Group2 loca to test

LOCA Safety Analysis

ReviewVARANS - Group 2

Truong Cong Thang (Group leader)

Doan Quang Tuyen

Nguyen Dinh Quen

Do Trung Quan

Nguyen Hoang Anh

Dao Ngoc Phuong

5.1 Scope for Analysis

The initial conditions for the analysis of a postulated event shall be

specified to give the severest possible result with respect to the applied

criteria, taking into consideration the whole range of normal operation

and operating period of the nuclear reactor facility including long-term

physical change with burn-up in the core during various cycles and with

refueling and anticipated change in operational modes. The analysis

shall in general cover the time range up to the point where the event

terminates and it can be reasonably inferred that the reactor could

reach a cold shutdown state safely.

Review guide for safety evaluation

5.2 Assumptions on Safety Functions

(1) Of safety functions designed to address postulated events, those which are allowed to be taken

into account in the analysis, shall in general be limited to safety functions to be performed by

structures, systems and components belonging to MS-1 and MS-2 specified in the Review

Guide for Safety Importance Classification. Safety functions of structures, systems and

components belonging to MS-3 may be taken into account in the analysis only if taking

credit for these functions is proved to be justifiable.

(2) The analysis shall, in addition to a postulated event for the systems and components necessary to

deal with an accident, assume a single failure of a component that could give the severest

possible consequence for each of the fundamental safety functions such as reactor

shutdown, core cooling and radioactivity confinement. For a short-term period after

occurrence of an accident, a single failure on an active component shall be assumed, while

for a long-term period, a single failure on an active component or a passive component

shall be assumed. The failure may not generally be assumed on a component which is

operated before the occurrence of the event and will be operated after that. The failure of a

passive component may not be assumed, if a single failure is assumed and when the

system which includes the said component is designed to fulfill its required safety

functions, or when the failure can be removed or repaired within time so as not to impair

the safety, or when the probability of the failure is sufficiently low.


(3) The analysis shall take into account an appropriate margin of time for manual

operations by operators to cope with the postulated event.

(4) If functions of the safety protection system are expected in the analysis, the kinds

of signals to actuate it and the timing that the signals are issued shall be defined. The

same requirement shall also be applied to other systems if their expected

performance affects the result of the analysis.

(5) The analysis of an accident shall take into account unavailability of off-site power if

functions of the engineered safety features are expected.

(6) If in the analysis the effect of reactor scram is expected, the kinds of signals to

initiate the scram shall be defined, and appropriate delay times for effective scram

initiation shall be considered. In addition the shutdown effect shall be evaluated on the

assumption that a control rod (or a group of control rods connected to a common

drive mechanism) with the maximum reactivity worth in the postulated

conditions is held at the fully withdrawn position.


Categorization of eventSAR Appendix 10: 1.1.2.1 Definition:

“ Accidents" are abnormal states going beyond the abnormal operational transients described in Section 1.1.1, "Abnormal Operational Transients." Although they occur with a very low frequency , radioactive substances might possibly be released from the reactor installation should they occur. Therefore, they are events which must be postulated from the view point of evaluating the safety of the reactor installation.

Categorization of event

Accident includes following events:

(1) Loss of reactor coolant or significant change of reactor core cooling

a. Loss of reactor coolant (hereinafter called "LOCA")

b. Loss of reactor coolant flow

c. Seizure of reactor coolant pump

Acceptance CriteriaReview guide: 4.2

Acceptance criteria for accidents

(1) The core shall not be considerably damaged, and can be sufficiently cooled.

(2) Fuel enthalpy shall not exceed the specified limit.

(3) Pressure on the reactor coolant pressure boundary shall not exceed 120% of the maximum allowable working pressure.

(4) Pressure on the reactor containment boundary shall not exceed the maximum allowable working pressure.

(5) There is no significant risk of radiation exposure to the surrounding public.

Why has a acceptance criterion of MCPR for AOO, but not for

accident?

