ML023380021
| ML023380021 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/21/2002 |
| From: | Rosalyn Jones Duke Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML023380021 (17) | |
Text
I Duke R. A JONES
- wPower, Vice President A Duke Energy Company Duke Power 29672 / Oconee Nuclear Site 7800 Rochester Highway Seneca, SC 29672 864 885 3158 864 885 3564 fax November 21, 2002 U.
S. Nuclear Regulatory Commission Washington, D.
C.
20555 Attention: Document Control Desk
Subject:
Oconee Nuclear Station Docket Numbers 50-269, 270, and 287 Supplement to License Amendment Request for Full Scope Implementation of the Alternate Source Term Technical Specification Change (TSC) Number 2001-07 In a letter dated October 16, 2001, Duke Energy (Duke) submitted a license amendment requesting approval of the Alternate Source Term (AST) analysis methodology for Oconee Nuclear Station.
Duke received questions from the NRC related to the AST LAR and discussed these questions with the NRC in a March 21, 2002 meeting.
A common understanding of the questions and required responses were obtained, and an initial supplement to the LAR was submitted on May 20, 2002.
Follow-up discussions indicated that some additional information would be helpful for the NRC review.
That information was submitted as supplement 2 on September 12, 2002.
Information provided in the LAR describes preliminary design information.
Some variation is expected in the final design and field installation.
In a conference call on November 19,
- 2002, Duke discussed the preliminary versus final design with the NRC.
Duke made assumptions and performed additional analyses that demonstrates the effect of changes in final design values on dose analysis results.
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www duke-energy.corn
U.
S. Nuclear Regulatory Commission November 21, 2002 Page 2 of 5 The staff has requested that this information be submitted for further consideration.
It is included as Attachment 1. contains the revised X/Q values.
contains the dose results for the Loss of Coolant Accident and Fuel Handling Accident.
The submittal dated May 20, 2002 contained a typographical error in the offsite dose tabulated for one Fuel Handling Accident case.
Case 7 in Duke's response to Request for Additional Information (RAI) 18, which represents a Transport Cask drop in the Unit l&2 Spent Fuel Pool, was previously reported as 1.2 rem TEDE at the EAB.
This dose should have been reported as 1.8 rem TEDE at the EAB.
A revised page is provided in Attachment 4.
Duke committed to three modifications in the AST submittal dated 5/20/02.
These modifications included:
a dual air intake system to the Control Room; a high pressure/low pressure injection relief valve discharge to the reactor building emergency sump; and a passive caustic addition system.
The technical specifications will be fully implemented upon completion of these modifications on all three Oconee units.
Pursuant to 10 CFR 50.91, a copy of this proposed license amendment is being sent to the State of South Carolina.
If there are any questions regarding this submittal, please contact Reene' Gambrell at (864) 885-3364.
Ver tru y yours, R. A. Jones, Vice President Oconee Nuclear Site
U.
S. Nuclear Regulatory Commission November 21, 2002 Page 3 of 5 cc:
Mr.
L.
N. Olshan, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 0-14 H25 Washington, D.
C.
20555 Mr.
L.
A. Reyes, Regional Administrator U.
S. Nuclear Regulatory Commission - Region II Atlanta Federal Center 61 Forsyth St.,
SW, Suite 23T85 Atlanta, Georgia 30303 Mr. M. C.
Shannon Senior Resident Inspector Oconee Nuclear Station Mr. Virgil R. Autry, Director Division of Radioactive Waste Management Bureau of Land and Waste Management Department of Health,& Environmental Control 2600 Bull Street Columbia, SC 29201
U.
S. Nuclear Regulatory Commission November 21, 2002 Page 4 of 5 R. A. Jones, being duly sworn, states that he is Vice President, Oconee Nuclear Site, Duke Energy Corporation, that he is authorized on the part of said Company to sign and file with the U. S. Nuclear Regulatory Commission this revision to the Renewed Facility Operating License Nos.
