Other Operational

ModesLecture 6-3

1

Key Topics

• Operational experience

• Early studies

• LPSD conditions and effects on modeling

• Current situation

2

Overview

Resources

• U.S. Nuclear Regulatory Commission, “Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States,” NUREG-1449, September 1993.

• F.E. Haskin, A.L. Camp, S.A. Hodge, and D.A. Powers, “Perspectives on Reactor Safety,” NUREG/CR-6042, Revision 2, March 2002.

• M. Barriere, et al., “An Analysis of Operational Experience During Low Power and Shutdown and a Plan for Addressing Human Reliability Assessment Issues,” NUREG/CR-6093, June 1994.

• D.W. Whitehead, et al., "Evaluation of Potential Severe Accidents During Low Power and Shutdown Operations at Grand Gulf Unit 1: Analysis of Core Damage Frequency from Internal Events for Plant Operational State 5 During a Refueling Outage," NUREG/CR-6143, 1994.

• T.L. Chu and W.T. Pratt, “Evaluation of Potential Severe Accidents During Low Power and Shutdown Operations at Surry, Unit 1: Summary of Results,” NUREG/CR-6144, Vol. 1, October 1995.

3

Overview

Other References

• Nuclear Energy Agency, “Shutdown and Low Power Safety Assessment” NEA/CSNI/R(93)19, Boulogne-Billancourt, France, November 1993. (Available from: http://www.oecd-nea.org/nsd/docs/indexcsni.html).

• Nuclear Energy Agency, “Proceedings of the Workshop on Precursor Analysis,” NEA/CSNI/R(2003)11, Boulogne-Billancourt, France, 2003, (Available from: http://www.oecd-nea.org/nsd/docs/indexcsni.html).

• Nuclear Energy Agency, “Improving Low Power and Shutdown PSA Methods and Data to Permit Better Risk Comparison and Trade-Off Decision-Making, Vols. 1-3” NEA/CSNI/R(2005)11/VOL1, NEA/CSNI/R(2005)11/VOL2, and NEA/CSNI/R(2005)11/VOL3, Boulogne-Billancourt, France, September 2005, (Available from: http://www.oecd-nea.org/nsd/docs/indexcsni.html).

• Nuclear Energy Agency, “Low Power and Shutdown Operations Risk: Development of Structure for Information Base and Assessment of Modelling Issues,” NEA/CSNI/R(2009)17, Boulogne-Billancourt, France, December 2009, (Available from: http://www.oecd-nea.org/nsd/docs/indexcsni.html).

• T.L. Chu, et al., “Evaluation of Potential Severe Accidents During Low Power and Shutdown Operations at Surry, Unit 1: Analysis of Core Damage Frequency from Internal Events During Mid-Loop Operations,” NUREG/CR-6144, Vol. 2, June 1994.

• Z. Musicki, et al., "Evaluation of Potential Severe Accidents During Low Power and Shutdown Operations at Surry, Unit 1: Analysis of Core Damage Frequency from Internal Fires During Mid-Loop Operations," NUREG/CR-6144, Vol. 3, July 1994.

4

Overview

Other References

• Nuclear Energy Agency, “Use and Development of Probabilistic Safety Assessment: An Overview of the Situation at the End of 2010”, NEA/CSNI/R (2012)11, Boulogne-Billancourt, France, 2012. (Available from: http://www.oecd-nea.org/nsd/docs/indexcsni.html)

• International Atomic Energy Agency, “The Fukushima Daiichi Accident: Report by the IAEA Director General,” STI/PUB 1710, Vienna, Austria, 2015.

• U.S. Nuclear Regulatory Commission, “Loss of Vital AC Power and the Residual Heat Removal System During Mid-Loop Operations at Vogtle Unit 1 on March 20, 1990,” NUREG-1410, June 1990.

• Department of Energy, Electric Power Research Institute, Environmental Protection Agency, Federal Emergency Management Agency, Institute of Nuclear Power Operations, and the U.S. Nuclear Regulatory Commission, “ Report on the Accident at the Chernobyl Nuclear Power Station,” NUREG-1250, January 1987.

5

Overview

Who Am I?