3.2.1.4 Review of conformance to acceptance criteria

a. A maximum calculation of a fuel clad temperature must be 12000C or less

b. A calculation of a stoichiometric oxidization quantity of fuel cladding must be 15% or less the thickness of the cladding before significant oxidation

c. The quantity of hydrogen that is generated in the core with the reaction between the fuel clad and structural members and water must be low enough to ensure the integrity of a containment vessel should be clearly identified

d. The design must enable long-term removal of decay heat even if variations in fuel geometry are consideredFuel enthalpy should be covered.

Number of damaged fuel rods should be given.

Acceptance Criteria

Computer codes

(2) Analytical methods

• The following three analytical codes described above in Section 1

,3 , "Calculation Codes Used in Analysis ," are to be used in the

analysis

a. LAMB: Analytical code for short-term thermohydraulic

transients

b , SCAT: Single-channel thermohydraulic analytical code

c. SAFER: Analytical code for long-term thermohydraulic

transients

Availability of safety systems

5.2 Assumptions on Safety Functions

Of safety functions designed to address postulated events, those which are allowed to be taken

into account in the analysis, shall in general be limited to safety functions to be performed by

structures, systems and components belonging to MS-1 and MS-2 specified in the Review

Guide for Safety Importance Classification. Safety functions of structures, systems and

components belonging to MS-3 may be taken into account in the analysis only if taking

credit for these functions is proved to be justifiable.

-> This requirement is compliedTable 1.2-1 Main safety functions considered for effect mitigating in analysis (1) and (2)

Classification Function Structure, System and/or

Components

MS-1 (ab.norm. transient and

accident)

Rapid shut down of reactor Control rods and Control rod driving

systems(Scram function)

Maintaining sub-criticality of core Control rods and Control rod driving

systems

Question: For “Maintaining sub-criticality of core” in the case of transient, Boron dilution could be used,

not only control rod.

Classificati

on

Function Structure, System and/or Components

MS-1

(ab.norm.

transient)

Actuating signal generation for

Engineering safety facilities and

reactor shutdown systems

Safety protection systems (list of safety protection systems

are needed !!)

MS-1

(accident)

Over pressure prevention of

reactor pressure boundary

Safety relief valves (Safety valve open function),

Heat removal after reactor shut

down

Residual Heat Removal System, Reactor Core Isolation

Cooling System, Safety relief valves (Relief valve manual

function),

Core cooling Low Pressure Core Flooder, High Pressure Core Flooder,

Reactor Core Isolation Cooling System, Automatic

Depressurization System

Question: Function Core cooling is also needed for ab.norm. transient !!

List of safety protection systems are needed to be specified !!

Table 1.2-1 Main safety functions considered for effect mitigating in analysis (1) and (2)



Components

MS-3 (ab.norm. transient) Mitigation of reactor pressure

increase

Safety relief valves (Relief valve

function), Turbine bypass valves

Restriction of reactor power

increase

Reactor coolant recirculation

system (Recirculation pump trip

function), Control rod withdrawal

monitoring system

Power maintaining of Reactor

coolant recirculation pumps

Recirculation pump MG sets

Question: How the Reactor coolant recirculation system can restrict

the increase of reactor power (depend on moderator coeff., Doppler

coef., etc.)?

Table 1.2-1 Main safety functions considered for effect mitigating in analysis (1)

and (2)



Components

MS-3 (accident) Confirmation of abnormal

situation

Monitors of Heating,

Ventilating and Air

Conditioning systems

Question: Need the monitoring of pressure, flow??

Table 1.2-1 Main safety functions considered for effect mitigating in analysis (1)

and (2)


3.2.1 Loss-of-coolant accidentsSAR KK6&7 (Appendix 10)

3.2.1.1 Causes

• Suppose the nearly inconceivable circumstance in which one of

the various pipes connected to the pressure vessel should break

during reactor operation for some reason. Should this occur, the

coolant will leak out of the pressure vessel or will be lost. In this

case, if the coolant cannot be replenished, it will become

impossible to cool the core sufficiently , and in the worst case

the fuel temperature will rise excessively on account of the decay

heat, and fission products may possibly be released from the fuel Satisfied

Boundary and Initial Conditions

SAR KK6&7 (Appendix 10)

3.2.1.3 Analysis of Accident Process

(1) Analysis conditions

The analysis of the HPCF pipe ends rupture accident shall be

conducted based on the following assumptions.