DPR-38, DPR-47, DPR-55; and that all the statements and matters set forth herein are true and correct to the best of his knowledge.
R.. A. JtU, Vicne President Oconee uclear Site Subscribed and sworn to before me this L
day of
- k 2002 Notary Public My Commission Expires:
My Commission Expires Aug. 19, 204 S.~~--. '
U.
S. Nuclear Regulatory Commission November 21, 2002 Page 5 of 5 bcc: B.
H. Hamilton D.
A. Baxter W. W. Foster T.
D. Curtis C.
W. Boyd R.
A. Jones P.
M. Abraham G.
B. Swindlehurst L.
F. Vaughn C. J.
Thomas -
MINS L. E. Nicholson L. V. Wilkie D. B.
Coyle R. T. Repko M. T.
Cash -
GO G. D.
Gilbert CNS J.
E. Burchfield S.
P. Schultz C.
B.
Taylor D.
M. Hubbard W. B.
Edge M. C.
Bolch J.
T. Fuller D.
F.
Rohrer D.
G.
Moore
- NSRB, EC05N
- ELL, EC050 File T.S. Working BWOG Tech Spec Committee (5)
ONS Document Management
ATTACHMENT 1 Final Design Analysis Final Design Analysis Information provided in Duke Energy's License Amendment Request (LAR) submittal that describes the features and analytical impact of future plant modifications is based on preliminary design information; therefore, some variation is expected in the final design and field installation.
Once the modifications are installed and post-modification testing has been performed, changes in X/Q values and other modeling values along with the resultant change in the calculated doses will be reflected in the Oconee UFSAR update that implements this change.
Also, any flow imbalance or other operational differences identified in post-modification testing of the dual control room air intakes will also be addressed in this analysis of the installed modification.
As described in Duke's May 20, 2002 response to Request for Additional Information (RAI) 12, the selection of bounding values from the 1998 testing for unfiltered inleakage for the analyses as submitted in the LAR provides Duke with margin to accommodate changes in input assumptions that could be required to account for possible plant operational changes, such as increases in ECCS system leakage flow, imbalances in ventilation system flow rates, or reductions in filtration efficiencies.
Duke has concluded that the appropriate input values for unfiltered inleakage as derived from the tracer gas test results should correspond to the nominal values determined from the testing programs.
For the UFSAR analysis, these nominal inleakage values will partially or wholly offset any increases in dose due to final design implementation and operational differences.
As a demonstration, we have evaluated a change of inputs for control room unfiltered inleakage, intake location, and airflow imbalance between dual intakes.
For this demonstration, we have chosen the 1998 nominal values for unfiltered inleakage into the control room, with an addition of 10 cfm allowance for unfiltered inflow due to Control Room ingress and egress during the course of an accident.
To represent a variation in planned installation, we have chosen a control room intake location which is 10 feet higher (closer to the unit vent release) than in our LAR submittal.
This variation causes a dose Page 1
increase of approximately 1%.
To represent an operational difference, we have chosen a 60% / 40% flow imbalance between the dual control room intakes, which is based on operational experience with dual intakes at Catawba Nuclear Station.
This variation causes a dose increase of approximately 20%.
Calculated doses for the LOCA and the limiting Fuel Handling Accident case were evaluated using these assumptions.
The results of this demonstration are shown in the table below.
CR Intake Unfiltered Height of LOCA Limiting FHA Air Inleakage Control Room Control Control Room Imbalance Before / After Intake Room Dose Dose (rem Booster Fans (rem TEDE)
TEDE)
(cfm)
Balanced 1150 / 150 (1) Preliminary 3.2 2.8 design 60 / 40 1075 / 90 10' higher 2.7 3.4 Imbalance (2) than preliminary design (1) Bounding values selected from the 1998 CR testing (with 10 cfm for ingress/egress)
(2)
Nominal value results from the 1998 CR testing (with 10 cfm for ingress/egress)
As demonstrated, the LOCA analysis with nominal unfiltered inleakage values results in a dose that is lower than previously reported.