• During a maintenance outage, my operators started to perform a safety test (which had been successfully done on some of my sister plants) to verify that during a loss of offsite power (LOOP), my main generator (coasting down) would provide sufficient power until my EDGs kicked in.

• Early in the test, my operators disabled my Emergency Core Cooling System (ECCS) to prevent it interfering with the test. At this point, outside dispatchers requested that the test be delayed so I could provide power to the grid for several hours. Late in the evening the test resumed.

• Complications following the delay led to xenon buildup, which caused difficulties in reaching needed test conditions and put me in an unstable operating regime. My operators decided to proceed with the test.

• In the early morning, my operators partially disabled my automatic scram system (to enable a rapid test repeat, should the first attempt fail) and started the test, ignoring a computer printout indicating that I should be shut down immediately. My power started to rise rapidly.

• 36 seconds after the start of the test, the shift manager ordered an emergency scram, but it was too late. My core disassembled a few seconds later.

6

Operational Experience

Who Am I?

• During an outage for routine inspections, I was in cold shutdown and

undergoing a pressure leak test. My safety relief valves were disabled

to prevent inadvertent opening during the test.

• In the early afternoon, an earthquake tripped breakers at a substation

and I lost all offsite power. A little while later, a flood failed my

emergency service water pumps, my emergency diesel generators

tripped (no cooling), and I entered station blackout (SBO).

• With considerable effort under trying circumstances, my operators

were able to restore power (laying cables to an air-cooled diesel at a

neighboring unit that had escaped flooding damage) in 16.5 hours

and establish cooling (using a low-pressure non-safety system) in 2.5

days.

• I reached a safe and stable condition 9 days after the earthquake.

7

Operational Experience

Who Am I?

• During a scheduled refueling outage, I had been refueled and was

in mid-loop condition,* waiting for final tensioning of my reactor

vessel head studs. One of my two emergency diesel generators

(EDGs) was undergoing maintenance.

• In mid-morning, a service truck backed into a power pole and

caused a loss of offsite power. My remaining EDG started but

tripped and I had no power to my safety buses.

• Lacking power, my operating residual heat removal (RHR) pump

tripped and my water temperature started to rise.

• After unsuccessful attempts to find the cause of my EDG trip, my

operators manually restarted my EDG 36 minutes into the incident

and RHR cooling was restored.

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*Mid-loop: reactor coolant system water level is lower than the top of the

junction of the hot leg with the reactor vessel.

Operational Experience

Risk During Low Power and Shutdown

Operations

• Originally discounted because low decay heat => long

times available for recovery

• Concerns raised about possibility of a slug of unborated

water causing a reactivity accident

• More careful consideration

– Operational experience with loss of decay heat removal, loss of

inventory, reactivity additions

– Early PRAs indicating potential for high (conditional) CDFs

9

Historical Analyses

A Combinatorial Problem

• Multiple combinations of outage types,

and plant operational states (POS)

• Every outage is different (‘there is no

such thing as an “average outage”’)

=> Potentially large number LPSD PRA

models to develop/analyze

10

Conditions and Modeling

Outage Type

Non-Drained Maintenance w/o RHR

Non-Drained Maintenance w/RHR

Drained Maintenance w/RHR

Refueling

POS Mode* Description

0 1 Full power

1 1,2 Low power, rx shutdown

2 3 Cooldown w/SGs to 177C (350F)

3 4 Cooldown w/RHR to 94C (200F)

4 5 Cooldown to ambient

5 5,6 Draining to midloop

6 5,6 Midloop

7 6 Filling refueling cavity

8 6 Refueling (old core)

Defueled

8 6 Refueling (new core)

9 6 Draining to midloop

10 5,6 Midloop

11 5,6 Refilling RCS

12 5 RCS heatup

13 4 RCS heatup to 177C (350F)

14 3 Startup w/SGs to Hot Stdby

15 1,2 Rx startup

*PWR modes per NUREG-1431

Thought Exercise

Outages can involve periods of:

• Low reactor coolant system inventory

• Equipment testing and/or maintenance

• Open containment

How might these conditions lead to differences with an at-power PRA

model?

• Hazards

• Initiators

• Event Trees

• Fault Trees

• Basic Events

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Conditions and Modeling

Early U.S. LPSD PRAs

Date Study CDF (/ry) Notes

1985Zion

(NSAC-84)1.8E-5

Extension of Zion PRA. Reduced inventory scenarios = 61% total CDF.

