ML13238A307
| ML13238A307 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 08/29/2013 |
| From: | Duke Energy Carolinas |
| To: | Gratton C Plant Licensing Branch II |
| Gratton C | |
| References | |
| Download: ML13238A307 (28) | |
Text
CR-3 Decommissioning Submittals August 29, 2013
Duke Participants John Elnitsky - VP Project Management and Construction Terry Hobbs - CR3 Decommissioning Director Blair Wunderly - CR3 Plant Manager Blair Wunderly CR3 Plant Manager Mike Delowery - CR3 Decommissioning Planning Manager Phyllis Dixon - CR3 Decommission Support Manager Matt Widener CR3 Security Manager Matt Widener - CR3 Security Manager Kate Nolan - Associate General Counsel Dan Westcott - CR3 Licensing Supervisor 2
Overview Decommissioning status Permanently defueled (PD) accident analysis Near-term submittals Near term submittals PD Emergency Plan (PDEP) LAR PD Technical Specifications (PDTS) LAR Post-Shutdown Decommissioning Activities Report (PSDAR)
Post Shutdown Decommissioning Activities Report (PSDAR)
PD Security Plan NRC expectations on review durations 3
Decommissioning Status Announcement to decommission under the SAFSTOR alternative made on 2/5/13 10 CFR 50.82(a)(1) certifications submitted on 2/20/13 NRC acknowledgement pursuant to 10 CFR 50.82(a)(2) received on 3/13/13 NRC acknowledgement pursuant to 10 CFR 50.82(a)(2) received on 3/13/13 Certified Fuel Handler Training Program submitted on 4/15/13 LAR 313 on Section 5 of the Technical Specifications submitted on 4/25/13 Decommissioning Transition Organization in place 6/3/13 Decommissioning Transition Organization in-place 6/3/13 Initial staffing level about 50% of operating level Staff size expected to be reduced to dormancy levels in 2015 I
ti i ht t iti d t R i
I 8/1/13 Inspection oversight transitioned to Region I on 8/1/13 All fuel assemblies remain in the spent fuel pool (SFP) 177 fuel assemblies were last critical on 9/26/09 SFP and cooling lines are seismically qualified 4
PD Accident Analysis Anticipated Operational Occurrence Loss of SFP Cooling Design Basis Accidents g
Fuel Handling Accident Waste Handling Accident Beyond Design Basis Event Beyond Design Basis Event Loss of SFP Inventory 5
Loss of SFP Cooling Both loops of recirculating cooling are unavailable SFP contains greater than 380,000 gallons of water which provides abundant passive cooling SFP heat up rate is approximately 1°F/hr p
pp y
Time to boil is over 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> Slow heat up rate provides ample time for mitigating actions No radiological consequences expected No radiological consequences expected 6
Dose Receptors Exclusion Area Boundary (EAB) - 4,400 feet Low Population Zone (LPZ) - 5 miles p
(
)
7
PD Accident Analysis Dose Limits Alternative Source Term (AST) amendment was approved 9/17/01 10 CFR 50.67 dose limits apply RG 1.183 guidance implemented
§50.67(b)(2) dose limits Control Room (CR) - 5 rem TEDE EAB - 25 rem TEDE over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> LPZ - 25 rem TEDE EPA PAG Manual 1 rem TEDE offsite Accident analysis dose limits CR - 5 rem TEDE EAB - 1 rem TEDE EAB 1 rem TEDE 8
Fuel Handling Accident Current licensing basis (CLB):
All 208 rods in a fuel assembly are assumed to fail (100% gap release)
No credit taken for Auxiliary Building and Control Room ventilation filters Atmospheric dispersion factors (/Qs) were updated Meteorological data from 2003 - 2007 was analyzed ARCON96 was used to calculate onsite /Qs PAVAN was used to calculate offsite /Qs RADTRAD dose results CR 1.3E-04 rem TEDE EAB 5.9E-05 rem TEDE 9
Waste Handling Accident Waste gas decay tank rupture accident no longer possible Bounding accident is a dropped high integrity container (HIC)
Conservative primary resin source term assumed Conservative primary resin source term assumed Composite of highest radionuclide concentrations in shipments over last 5 years 10% release fraction assumed (factor of 10 larger than 10 CFR 30.72, Schedule C)
Offsite receptors analyzed consistent with CLB and BTP 11 5 Offsite receptors analyzed consistent with CLB and BTP 11-5 Dose results EAB 0.04 rem TEDE 10
Loss of SFP Inventory Thermal hydraulic and dose consequence analyses conducted Thermal hydraulic scenarios Air cooled analysis y
Peak clad temperature is 547°C (no credit for ventilation)
Cladding remains intact Adiabatic analysis
Time to 900°C approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />
Sufficient time is available to take mitigating actions and, if necessary, offsite protective measures without preplanning 11
Loss of SFP Inventory (contd)
Dose rates calculated using MCNP5, Version 1.60 CR 8 E-06 rem/hr TEDE EAB 2 E-07 rem/hr TEDE Time to reach dose limits much greater than a year 12
PDEP LAR Anticipated submittal: October 2013 Scope of LAR PD EALs Exemptions to §50.47 and Appendix E PDEP 13
PD EALs Current EAL scheme based on NUMARC/NESP-007, Revision 2 PD EAL scheme will be based on NEI 99-01, Revision 6 Recognition Category PD initiating conditions used g
g y g
No deviations taken Classification levels Unusual Event and Alert Unusual Event and Alert
Radiological conditions
Hazards
System malfunctions 14
Proposed Exemptions to 10 CFR 50.47
§50.47(b)
(1) assigned responsibilities (3) requesting assistance (4) classification (5) notification (6) communications (7) public information (9) offsite monitoring (10) protective actions for EPZs (14) exercises
§50.47(c)
(2) define EPZs 15
Proposed Exemptions to 10 CFR 50 Appendix E IV. Content of Emergency Plans A. Organization B. Assessment Actions C. Activation of Emergency Organization D. Notification Procedures E. Emergency Facilities and Equipment F. Training I. Onsite Protective Actions During Hostile Action 16
PDEP Significant changes Offsite organizations removed Shift and emergency staffing capabilities deleted Fire Brigade defined in the Fire Protection Plan Departures from the guidance in NUREG-0654 17
PDTS LAR Anticipated submittal: November 2013 Technical Specification changes based on 10 CFR 50.36 applicability to a PD plant License Condition deletions and revisions proposed License Condition deletions and revisions proposed 18