- The break size should be clearly given in values

- Where is the position of break?

- Why was the HPCF pipe break selected for severest case?

Table 3.2, 1-1 Main calculating conditions for loss-of-coolant accidents

(SAR)Item Values used

Reactor thermal powerApproximately 102% of rated power (4,005 MWt)

Maximum linear heat generation rate 44.0 kW/m x 1.02

Core flow Rated flow 52.2 x 103 t/h

Reactor dome pressure 73.1 kg/cm2 g

Core inlet enthalpy 294 kcal/kg

Flow rate, High-Pressure Core-Flooder System (rated value) 727 m3/h (per pump, at 7.0 kg/cm2 d)*

Flow rate, Low-Pressure Core-Flooder System (rated value) 954 m3/h (per pump, at 2.8 kg/cm2 d)*

Reactor Core Isolation Cooling 182 m3/h (per pump, at 82.8 - 10.5

System (rated value) kg/cm2 d)*

What are some others rated values?


Boundary and Initial ConditionsSAR KK6&7 (Appendix 10)

a. The reactor shall operate at about 102% of rated power (4 ,005 MWt) and at a rated core flow rate immediately before the accident. An initial value of reactor pressure shall be 73.1 kg/cm2g. An initial value of MCPR will not actually become smaller than the operation limit (1.22) , but shall be 1.19 , or a value commonly used for an analysis of a loss-of-coolant accident in a boiling water reactor (BWR).

- Reactor power is conservative

- It should shown that the value used of initial pressure

is conservative by comparing with rated value.

Boundary and Initial ConditionsSAR KK6&7 (Appendix 10).

a. The maximum power density of a fuel rod that is used for the analysis

shall be 102% of 44.0 kW, or the operation limit. For a gap heat

transfer coefficient between a fuel clad and pellet, a value that will

make the analysis result more stringent shall be used in consideration of

variations in the heat transfer during a period of fuel burning

The heat transfer through fuel gap has not clearly identified?


SAR KK6&7 (Appendix 10).

c. For the decay heat after the shutdown of the reactor, a value

determined from an equation (GE (mean) + 3σ) that incorporates a

safety margin into actual measurements, shall be used. For

reference, this equation incorporates a decay heat of actinide.

Satisfied



d. Off-site power shall be lost concurrently with the occurrence of the accident Consequently, a recirculation pump will instantly be tripped. The reactor shall scram at a signal of a sharp drop in the core flow rate. Fig. 3.2.1-1 shows set values for the scrams at the sharp drop in the core flow rate

The description of SCRAM process was

mentioned

The time delay of scram was not specified in

value.

Review guide 5.2:

(6) If in the analysis the effect of reactor scram is expected, the kinds

of signals to initiate the scram shall be defined, and appropriate delay

times for effective scram initiation shall be considered.


e. It shall be considered that a signal for high pressure of a drywell

as a ECCS startup signal will be given earlier than a signal for a

low water level in the reactor (Level 2 or 1) , but ECCS is assumed

to conservatively start up at the signal for the low level

Conservative assumption of ECCS startup signal was used

Boundary and Initial ConditionsSafety guide: 5.2 Assumptions on Safety Functions (2)

The analysis shall, in addition to a postulated event for the systems and components necessary to deal with an accident, assume a single failure of a component that could give the severest possible consequence for each of the fundamental safety functions such as reactor shutdown, core cooling and radioactivity confinement. For a short-term period after occurrence of an accident, a single failure on an active component shall be assumed, while for a long-term period, a single failure on an active component or a passive component shall be assumed. The failure may not generally be assumed on a component which is operated before the occurrence of the event and will be operated after that. The failure of a passive component may not be assumed, if a single failure is assumed and when the system which includes the said component is designed to fulfill its required safety functions, or when the failure can be removed or repaired within time so as not to impair the safety, or when the probability of the failure is sufficiently low.



