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LTA c 15 15 Note: RCCA Locations are shaded 8 of 8

_________ .- __ - _ - _ _ --.- ___-__ - _-_____ _ -. . _ _ - _ _ _ . _ - - - _ . ._ . - _ - . .__-- -_