ML100500476

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Areva Fuel Transition Technical Specification Change Request Submittal
ML100500476
Person / Time
Site: Browns Ferry Tennessee Valley Authority icon.png
Issue date: 02/17/2010
From:
Tennessee Valley Authority
To:
Office of Nuclear Reactor Regulation
Bailey S, NRR/ADRO/DORL, 415-1321
References
Download: ML100500476 (28)
v • d • e


Text

1m Browns Ferry Nuclear Plant, Unit 1 AREVA Fuel Transition Technical Specification Change Request Submittal February 17, 2010 1

1m Agenda

  • Submittal Background

- Automatic Depressurization System (ADS) Single Failure

- Standby Liquid Control System (SLCS) Shutdown Margin (SDM)

Analysis Uncertainties

- Lack of Inclusion of Stability Topical Reports in Technical Specifications

- Rod Drop Analysis (RDA) Parameterization

  • Resubmittal Plans 2

~ Submittal Background

  • Browns Ferry Nuclear Plant (BFN) Unit 1 Technical Specification Change Request

- To Allow Use of AREVA Fuel, Methodologies, and Blended Low Enriched Uranium Similar to Change Previously Approved for BFN Units 2 and 3 Initially Performed for Extended Power Uprate (EPU) Conditions

  • 1200/<) of Original Licensed Thermal Power (OLTP)

Submitted October 23, 2009

  • Addressed Previous Applicable NRC Requests for Information

- Tennessee Valley Authority (TVA)/NRC Meetings January 28,2009, and March 16,2009

- Supplemented November 17, 2009

  • To address Current Licensed Thermal Power (CLTP) - 105% OLTP

- NRC Completed Acceptance and Identified Need for Supplemental Information to Acceptance Submittal for Review on December 23, 2009

- TVA Responded with Supplemental Information on January 15, 2010

- NRC Issues with Supplemental Information Identified and TVA Withdrew Submittal on February 2, 2010 3

1m Submittal Issues

  • Issues with TVA Responses to 4 Questions

- ADS Single Failure

- SLCS SDM Analysis Uncertainties

- Lack of Inclusion of Stability Topical Reports in Technical Specifications

- RDA Parameterization 4

~ ADS Single Failure Question

  • NRC Questioned why Single Failure of ADS, or of a Single ADS valve, is Not Considered

- BFN ADS is not Single Failure Proof

- No Single Failure can Disable Both ADS and High Pressure Coolant Injection (HPCI)

System

- There is a Single Battery Failure that can Disable ADS Logic 5

Ii!lJ Browns Ferry Nuclear Plant ADS Design

  • Both ADS Logic Trains are Powered by Same Reactor Motor Operated Valve (RMOV) Board
  • Battery Failure, Supplying Power to RMOV Board B, Leads to Failure of Automatic ADS Function

- 4 ADS Valves can Still be Manually Actuated

- HPCI System is Operable for this Battery Failure

  • RMOV Board A Failure Disables HPCI System, but Leaves ADS Operable
  • Single Failure Concern had been Previously Identified by TVA

- Evaluated by General Electric (GE) in 1980 6

Ii!il ADS Single Failure Evaluation History

- Evaluated Battery Failure Disabling Automatic ADS Initiation

- SAFE/REFLOOD Loss of Coolant Accident (LOCA) analysis at 100°A>

OLTP

- Considered Both Recirculation and Feedwater Line Breaks

- Assumed Manual Opening of 4 ADS Valves at 10 minutes

- Limiting Recirculation Break was 0.3 ft2

  • Board B Failure was Limiting Single Failure

- Results Influenced by Crediting of Low Pressure Coolant Injection (LPCI)

- Report was Submitted to NRC on March 26, 1986 7

Ii!lJ ADS Evaluation History (continued)

  • BFN Transitioned to SAFER/GESTR in 1995

- Initial SAFER/GESTR report

- 105°~ OLTP

- Single Failure Table Shows RMOV Board A as Limiting Single Failure

  • Prior Evaluations Showed RMOV Board B (i.e., ADS single failure) as Limiting.

- Single Failure Table Footnote for Battery Failure States ADS Functions per Previous Analysis Report

- Credit for LPCI Removed

- Single Failure Table Footnote Re-worded

  • Stated 4 ADS Valves Would be Available with HPCI

- RMOV Board B Single Failure Case

  • Stated This Case is Bounded by 5 ADS Valves Available without HPCI

- RMOV Board A Single Failure Case 8

U!lJ ADS Evaluation History (continued)

- UFSAR Table 6.5-3 No Longer Addressed

  • Manual versus Automatic ADS Function
  • RMOV Board B Impact on ADS Function

- Single Failure Table Footnote Revised to Remove All Discussion Related to RMOV Board B Failure

  • General Electric Hitachi (GEH) Review of Current SAFER/GESTR Analyses (January 2010)

- Concluded ADS Single Failure Issue Not Adequately Addressed 9

1m ADS Evaluation History (continued)

  • AREVA LOCA Analysis (December 2003)