g. The most stringent single failure shall be assumed in the ECCS

network from the viewpoint of the capability of reactor cooling. The

most stringent single failure in the case of the HPCF pipe rupture

accident shall be a failure of a diesel generator that supplies power

to an integrity high-pressure core injection system

A diesel generator was assumed to be fail to

conform with single failure criterion



h. The leakage of coolant from the fractured area shall be

calculated based on a uniform critical flow model

i. In a safety and relief valve, the relief valve works earlier than the

safety valve, but the safety valve shall be assumed to work earlier

Please explain the effects to safety when the safety valve is

assumed to work earlier than relief valve? (pressure boundary

was fail)



j. In the calculation of a clad temperature , the following correlation equations are used to determine a heat transfer coefficient between the clad and the coolant

(1) Cooling of nucleate boiling: Correlation equation as a function of coolant void fraction

(2) Cooling of film boiling: Correlation equation that uses a correlation equation of spray flow cooling and the corrected Bromley Equation as a function of the void fraction

(3) Cooling of transition boiling: Correlation equation obtained after a heat transfer coefficient between nucleate and film boiling is interpolated with a degree of overheating of fuel cladding

(4) Steam cooling: Dittus-BoeIter Equation

(5) Spray flow cooling: Sun-Saha Equation

(6) Spray (falling water) cooling: Correlation equation based on actual measurements of SHTF test

(7) Wet cooling: A heat transfer coefficient after being wetted is based on the model of Andersen.


There is no information about reactivity feedback

and fuel burn-up?

Evaluation of Results

Variations of core flow rate


The loss of offsite power

occurring simultaneously with

the accident , the core flow will

decrease rapidly because of

the shutdown of the

recirculation pumps.

Variations of reactor water level

SAR Appendix 10

The water level inside the core shroud

will start to drop after about 60

seconds. However , the Reactor Core

Isolation Cooling System will be

activated by low level (level 2) signals

of reactor water and start water

injection in about 59 seconds after the

accident. Why does the water level inside

the core keep high for 60s after

break occur?

Generally it should be dropped

rapidly.

Variations of core average pressureSAR KK6&7 (Appendix 10)

The Automatic Depressurization

Systems will also be activated by high

pressure signals of drywell and low

level (level 1) signals of reactor water

in about 176 seconds after the accident

to lower the reactor pressure , and two

Low-Pressure

Flooder Systems will begin to inject

water in about 364 seconds. The water

level inside the core shroud will not

drop below top of the active fuel and

the core will be kept f1 flooded.

Why does the reactor core can keep high pressure for

longtime (176s) after LOCA occurs?

Heat transfer coefficient change

Fuel cladding temperature change

- Figure 3.2.1-8 was not

provided!

- Figure 3.2.1-7 shows the

cladding temperature. The pick

cladding temperature is shown

as 6000C, not 5550C as

described in SAR!

SAR Appendix 10

Fig. 3.2.1-8 shows that the fuel

cladding-tube temperature is

about 5550C or less during a

complete break of the HPCF

lines.

Oxidation

SAR Appendix 10

There is very little increase in the thickness of the oxide layer on the

fuel cladding tubes because of the low temperature of the fuel

cladding tubes. Moreover, the zirconium-water reaction fraction in

all of the fuel cladding tubes is negligibly small

It should give a calculation results of the

total amount of hydrogen generated.

Other comments

- It should provide the time sequences (in table) that show

the operation of every system or component.

- The leakage flow depending time was not provided.

- There was not any description of operation action.

- There is not clear about the effect of single failure of

ECCS to the results.

- Cần bổ sung nhận xét về các mô hình phân tích ví dụ:

- Blowdown model

- Drywell model

- Vent-clearing model v.v…….

- (yêu cầu các mô hình này phải được dẫn chiếu đến tài

liệu thẩm định mô hình liên quan)

Câu hỏi liên quan tới Fukushima: có 3 lò bị nóng chảy vùng hoạt, tuy nhiên hydro lại từ 3 lò này qua hệ

thống thông gió sang lò thứ tư và dẫn tới nổ ở lò thứ 4, trong ABWR có thể xảy ra tình huống này ko?

Cách ngăn chặn ?