The lower inleakage values more than offset the increases in dose from operational differences.
For the limiting FHA event, the increased dose is partially offset by nominal inleakage values, but the calculated dose is higher than previously reported primarily due to the assumed CR intake air imbalance.
The limiting FHA is considered extremely unlikely, and it involves the assumption of cladding damage and gap activity release from hundreds of assemblies stored in the spent fuel pool by a transport cask drop onto the pool.
The calculated FHA dose remains within the regulatory limits.
Duke Energy Corporation will assure that the UFSAR will provide analysis results derived from the final design and will document the demonstrated margins to limits for all Page 2
events.
A revised set of values as presented in tables provided in Duke's response to Request for Additional Information (RAI) 18 for the FHA dose response and in Section I of the Duke Energy Corporation Supplement to the License Amendment Request (letter, dated September 12, 2002) for the LOCA dose response is provided in Attachment 3.
Page 3
ATTACHMENT 2 Revised X/Q Values
ATTACHMENT 2 Revised X/Q Values Revised X/Q values which represent a 60/40 airflow imbalance in dual control room air intakes, and an intake location that is 10' higher that preliminary design.
Release Point Type Bounding X/Q value Vent Release 0 to 2 hr 5.38E-04 2 to 8 hr 3.74E-04 8 to 24 hr 1.57E-04 1 to 4 days 1.24E-04 4 to 30 days 1.01E-04 Equipment Hatch Release 0 to 2 hr 3.73E-04 2 to 8 hr 2.90E-04 8 to 24 hr 1.23E-04 1 to 4 days 9.30E-05 4 to 30 days 7.20E-05 Fuel Handling Building Roll-up Door Release 0 to 2 hr 1.72E-04 2 to 8 hr 1.37E-04 8 to 24 hr 5.75E-05 1 to 4 days 4.43E-05 4 to 30 days 3.36E-05 BWST Release 0 to 2 hr 2.32E-04 2 to 8 hr 1.76E-04 8 to 24 hr 7.26E-05 1 to 4 days 5.64E-05 4 to 30 days 4.42E-05
ATTACHMENT 3 Dose Results for Loss of Coolant Accident and Fuel Handling Accident
ATTACHMENT 3 Final Design Values for LOCA and FHA Dose results with a 60 / 40 airflow imbalance, and a control room intake location 10' higher than preliminary design.
Nominal 1998 unfiltered inleakage values (plus 10 cfm for doors) were used to offset the increase in dose due to airflow imbalance.
Tables also contain a correction of a typographical error in EAB dose for Case 7.
