ML20203K866

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Rev 0 to Training Lesson Plan LO-LP-36001-00-C, Introduction to Mcd:Accident Analysis Methods
ML20203K866
Person / Time
Site: Vogtle  Southern Nuclear icon.png
Issue date: 04/08/1986
From: Brigdon R, Swartzwelder J
GEORGIA POWER CO.
To:
Shared Package
ML20203K798 List:
References
LO-LP-36001-, LO-LP-36001-00, NUDOCS 8608210387
Download: ML20203K866 (23)
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. . - _ - - - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _

GeorgiaPower

, cowan canauron canarusar ikk H Ro - s.P- / S6 VOGTLE ELECTRIC GENERATIN TRAINING LESSON PL s le WLP-15y 3 INTRO TO MCD: " # 077 TITLE. ACCEST ANALYSIS MEMMUMBER-

  • I.IC DSED OPERATOR TRAM M L.0- t P-3sco t-co PROGRMl: REVIS10N: D -"

AUTHOR: R. BRI@0N 3'"'"

DATE:

APPROVED: j f INSTRUCT 0[ GUIDELINES: O

) DATE: f-g_.gM; 1

STUDENT MATERIALS j_ O MO .%00 l -OO- C~ OO l HANDOUT: -9R-He-4H-49&, " INTRODUCTION TO MCD: ACCIDENT ANALYSIS METHODS" TRANSPARENCIES "

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~ - n CHF (Figure -

1)

Lo-TF-34ao 8-oo-C-003 M di4b9ff-002- DATA SPEAD FOR PREDICTED vs ACTUAL CHF (Figure 2) '

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60-TF-36 col-00-C-oos[d M eh-009 CHF vs ACTUAL CORE HEAT FLUI (Figure 3) l

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.%- . . -t--404-g-77-J60s1-00-C.- cos - s 6R-TP-Off-004- FIGURE 2.1-1 TECH SPEC SAFETY LIMIT CURVE (Figure 4) 60-7F-360o 1-60-(- ce6 --* en-9P-eff=005- RCCA BANK REACTIVITY WORTH vs ROD DROP TIME  ;

(Figure 3)

. ~ , , , , _ . . . . ,

to-77-36001 C- oo7-+-St=TP-Off=000 DOPPLER POWER COEFFICIENI USED IN ACCIDENT l ANALTSIS (Figure 5)  !

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co- rP- 26aos -0 *- c- 007 e en-53-etP=00t- TRIP SETPOINT CORRECTIONS (p. 12-17)  ;

? PROTECTION SYSTEM DESIGN PHILOSOPHY

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PURPOSE STATEMENT: _

THIS LESSON DESCRIBES IN BASIC TERMS THE METHODS USED IN PERFORMING ACCIDENT ANALYSIS. THE LESSON FAMILIARIZES THE STUDENT WITH CERTAIN ACCIDENTS PRESENTED IN THE FSAR. IT ALSO ENCOMPASSES THE OPERATORS ROLE IN PREVENTING AND/OR MITIGATING CORE DAMAGE.

II. LIST OF OBJECTIVES: -

/ 1. Describe,{ genera terms, the concept of accident analysis.

2. State the goal of accident analysis.
3. Describe the concept of risk, and with the aid of appropriate information, calculate risk factors for a given plant accident.
4. Describe the multiple defense concept of reactor safety.
5. State the three physical barriers to fission product release.
6. State two core thermal limits and basically describe how each liNit is ~

determined.

7. Define DNBR and state its minimum acceptable balance. ~
8. State four factors in the core affecting the local proximity to DNB.

i 9. Draw the Safety Limit Curve '(Tech. Specs. 2.1-1) and Describe the curve including' bases for each line segment for the 2250 psia curve and how the curve encompasses the fcur factors listed in objective 8.

10. Describe, in general terms, factors used to determine Rasctor Protection trip setpoints.
11. State the ECCS acceptance criteria specified in 10CFR50.46 and the significance of each limit in pr core damage.
12. State the M en accident analysis and Technical Specifications.

A, ~"

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n 4.6

13. State the T - i begween accident analysis and ESF System design.

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REFERENCES:

1. " Mitigating Core Damage Training", Chapter 13, Item 13.2.1.1.6 Final Safety Analysis Report, Volume, Vogtle Electric Generating Plant, Revision 19.

September 1985.

2. " Mitigating Core Damage", Item II.B.4, NUREG 0737 Enclosure 3.
3. " Introduction", Chapter 1, Mitigating Reactor Core Damage, General Physics Corporation, 1981.
4. " Introduction to Accident Analysis", Transient and Accident Analysis, Yol I. ,

Chapter 12 Westinghouse Electric Corporation, Revision 1.