• Trong kết quả có nói về tăng giảm nhiệt độ vỏ thanh nhiên liệu mà

không có hình vẽ cần có hình vẽ thể hiện nhiệt độ vỏ thanh nhiên

liệu để làm rõ được sự thay đổi bề mặt vỏ thanh ?

Theo US-NRC -The calculated total amount of hydrogen generated from the chemical reaction of

the cladding with water or steam does not exceed 1% of the hypothetical amount that would be

generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding

surrounding the plenum volume, were to react. SAR: A rate of a zirconium-water reaction shall be 5 times larger than the result of loss-of-coolant analysis,

or shall be a rate obtained when the reaction is made in fuel dads with a thickness of 0.23 mil (equivalent

to 0.73% of the total quantity of fuel clads), but a higher result shall be used. In the analysis, the rate shall

be 0.73%.

-> conservative ?

3.5.1 Loss of Reactor coolant

G33. Systems for Controlling Containment Facility Atmosphere

(1) The containment facility atmosphere cleanup system shall be designedto be capable of reducing the concentration of radioactive materialsreleased to the environment in case of postulated events for reactorcontainment design.

(2) The flammable gas concentration control system shall be designed tobe capable of controlling the concentration of hydrogen or oxygen presentin the reactor containment in case of the postulated events for reactorcontainment design, thereby maintaining the integrity of the containmentfacility.

(3) The systems for controlling containment facility atmosphere shall bedesigned with redundancy or diversity and independence so that they canfulfill their safety functions even in case of unavailability of off-site power inaddition to an assumption of a single failure of any of the components thatcomprise the systems. They shall also be designed to allow testing withrespect to their functional capability.

Safety review guide

GUIDE_005

(1) An event is assumed that combustible gas is generated during a loss of the reactor coolant assumed in 3.4.1.

(2) The amount of hydrogen generated by metal-water reaction shall be the larger value of either five times the

amount generated by metal-water reaction that is calculated in 3.1.1 or the amount generated when the metal of

0.0058mm thickness from the surface of the cladding tubes of all fuel rods reacts with water.

(3) Assuming that 50% of halogen and 1% of the fission products excluding noble gas and halogen out of the

fission products inventory in the reactor core exist in the liquid phase of the water in the reactor containment, the

radiolytic decomposition of the water in the reactor containment shall be appropriately evaluated. Furthermore,

assuming that all other fission products excluding noble gas exist in the reactor core, the radiolytic decomposition

of the water in the reactor core shall be appropriately evaluated. The decomposition rate of the water per unit

energy absorbed shall be the value confirmed by experiments with an appropriate margin taken into account.

(4) For a design that adds materials such as alkali in the reactor containment water, the hydrogen generated by

chemical reaction with metal structures in the reactor containment shall be appropriately evaluated.

(5) For a design that provides a system to control the concentration of combustible gases, such as a hydrogen

recombiner, the function may be expected within the design range of these systems.

(6) As criteria, the concentration of either oxygen or hydrogen in the reactor containment atmosphere shall be 5%

or 4% or less, respectively, for at least 30 days after the occurrence of the event.

Safety review guide

Evaluation of Results

Variations of pressure in drywell and suppression

chamber• The figure is not clear,

the values can not be

identified!• The activating time of

ACCS was not shown

clearly.• The point that the

pressures drop suddenly

is the time of initiating

RHRs?• What are the effects

of single failure?

Variations of Temp. in Drywell and suppression

chamberThe assumption describers that:

It is assumed that the Residual Heat

Removal System will be manually switched

to the Containment-Vessel Spray-Cooling

System 10 minutes after the accident

But the Result shows:

The Residual Heat Removal System is

used at first as a Low-Pressure Flooder

System, but 15 minutes after accident it

is switched manually so that one pump will

be used as a Containment-Vessel Spray-

Cooling System to lower the pressure in the

containment vessel Do they conflict?

Results

• The temperature can drop suddenly?

Normally it will decrease slowly.

• The pick of temperatures and

pressures are lower than “limit

values”

But the acceptance criteria for

drywell and suppression chamber were

not listed!

3.5.2 Generation of Flammable Gases

Results

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