- Single Failures Considered

  • Consistent with TVA Input and UFSAR Table 6.5-3

- UFSAR Table 6.5-3 Indicates Limiting Battery Failure Case is Loss of HPCI System with ADS Operable

- Based on these TVA Inputs, AREVA LOCA Analysis was Not Required to Consider Battery Failure Case which Disables ADS, but leaves HPCI System Operable

  • From Inputs, ADS Appeared to be Single Failure Proof 10

[ml Preliminary Root Cause Information

  • Initial SAFER/GESTR Validation that ADS Single Failure Scenario was Non-limiting

- Limited Cases Run

- Removal of LPCI May Have Altered Relationship Between RMOV Board A and RMOV Board B Single Failures

- Limiting Break Size May Have Changed with New Fuel Introductions and Power Uprates

  • TVA Review of SAFER/GESTR Reports was Not Adequate 11

Ii!i! Preliminary Root Cause Information (continued)

  • TVA Did Not Adequately Update UFSAR Text as Issue Evolved

- Single Failure Table ADS Footnote Adopted Wording from SAFER/GESTR Revision 1 Report

  • Led to Overly Simplified Single Failure Table

- Failure Effects by Individual Battery Not Listed

- Previous GE Analysis Reports Reflecting ADS Single Failures Not Referenced

- Discussion of ADS Does Not Fully Describe Battery Failure Effects

  • Alternate Battery Failure Scenarios Not Considered in Current AREVA LOCA Analysis

- Used UFSAR Single Failure Table, which is Incomplete 12

Ii!4l ADS Issue Corrective Actions

  • TVA Will Provide Both GEH and AREVA LOCA Addendum Reports by April 2010

- Based on Expanded Emergency Core Cooling System (ECCS)

Equipment Availability Table

- Will Address ADS Single Failure with Current ADS Configuration

- Will Include Specific Analysis of HPCI System Line Breaks

- EPU Based with Additional CLTP Cases

  • CLTP Cases Will Include a Spectrum of Break Sizes

- Will Credit Technical Specification HPCI System Flow Rate and Faster HPCI System Start Time

  • LOCA Addendum Reports Included as Part of Resubmittal 13

[i!JJ ADS Issue Corrective Actions (continued)

  • TVA Investigating Potential ADS Modifications
  • Potential Modifications
1. Single Failure Proof ADS with 6 Automatically Initiated ADS Valves
2. Single Failure Proof ADS with 4 Automatically Initiated ADS Valves
3. Same as #1 , with Addition of ADS Timer Modification
4. Same as #2, with Addition of ADS Timer Modification
  • ADS Timer Modification BFN ADS Requires Low Pressure ECCS Running to Start ADS timer Changing Logic so Timer Starts on Just Water Level and Drywell Pressure Would Start Timer Significantly Earlier
  • Modifications 2, 3, and 4 Would Require LOCA Reanalysis 14

lim ADS Issue Corrective Actions (continued)

  • Extent of Condition

- TVA has Validated All Other Parameters and Single Failure Assumptions Provided to Vendors for LOCA and Transient Analyses

  • Created Expanded ECCS Equipment Availability Table

- No Other Issues Found

  • Planned Actions to Prevent Recurrence

- Revise Licensing and Design Basis Documentation to Properly and Consistently Document ADS Design Relative to Single Failure

  • BFN Design Basis Documents 15

um SLCS SDM Uncertainties Issue

  • Concerns

- Uncertainties Discussed were Not Comprehensive

- Application of Manufacturing Tolerances to GE14 Not Addressed

  • AREVA Considers All Uncertainties Associated with Performing SLCS SOM Analysis Using the NRC approved CASMO-4/MICROBURN-B2 Methods

- Manufacturing Tolerances and Calculational Uncertainty

- Cold Critical Measurements

- Comparison to Higher Order Calculation

- Cross Section Interpolation 16

~ 8LC8 8DM Uncertainties Issue (continued)

Specific uncertainty values used Uncertainty One sigma K95/95 X RMS (K95/95 X RMS)2 value

~k ~k2

~k Manufacturing and 0.00191 0.0038 1.4440E-05 Calculation Cold critical 0.00149 0.0033 1.0890E-05 Casmo-4 vs. MCNP 0.00284 0.0070 4.9000E-05 X-sec interpolation 0.00057 0.0016 2.5600E-06 Sub Total 7.6890E-05 Total uncertainty = Square root of 7.6890E-05 = 0.0088 ~k = 0.88 %~k 17

rnJ SLCS SDM Uncertainties Issue (continued)

  • Manufacturing Variability

- AREVA has Performed a Specific Analysis for ATRIUM-10 fuel

  • One Sigma Value of 0.00123 ~k Determined

- GE Historical Value is 0.00127 ~k (from NEOO-2481 0)

  • Not a GE14 Specific Value
  • Note - GE Does Not Directly Account for this Uncertainty in SLCS SDM

- AREVA Conservatively Uses GE Historical Value in Their SLCS SOM Uncertainty Analysis for All Fuel Types