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111. LESSON OUTLINE: NOTES I. INTRODUCTION A. Lesson Intent 77 001

1. Provide the student with and understanding of the methods employed in accident analysis including:
a. Familiarization with accidents specified in the FSAR.
b. Purpose of accident analysis and methodology.
c. Relationship of accident analysis to both the operator and general public.
2. The student will gain an understanding of the con-capt of risk as applied to accident analysis as well as conditions that could lead to core damages. ,

B. Lesson Content

1. Risk analysis and accident analysis methodology.

~

2. Review of core thermal, limits, hot channel factors, and DNB. .

II. PRESENTATION

~

A. Purpose of Accident Analysis and the Concept of Risk Objective 1

_AL*__t.Ia '1

1. Fundamental purpose of accident analysis is to N""~'~

demonstrate to the general public that they will be protected from harm as a result of an reactor accident.

2. Principle source of danger is from the release of radioactive materials to the environment.

, a., ,This the fundamental design of the training of operating personnel.

q'p,plemt, v .

.g.~

.bDdBa: , it must be shown that only a g, vest ~ chance exists, that an accident t ,. t the plant from returning to

~ power ' ation following the accident. -

B. Concept of Risk , Objective 3 l 1. Definition in accordance with WASH-14DO report of '

1974.

. 4

. L o-t-P- 36 00 n -co L

=

Ill. LESSON OUTLINE: NOTES

_ > a. RiskfFrequency X Magnitude where: Risk is consequences [gn d flyw f.

.- f frequency is events /()hki "IVHf.

M **

Magnitude is consequences /CV6%I

  • W
b. Example:

In 1930, approximately 25 million automobile accidents occurred in the United States. One out of every 400 accidents resulted in death.

Therefore the risk of death in an auto accident is: -

6 Risk = (25x10 accidents)(1 death ) .: _

, yr 400 accident Risk approximately 63, 000 deaths each year

c. Assuming a population of 270 million people the risk of death per person in any year is:

Risk (perperson)=63,000 deaths /yr[

4 270x100 persons 2.3x10 deaths /g p, p6 W -yr

2. " Studies t riska from all types of events s * ', f. .

rk,. .

~

. ~ M deaths / person-yr.

. } }~~ Q", ;tfl; aJ.yg .2vat:

, M . .ater is unacceptable to society pobablyresultinimmediateaction

' 't'4 rednes'the risk to an acceptable level. -

As risk decreases, society (people) become ,

less concerned and are move apt to accept ~ ~

the risks.

. 5

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.- w -tp %so: 00~

Sr ' -

077 lll. LESSON OUTLINE: NOTES

b. It has been determined that if the risks from nuclear plants can be shown to be less than 10 deaths / person-yr, society may be willing to accept the risks.

f

[ In addition if the benefits of nuclear power can also be shown, the public may be even more willing to accept nuclear power if they receive something in return.

3. Risk evaluation is complicated by the publics greater concern with catastrophic events.
a. Example: Commercial airline crashes were above average in 1985, but the number of deaths were still far below, those from automobile accidents. However, the news media reported the big accidents, not the individual automobile accidents.

g j

b. Since severgauclear accidents could result i in a catastrophic release of activity it must .: _

l be shown that these accidents have a very low 1 probability of oscurring. .

4. Risk from Radiation Release Calculations Objective 3
a. Risk = PCD where: P is probability of accident (events) yr C is number of curies of released for
a particular accident (curies) event D is number o,f deaths postulated to occur per curia released from

. . icular event (daaths) x gy '

"J r curie i Idl- on factors such as

s,.

. -)Q?

2) ease atmospheric conditions Rffect of type of radiation released on the body. .

l c. Note that if any or all of the forms increase',

the risk of the event also increases.

I Risk - PCD l

l

. 6

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lil. LESSON OUTLINE: NOTES

5. The goal of safety analysis is to show the total Objective 2 risk to the public from all accidents is less than an acceptable valve.

I

a. Note that high probability events must have low release rates and vice versa.
b. High probability events must show that release i rates are within the limits of 10CFR20 limitations.
c. Extremely severe events must show that release rates are within the higher limits of 10CFR100.

B. Reactor Safety Concepts

1. Multiple Defense Concept used in power plant design 0bjective 4
s. Build and operate the plant in such a way to minimize the likelihood of accidents. -
1) Proper design, construction techniques ,.

and suitable, materials.

2) Strict quality' control to maintain high standards.

~

3) Properly trained operators.
b. Assume that accidents can happen despite all precautions taken.
1) Design in reactor protection system.

i 2) Shutdown should occur before significant damage occurs. ,

"c. Aaouns.

l accident.