LORHR = 56%, LOCA = 6%. Operator errors dominant.

1988NUREG/

CR-50155.2E-5

Modified NSAC-84. LORHR = 82%, LOOP = 10%, LOCA = 8%. Operator

errors dominant.

≥1983Seabrook

(PLG)4.5E-5

Supplement to at-power Level 3 PRA. Analyzed Modes 4 (hot shutdown), 5

(cold shutdown), and 6 (refueling). Included fires and floods. LORHR = 82%,

LOCA = 18% (but dominates health risk – equipment hatch can’t be closed in

time; overpressure => stuck open relief valves or ruptured RHR pump seals).

Reduced inventory = 71%.

1985Brunswick

(NSAC-83)7.0E-8

Loss of RHR only. Dominant contributors: RHR/RHRSW maintenance;

RHR/RHRSW pump failures; RHRHX CCF

Sequoyah

(SAIC)

8-5E-7 to

7.5E-5

LOCA during Mode 5 only. Initiators: safe shutdown earthquake, operator

error. Dominant contributors: operator-induced LOCA, availability of power,

maintenance, operator errors during response, RHR pump air binding, RHR

suction failure

12

Historical Analyses

Adapted from: U.S. Nuclear Regulatory Commission, “Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States,” NUREG-1449,

September 1993

Early International LPSD PRAs

StudyCDF

(% Total)Dominant Notes

Doel 3

Tihange 2

Full PSA including LPSD; Level 1 + containment

response; internal events

TVO-SEPRA >50%1. Leakage below core

2. Loss of RHR

LPSD; Level 1; internal events; improvements reduce

CDF ~10%

EPS 900 32% LOCA + loss of SIFull PSA including LPSD; Level 1; internal events;

human error dominant

EPS 1300 70%

GRS-BWR 72Phase 1: LPSD screening; Level 1; internal events

Phase 2: Full PSA including LPSD, fire, flood; Level 2

GRS-PWR 13001. Deboration

2. Loss of RHR at midloopLPSD screening; Level 1; internal events

Vattenfall –

Ringhals 2Loss of RHR at midloop

LPSD; Level 1; internal events; pessimistic

assumptions

Sizewell B >60% Fires Full PSA including LPSD, fire, flood; Level 3

Borssele LPSD (planned)

13

Historical Analyses

Adapted from: Nuclear Energy Agency, “Shutdown and Low Power Safety Assessment” NEA/CSNI/R(93)19, Boulogne-Billancourt, France, November 1993.

NUREG-1150 follow-on studies

• Grand Gulf (BWR) – NUREG/CR-6143

– Internal events

– Mode 5 (cold shutdown)

• Surry (PWR) – NUREG/CR-6144

– Full scope (internal events, fire, flood, seismic)

– Screening Level 1 analysis for all plant operating states, four

outage types (refueling, drained maintenance, non-drained

maintenance with RHR, non-drained maintenance without

RHR)

– Detailed analysis for mid-loop (Level 3)

14

Historical Analyses

Example Results – Surry Level 1

CDF (/ry) Mid-Loop At

PowerMean 5th 95th

Internal Events 5E-6 5E-7 2E-5 4E-5

Internal Fires 2E-5 1E-6 8E-5 1E-5

Internal Floods 5E-6 2E-7 2E-5

Seismic 9E-8 3E-10 4E-7 3E-5

Total 3E-5 8E-5

15

Notes:

1) Sources: NUREG-1150 and NUREG/CR-6144

2) LPSD results include fraction of time in mid-loop (0.066)

3) Seismic results are based on EPRI hazard curves

Screening:

Detailed:T.L. Chu and W.T. Pratt, “Evaluation of Potential Severe

Accidents During Low Power and Shutdown Operations

at Surry, Unit 1: Summary of Results,” NUREG/CR-6144,

Vol. 1, October 1995

Historical Analyses

Current Situation

• U.S.

– Not required for risk-informed applications

– Included in ongoing Level 3 PRA project

– “Trial Use” PRA standard

• International

– Full scope (internal events, internal hazards,

external hazards) required in many countries

16

Current Situation