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RIA Public Workshop #4Implementation of Revised Acceptance

Criteria

September 25, 2008

Paul M. Clifford

Division of Safety Systems

Nuclear Reactor Regulation

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Opening Remarks

• Interim criteria and guidance provided in updated Standard Review Plan NUREG-0800 in March 2007.

Being implemented in ESBWR and EPR DCDs

• Interim criteria needs to be finalized and implemented across the fleet.

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Opening Remarks

• RIA public workshop #4 provides the opportunity for members of the public and industry to express concerns and provide input to the staff.

1. Propose revisions to interim criteria and guidancea) Status of NEI/EPRI topical report

2. Provide input on implementation schedulea) Enhancements of analytical methods and assumptions

– Future NRC submittalsb) Cost/Benefit Assessment

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RIA Radiological Consequences– Fuel Cladding Failure Criteria

– Fission-Product Inventory Guidance

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Fuel Cladding Failure Criteria

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Previous Failure Criteria

• NUREG-0800 Standard Review Plan Section 4.2 defines RIA fuel clad failure criteria.

– Radial average fuel enthalpy greater than 170 cal/g for BWRs at zero or low power,

– Local heat flux exceeding fuel thermal design limits (e.g. DNBR and CPR) for all PWR events and at-power events in BWRs.

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Issue with Criteria

• Current criteria is not always adequate to protect fuel rod integrity.– Does not recognize PCMI failure mechanism.– Does not address high burnup fuel.– Does not address cladding corrosion.

0

50

100

150

200

250

0 10 20 30 40 50 60 70

Burnup (GWd/t)

Pea

k F

uel

En

thal

py

(cal

/g)

BIGR CABRI IGR NSRR PBF SPERT

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Interim Cladding Failure Criteria

• The high cladding temperature failure criteria for zero power conditions is a peak radial average fuel enthalpy greater than 170 cal/g for fuel rods with an internal rod pressure at or below system pressure and 150 cal/g for fuel rods with an internal rod pressure exceeding system pressure. For intermediate and full power conditions, fuel cladding failure is presumed if local heat flux exceeds thermal design limits (e.g. DNBR and CPR).

• The PCMI failure criteria is a change in radial average fuel enthalpy greater than the corrosion-dependent limits.

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PWR PCMI Failure Criteria

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BWR PCMI Failure Criteria

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Radiological Guidance

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Previous Criteria and Guidance

• Offsite dose consequences limited to “well within”the guidelines in 10CFR Part 100.

• Appendix B to RG 1.77 (1974) provides guidance and assumptions for evaluating the radiological consequences of a control rod ejection accident. These assumptions are supplemented by guidance given in RG 1.183 (2000) and RG 1.195 (2003). – The amount of activity accumulated in the fuel-clad gap

should be assumed to be 10% of the iodines and 10% of the noble gases accumulated at the end of core life, assuming continuous maximum full power operation.

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Reported Transient FGR

0

5

10

15

20

25

30

35

40

0 50 100 150 200

Peak Enthalpy Increase, cal/g

Me

as

ure

d F

GR

, %

Cabri UO2 Cabri MOX NSRR UO2

BWR Transient BIGR VVER

Red: > 55 GWd/MTUBlue: 45 - 55 GWd/MTUGreen: 28 - 45 GWd/MTU

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Fission Product Inventory

The total fission-product gap fraction available for release upon cladding failure equals the steady-state gap inventory (present prior to the event) plus any fission gas released during the event.

– Gap inventory provided in Reg. Guides.

– Transient FGR = [(0.23 ΔH) – 7.14]

0

5

10

15

20

25

30

35

40

0 50 100 150 200

Peak Enthalpy Increase, cal/g

Mea

sure

d F

GR

, %

Cabri UO2 Cabri MOX NSRR UO2

BWR Transient BIGR VVER

Red: > 55 GWd/MTUBlue: 45 - 55 GWd/MTUGreen: 28 - 45 GWd/MTU

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Long-Term Core Cooling

and Reactor Pressure Vessel Integrity

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Regulations

10CFR50 Appendix A, General Design Criterion 28 (GDC28) requires the reactivity control system to be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither:

– Result in damage to the reactor coolant pressure boundary greater than limited local yielding, nor

– sufficiently impair core cooling capability.