LOCA Calculated Doses Containment RBES Model Total TEDE Model (rem TEDE)
(rem)
EAB 8.6 0.2 8.8 LPZ 1.6 0.1 1.7 Control Room 2.2 0.5 2.7 Page 1
- 1 Calculated Doses~to Control Room Operators due to Fuel Handling Events Spent Fuel Pool (SFP) and Containment Case Group Source Release Control Room TEDE Point Unit (rem)
Destination 1
1 Fuel Assembly Any Unit Unit 1&2 2.2 Accident in Vent Either SFP 2
1 Fuel Assembly Either Unit 1&2 0.7 Accident in Roll-Up Either SFP Door 3
1 Fuel Assembly Any Unit Unit 3 1.4 Accident in Vent Either SFP 4
1 Fuel Assembly Either Unit 3 0.4 Accident in Roll-Up Either SFP Door 5
2 Fuel Assembly Any Unit Unit 1&2 1.3 Accident in Any Vent Containment 6
2 Fuel Assembly Any Unit Unit 3 0.8 Accident in Any Vent Containment 7
3 Transport Cask Any Unit Unit l&2 3.4 Drop in Unit Vent 1&2 SFP 8
3 Transport Cask Any Unit Unit 3 2.2 Drop in Unit Vent 1&2 SFP 9
3 ISFSI Cask Drop Any Unit Unit 1&2 1.5 in Vent Unit 1&2 SFP 10 3
ISFSI Cask Drop Any Unit Unit 3 0.9 in Vent Unit 1&2 SFP 11 3
ISFSI Cask Drop Either Unit l&2 0.5 in Roll-Up Unit 3 SFP Door 12 3
ISFSI Cask Drop Either Unit 3 0.3 in Roll-Up Unit 3 SFP Door Page 2
Page 3 Calculated Offsite Doses due to Fuel Handling Events Spent Fuel Pool (SFP) and Containment Case Group Source Release EAB LPZ Point TEDE (rem)
Fuel Assembly Any Unit 1.2 0.1 Accident in Vent Either SFP 2
1 Fuel Assembly Either 1.2 0.1 Accident in Roll-Up Either SFP Door 5
2 Fuel Assembly Any Unit 0.7 0.1 Accident in Any Vent Containment 7
3 Transport Cask Any Unit 1.8 0.2 Drop in Unit 1&2 Vent SFP 9
3 ISFSI Cask Drop Any Unit 0.8 0.1 in Vent Unit 1&2 SFP 11 3
ISFSI Cask Drop Either 0.8 0.1 in Roll-Up Unit 3 SFP Door
ATTACHMENT 4 Revised Page from Submittal dated May 20, 2002
U. S. Nuclear Regulatory Commission May 20, 2002 Page 68 Calculated Doses to Control Room Operators due to Fuel Handling Events Spent Fuel Pool (SFP) and Containment Case Group Source Unit and Control Room Unit TEDE Release Point Destination (rem) 1 1
Fuel Assembly Accident Unit 2 Unit 1&2 1.8 in SFP Unit Vent 2
1 Fuel Assembly Accident Unit 3 Unit 1&2 0.6 in SFP Roll-Up Door 3
1 Fuel Assembly Accident Unit 2 Unit 3 1.2 in SFP Unit Vent 4
1 Fuel Assembly Accident Unit 3 Unit 3 0.4 in SFP Roll-Up Door 5
2 Fuel Assembly Accident Unit 2 Unit 1 &2 1.0 in Containment Unit Vent 6
2 Fuel Assembly Accident Unit 3 Unit 3 0.7 in Containment Unit Vent 7
3 Transport Cask Drop in Unit 2 Unit 1&2 2.8 SFP Unit Vent 8
3 Transport Cask Drop in Unit 2 Unit 3 1.9 SFP Unit Vent 9
3 ISFSI Cask Drop in SFP Unit 2 Unit 1 &2 1.2 Unit Vent 10 3
ISFSI Cask Drop in SFP Unit 2 Unit 3 0.8 Unit Vent 11 3
ISFSI Cask Drop in SFP Unit 3 Unit 1 &2 0.4 Roll-Up Door 12 3
ISFSI Cask Drop in SFP Unit 3 Unit 3 0.3 Roll-Up Door Calculated Off site Doses due to Fuel Handling Events Spent Fuel Pool (SFT) and Containment Case Group Source Unit and EAB LPZ Release Point TEDE (rem)
Fuel Assembly Accident Unit 2 1.2 0.1 in SFP Unit Vent 2
1 Fuel Assembly Accident Unit 3 1.2 0.1 in SFP Roll-Up Door 5
2 Fuel Assembly Accident Unit 2 0.7 0.1 in Containment Unit Vent 7
3 Transport Cask Drop in Unit 2 1.8 0.2 SFP Unit Vent 9
3 ISFSI Cask Drop in SFP Unit 2 0.8 0.1 Unit Vent 11 3
ISFSI Cask Drop in SFP Unit 3 0.8 0.1 I_ _
Roll-Up Door