10 CFR 50.36 Requirements Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity.
The Safety Limits in the ITS relate to the operating reactor core and RCS pressure which are not needed in PDTS.
Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary significant abnormal degradation of the reactor coolant pressure boundary.
The reactor coolant pressure boundary performs no safety function with all fuel stored in the fuel pools.
No Limiting Conditions for Operation (LCOs) included in the ITS to satisfy Criterion 1 are needed in the PDTS.
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10 CFR 50.36 Requirements (contd)
Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
The only fission product barrier that is still performing a safety function is the fuel cladding. No LCOs associated with the reactor coolant or containment fission product barriers that satisfy Criterion 2 are needed in the PDTS.
Criterion 3 A structure system or component that is part of the primary success path and
Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
The fuel handling accident and the radioactive waste handling accident are the only design basis g
g y
g accidents which can occur in a permanently defueled plant. There are no structures, systems, or components which actuate to mitigate either of those accidents 20
10 CFR 50.36 Requirements (contd)
Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
Each limiting condition for operation (LCO) in the ITS was evaluated for applicability to a permanently d f l d l t
d it i
ifi t
bli h lth d
f t If LCO i t
li bl f th defueled plant, and its significance to public health and safety. If an LCO is not applicable for the protection of nuclear fuel stored in the spent fuel pools it was not included in the PDTS.
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Remaining Technical Specifications 3.7.13 Fuel Pool Water Level 3.7.14 Spent Fuel Pool Boron Concentration 3.7.15 Spent Fuel Assembly Storage 3.7.15 Spent Fuel Assembly Storage 22
License Condition Deletions Mitigating Strategies for addressing large fires and explosions 10 CFR 50.54(hh)(3) states that §50.54(hh) does not apply to a nuclear power plant for which certifications under §50.82(a) have been submitted.
NRC letter dated 11/28/11 rescinded Item B.5.b of the Interim Compensatory Measures Order EA 026 Cyber Security 10 CFR 73 54 t t B N b
23 2009 h li tl li d t t
l 10 CFR 73.54 states, By November 23, 2009 each licensee currently licensed to operate a nuclear power plant under part 50 of this chapter shall submit, as specified in §50.4 and §50.90 of this chapter, a cyber security plan that satisfies the requirements of this section for Commission review and approval.
NRC letter dated 3/13/13 acknowledged that pursuant to 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for CR3 no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel.
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Remaining Programs, Manuals, and Reports 5.6.2.3 - Offsite Dose Calculation Manual 5.6.2.17 - Technical Specifications Bases Control Program 5.7.1.1.b - Annual Radiological Environmental Operating Report 5.7.1.1.b Annual Radiological Environmental Operating Report 5.7.1.1.c - Radioactive Effluent Release Report 24
PSDAR Anticipated submittal: December 2013 SAFSTOR approach will be pursued Spent fuel storage strategy under evaluation Spent fuel storage strategy under evaluation Dry storage in ISFSI Wet storage in nuclear island LLW storage footprint will be minimized LLW storage footprint will be minimized Environmental contamination is minimal GEIS bounds expected environmental impacts 25
Defueled Security Plan Anticipated docketing: June 2014 Changes will not decrease effectiveness of the following plans:
Physical Security Plan y
y Guard Training and Qualification Plan Safeguards Contingency Plan Changes will be docketed under 10 CFR 50.54(p)(2) within 60 days Changes will be docketed under 10 CFR 50.54(p)(2) within 60 days A closed meeting will be requested in early 2014 to discuss proposed changes 26
NRC Review Duke Energy is committed to carefully planning the decommissioning transition phase Changes in staffing levels are linked to the approval of the PDEP and PDTS CR3 decommissioning schedule assumes one year reviews Predictable review durations facilitate effective staffing transition Predictable review durations facilitate effective staffing transition 27
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