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Current Criteria

RG 1.77 identifies acceptable PWR analytical methods and assumptions as well as the following acceptance criteria:

– Fuel radial average energy density limited to 280 cal/g at any axial node.

– Maximum reactor pressure limited to the value that will cause stresses to exceed Service Level C as defined in the ASME Boiler and Pressure Vessel code.

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Issue with Current Criteria

• In 1980, MacDonald et al. reviewed test data from SPERT, TREAT, and PBF and concluded:– SPERT and TREAT results were misinterpreted. A more

appropriate coolability criteria would be 230 cal/g.

– LWR fuel rods subjected to the regulatory limit, radial average fuel enthalpy of 280 cal/g, will be severely damaged and post-accident cooling may be impaired.

• Fuel fragmentation and dispersal not addressed.

• Fuel rod ballooning not addressed.

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Empirical Database

• Loss of fuel rod geometry and molten fuel-coolant interaction reported in some SPERT and PBF tests.

• Fuel fragmentation and dispersal reported in other RIA test programs.

• Pressure pulses reported in some RIA tests.

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Interim Criteria

1. Peak radial average fuel enthalpy must remain below 230 cal/g.

2. Peak fuel temperature must remain below incipient fuel melting conditions.

3. Mechanical energy generated as a result of (1) non-molten fuel-to-coolant interaction and (2) fuel rod burst must be addressed with respect to reactor pressure boundary, reactor internals, and fuel assembly structural integrity.

4. No loss of coolable geometry due to (1) fuel pellet and cladding fragmentation and dispersal and (2) fuel rod ballooning.

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Fuel Melting Calculations

100

150

200

250

300

0 10 20 30 40 50 60 70

Peak Pellet Burnup (GWd/MtU)

Ra

dia

lly-A

ve

rag

ed

Fu

el E

nth

alp

y (

ca

l/g)

RG 1.77

MacDonald Limit

FRAPTRAN - 100ms

FRAPTRAN - 10ms

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Interim Criteria

1. Peak radial average fuel enthalpy must remain below 230 cal/g.

2. Peak fuel temperature must remain below incipient fuel melting conditions.

3. Mechanical energy generated as a result of (1) non-molten fuel-to-coolant interaction and (2) fuel rod burst must be addressed with respect to reactor pressure boundary, reactor internals, and fuel assembly structural integrity.

4. No loss of coolable geometry due to (1) fuel pellet and cladding fragmentation and dispersal and (2) fuel rod ballooning.

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Fuel Dispersal

• Empirical database related to fuel dispersal and fuel-coolant interaction limited.

• Technical challenges:– Flow channel blockage by fragmented fuel

and cladding particles and fuel rod ballooning.

– FCI mechanical energy from the dispersal of non-molten fuel fragments.

– Dispersal of fragmented fuel particles needs to be assessed.

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Implementation Strategy

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Implementation

1. Interim criteria and guidance included in revised Standard Review Plan (NUREG-0800).a) Issued to support the initial licensing of new reactor

fleet.b) Applicable to all advanced reactor design

certifications (DCD) and combined operating license (COL) applications.

2. Encourage licensees and vendors to develop and submit new 3D core neutronic methods and strategy to disposition long-term coolability concerns.

3. Complete further evaluation and adjust, if necessary, criteria and guidance.

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Implementation

4. Finalize criteria and revise Regulatory Guides and SRP.

5. Backfit determination per 10 CFR 50.109.a) If the proposed change in regulatory staff position

qualifies as either an exception (e.g., compliance, adequate protection) or cost-justified substantial increase in safety under the provisions of 10 CFR 50.109, an implementation schedule for applying the final criteria and guidance to both the current reactor fleet and new reactor fleet will be proposed.

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Proposed Schedule

1) Finalize RIA criteria - December 2009

2) Revise guidance documents - June 2010

3) Backfit determination - June 2010

4) NRC internal review / ACRS - TBD

5) FRN / Public Comment - TBD

6) Issue RIS and revised guidance documents - TBD

7) Vendor Analytical Methodology - TBD1) Corrosion / Hydrogen Model

8) Each licensee submits new RIA analysis - TBD

d t ti li if li bl