ML18302A092
| ML18302A092 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/29/2018 |
| From: | Hughes D Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18302A092 (3) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 October 29, 2018 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 10 CFR 50.4 10 CFR 50.54(q) 10 CFR 50, Appendix E 10 CFR 72.44(f)
Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket Nos. 50-259, 50-260, 50-296, and 72-052
Subject:
Browns Ferry Nuclear Plant - Site Emergency Plan Implementing Procedure Revision In accordance with the requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.54(q); 10 CFR 50, Appendix E; and 1 O CFR 72.44(f), the Tennessee Valley Authority (TVA) is submitting a description of changes to the Browns Ferry Nuclear Plant (BFN) Radiological Emergency Plan. The affected document is the BFN Emergency Plan Implementing Procedure (EPIP) named below.
EPIP EPIP-1 Revision 0057 Emergency Classification Procedure Description of Changes and Summary of Analysis Effective Date 09/28/2018 EPIP-1, Revision 57, was revised to incorporate changes to Initiating Condition (IC) RU1 [Release of gaseous or liquid radioactivity greater than 2 times the Offsite Dose Calculation Manual (ODCM) limits for 60 minutes or longer] to correct the values used during Emergency Classification Level (ECL) determination. Several changes were implemented in the table in step (2). This revision replaced "14,336 cps" with "3.0E+5 µCi/s" for radiation monitors (RM) 249, 250, 251, and 252 to ensure that the values can be accurately read and understood. This revision also removed "Offgas Post-Treatment Radiation Monitor (1, 2, 3-RM-90-265/266)" and its associated threshold because this instrument was previously off scale. The units for the GE Stack Gas Radiation Monitor (RM-90-147/148) was corrected from "cpm" to "cps" to accurately reflect the threshold value used
U.S. Nuclear Regulatory Commission Page 2 October 29, 2018 in EGL determination. The value for Offgas Pretreatment Radiation Monitor (1, 2, 3-RM-90-157) was corrected from "5,096" to "4.2E+05" based on the information found in O-Tl-15 (Radioactive Gaseous Effluent Engineering Calculations and Measurements). The period between 1 and 2 in
"(1, 2, 3-RM-90-157)" was replaced with a comma. These activities were evaluated using EPDP-17, Attachment 2 (Screening Evaluation Form) to determine whether a 10 CFR 50.54(q)
Reduction in Effectiveness Evaluation was required. The activities were determined to be a change to the radiological emergency plan that was not editorial or typographical. Upon completion of Attachment 2, it was determined that the activity could not be implemented without performing a 10 CFR 50.54(q) Reduction in Effectiveness Evaluation; therefore, EPDP-17, was completed. It was determined that the changes did not negatively impact the timeliness with which an EGL can be declared. Replacing the values for RMs 1, 2, 3-RM-90-249, -250, -251, and -252 do not change the threshold values as it relates to activity levels associated with the potential condition. These changes align the values in the table with the instrument readout in the Main Control Room (MGR) to ensure the value can be more accurately read and understood. The removal of the Offgas Post-Treatment RM for which the threshold value is no longer on scale does not negatively impact the ability of the operators to declare the event since the Emergency Action Level (EAL) value associated with this monitor is not within the usable response and display range of the instrument. Additionally, the Offgas Pretreatment RM is not removed as part of the change; therefore, an alternate EAL threshold for this effluent path continues to exist. Correcting the units for the Stack Gas RM ensures that the values contained in the EAL more accurately reflect the threshold values used in EGL determination. Correcting the value for the Offgas Pretreatment RM ensures that the value represented in the EAL more accurately reflects the approved value of 2 times the set point established by the ODCM, as discussed in O-Tl-15 (Radioactive Gaseous Effluent Engineering Calculations and Measurements).
Replacing the period with a comma to separate unit designators associated with the Offgas Pretreatment RM is considered editorial/typographical in nature. There is no change in assignment of responsibility and therefore no adverse affect on how the emergency classification/declaration function is performed by plant operators. In conclusion, the implemented changes improve the effectiveness of the Radiological Emergency Preparedness (REP) by aligning thresholds associated with this IC with correct design outputs and instrument readouts available in the MGR.
Therefore, the activities were determined to not constitute a reduction in effectiveness and were implemented without prior approval.
This revision also proposed changes to the Fission Product Barrier (FPB) Matrix in EPIP-1,
Revision 57. The threshold wording was edited by adding the word "ANY" to clarify that any RM reading greater than the values meets the threshold. The Unit 3 values for all three FPBs were updated based on implementation of extended power uprate (EPU). The Reactor Coolant System (RCS) Barrier Loss 4.A (Primary Containment Radiation) threshold bases was revised to reflect design output from calculation NDQ0090930050. The reference to O-Tl-88 was removed since the basis of the EAL is more appropriately described in calculation NDQ0090930050. The RCS Barrier Loss 4.A threshold was revised to remove an incorrect reference to 300µCl/gm 1-131 being equivalent to loss of the RCS barrier. Lastly, the references for the FPB bases were updated.
These activities were evaluated using EPDP-17, Attachment 2 to determine whether a 10 CFR 50.54(q) Reduction in Effectiveness Evaluation was required. The activities were determined to be a change to the REP that was not editorial or typographical. Upon completion of
U.S. Nuclear Regulatory Commission Page 3 October 29, 2018, it was determined that the activity could not be implemented without performing a 10 CFR 50.54(q) Reduction in Effectiveness Evaluation; therefore, EPDP-17, Attachment 4 (Effectiveness Evaluation Form) was completed. It was determined that none of the proposed changes were a reduction in effectiveness because they do not impact the timeliness with which an event can be declared. The proposed changes maintain the effectiveness of the REP by aligning with corrected design output. There is no change in assignment of responsibility; and therefore, no adverse affect on how the emergency classification/declaration function is performed by plant operators. The activity continues to comply with the requirements of 1 O CFR 50.47(b) (Emergency plans) and 10 CFR 50 Appendix E (Emergency Planning and Preparedness for Production and Utilization Facilities), and the activity does not constitute a reduction in effectiveness. Therefore, it was determined that the activities could be implemented without prior approval.
The following editorial/typographical changes were incorporated. The Emergency Classification Wall Boards were replaced with updated versions to reflect the changes made. The fuel clad and RCS barrier 2.A [Reactor Pressure Vessel (RPV) Water Level] thresholds were reformatted to better align with conditional logic. Lastly, "SG8" was replaced with "SS8" in the "References" section of IC SS8 in Attachment 3 (Emergency Classification Scheme Bases). These activities were all evaluated under EPDP-17, Attachment 2 to determine whether a 50.54(q) Reduction in Effectiveness Evaluation was required. These changes were determined to be changes to the REP; however, they were determined to be editorial/typographical in nature. Therefore, it was concluded that the activities could be implemented without performing a 50.54(q) Reduction in Effectiveness Evaluation.
There are no new regulatory commitments in this letter. If you have any questions regarding this submittal, please contact B. F. Tidwell at (256)729-3666.
D. L. Hughes Site Vice President cc:
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant NRR Project Manager - Browns Ferry Nuclear Plant NRC Director - Division of Spent Fuel Management, NMSS