  • GE Historical Value Bounds AREVA Determined Value

- Calculational Uncertainty Accounted For

  • AREVA has Performed a Specific Analysis to Determine a Value of 0.001 0 ~k
  • GE Historical Value for Fresh Core is 0.00142 ~k (NEDO-24810)
  • GE Historical Value for Exposed Core is 0.00132 ~k (NEDO-2481 0) 18

Im1 SLCS SDM Uncertainties Issue (continued)

  • Manufacturing Variability

- Combined Uncertainty

  • AREVA Method Treats Combined Uncertainty Conservatively

- Effect is Also Inherently Captured in Cold Critical Uncertainty

- Uncertainty is Effectively Being Captured Twice

  • AREVA Method is More Conservative than Global Nuclear Fuels Method and is Therefore Conservative for GE 14 Co-resident Fuel 19

um SLCS SDM Uncertainties Issue (continued)

  • Plant Measurement Uncertainty

- Analysis does Not Depend on Any Measurement of Plant Process Parameters

  • Worst Case Temperature Used - Measurement Uncertainty is Not Relevant
  • Operation Uncertainty

- Prior Cycle Variability Accounted For (Nominal, Short, and Long)

  • Uncertainty Due to Control Blade Burnup

- Analysis Generally Done with All Rods Out

- For Cases with Rods, Burnup Effects are Indirectly Captured Through Cold Critical Eigenvalue Benchmarking 20

lim SLCS SDM Uncertainties Issue (continued)

  • Conclusions

- All Key Uncertainties are Considered

- Manufacturing Uncertainty Treatment is Conservative

  • Applies to Both ATRIUM-10 and GE14 Fuel Types

- SLCS SDM Analysis Method is Conservative in Treatment of Uncertainty 21

Lack of Inclusion of Stability Topical Reports 1m in Technical Specifications

  • Issues

- Technical Specification 5.6.5.b does Not Include AREVA Stability Topical Reports

- Technical Specifications do Not Capture the Oscillation Power Range Monitor (OPRM) Setpoint

  • Actions to Resolve

- Revise Technical Specification 3.3.1.1 (Reactor Protection System Instrumentation) for OPRM Upscale Function to Indicate

  • OPRM Period Based Limits are Included in Core Operating Limits Report (COLR)

- Revise Technical Specification 5.6.5.a to Include OPRM Setpoint as COLR Item

- Revise Technical Specification 5.6.5.b to Include the Following Stability Related Topical Reports

  • EMF-CC-074(P)(A) - STAIF
  • BAW-10255(P)(A) - DIVOM using RAMONA5-FA 22

rnJ RDA Parameterization

  • Concerns

- Insufficient Detail on How Mass and Modern Fuel Designs are Captured

- Studies that Validated the Parameterized Function Not Provided

  • Parameterized Functions are Unchanged from Those Functions Provided in Topical Report XN-NF-80-19 (P)(A) Vol. 1 Supplement 2.

- The Generic Curves have been Validated for GE-11 and ATRIUM-1 0 Specific Calculations using the Methodology Described in XN-NF-80 19 (P)(A) Vol. 1 Supplement 2 23

[m1 RDA Parameterization (continued)

Enthalpy vs Rod Worth ATRIUM-10 Fuel C)

ClS u

Q.

ClS

.s::

s:::::

W 5 6 7 8 9 10 11 12 13 14 15 16 Rod Worth, mk

[ ~ c;en~r~m-~BOC 10. 7~OP~-MO~u1 O. 5dop ---*--EOC11 .2dop I 24

[m RDA Parameterization (continued)

Enthalpy vs Rod Worth GE-11 Fuel

~

ca

(,)

~ca

.c W

I:

-j 5 6 7 8 9 10 11 12 13 14 15 16 17 18 Rod Worth, mk L_~~- Generic.---- MOC 11 10 .Odop 25

lim Doppler Coefficient Vs. Fuel Weight CRDA Doppler Coefficients

-9.40E-06 GE-11/13

-9.60E-06

-9.80E-06

<,:: GE-14 a;

'u

!i Cl... -1.00E-05 Ql ATRIUM-9 ATRIlJM-1n c

o Q

-1.02E-05

-1.04E-05 9X9-2 8x8-2

-1.06E-05 162 164 166 168 170 172 174 176 178 Lattice Weight 26

.. I'!".

  • Ii!i! RDA Parameterization (continued)
  • GE-14 Fuel Design is Very Similar to ATRIUM-1 0 Fuel Design in Terms of Fuel Diameter and Nodal Weight Based Upon the Upper Lattice
  • Concl usion

- The Generic Parameterization Function is Applicable to Modern Fuel Designs

- GE14 Would be Expected to Show Similar Behavior Based on Similarity of Fuel Characteristics with Fuel Designs Studied 27

U!4! Resubmittal Plans

  • Schedule
  • Thermal Power Level
  • Scope

- Previous Submitted Documentation

- Four Acceptance Review Issues

- Other Issues

  • Level of Detail 28
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