[; . 3:.

venyts

. .dijg ' @q -entremely improbable but potent we t} S W that ESF system can protect the

. yM" ' ^ - a g or a least limit the damage from l C M 'ina accident that the public remains

<< V M ted.

2) It must be stressed, that even though ,

large amounts of money and effort are -

used to protect the public 'from such

~

an accident, the accident is not expected to ever occur.

. 7

~ - - _ _ _ _ _ _ . . - , _ _ _ _ - _ _ _ _ _ _ . _ _ _ . - _ . _ _ - __ _ _ ___- _ _ _ - _ . _ _ _ . _ -

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,- g.,0 . Q - 36061 - Od -'

ak-G--;77 lil. LESSON OUTLINE: NOTES

2. Major Goal in Plant Design is to prevent the release of fission products to the environment. {
a. Barriers to Fission Product Release. Objective 5
1) Fuel arrangement and cladding.

Extensive design effort goes into proving the integrity of this first barrier.

2) RCS pressure boundary.

a) Tech Specs limits RCS pressure and heatup and cooldown transients to minimize stresses on the vessel and piping.

b) Extreme brittle fracture studies ensure the vessel remains ductile.

3) Containment Building a) Designed to limit the release of fission products to the environment.

to an acceptable level.

, b) Pressure limits based on Design Bases Accident Studies.

C. Core Thermal Limits - B and PCT Objective 6

1. Core integrity can be lost if the fuel or clad is allowed to overheat.
a. Overheating can be caused by:

~

1) LOCA, until Emergency Care Cooling is
ted and becomes effective.

a:,

C A,g ,[1!ooling capability decreases below reduction.

g '4. 1 g N"$,;. ~ V.

.f :.,Q~ transients cause heat production te anceed core cooling capability.

2. Departure from Nucleate Boiling (DNB) - Limit 1 ,
s. Critical Heat Flux: The heat f1'ux that resul'ta in a rapid drop in heat transfer ability, as measured by a drop in the heat transfer coefficient or by an increase in the clad and fuel temperature.

8

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lli. LESSON OUTLINE: NOTES

1) At this heat flux - the core is said to be emergency DNB.
2) This is a local condition dependent upon  !

the local heat flux at the clad surface.

b. DNBR
  • Objective 7 l
1) DNB must be avoided to prevent core i overheating. l
2) The proximity of the core to DNB I conditions is called the Departure From Nucleate Boiling Ratio.

a) DNBR = Predicted Local heat flux actual Local heat flux

-f to _ cAusa. DNB i b) Note the word " predicted"

- The local heat flux necessary to cause DNB cannot be exactly - _

provi,ded.

- Local parameters depend on too many variables

~

c) Typically, the designee conducts l experiments to measure the heat flux (for a given pressure) i to cause DNB. -. .

This results in a fanily of curves 4R=TF=0PP=66t-that any be used to provide TP-CO2, I operational limits on the specific '

core. .

s ,, 3 3 ion and CNF sf ~~

. If, a relation is used by Westinghouse s 9%

  • ict the CHF (q")

A.@ TF-Oo E-MAL

_A 4 ,,, ,k-perfect, would assume the s- -'

' relationship shown in 48-WP=0Pf=60t- r -- ;; ani w 77 3&oos.co-cales,

- q" predicted = Q" measured ,

- DNBR = 1.0 i

9 I

. . . - - .,-~,------..r..- ,n.-, _..__.__,,_____,,,_,-._m._ - _ __._ ___-__,- m.-- - , . . _ - _ . _ _ __ _ . . - - _ _ . . _ _ _ -

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lil. LESSON OUTLINE: NOTES b) Since correlation is not perfect, the designers must allow for error in i the q" predicted. _ _ _ _

i

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- For W-3 correlation, a DNBR of "" " 0 7 ' ^ ^ ' l 1.3 must be used. T7-OD)

DNBR = q' predicted 1.3 q" measured

- At DNBR = 1.3, there is 95 percent probability that DNB is not taking place in the core

- 95% confidence level says that if experiments were repeated the data would agree favorably with previous experiments at least 95% of the time.

_? C'l

2) CEF Dependency on Core Height 6 a) Graph axis

.: Q 00'l

- horizontal - core height (s)

- vertical - core heat flux (BTU )

hr-ft" b) Two' curves on graph. '

- Lower curve shows heat flux in the curve as a function of

-l core height.

(measured CEF)

  • Function of time in core life

~

and core conditions t .c '

l f' "

  • May be a plot of the highest y .h y heat flux encountered during an

' . I,' ES accident i D y&Wj: . ei - Upper curve shows predicted CHF

- -** eWQ by CHF correlation. -

  • Decreases with increasing core ,

height ,

  • Due to higher coolant enthalpy at i

the top of the core due to the higher coolant temperature.

l l

l 10

. _ . . _ _ _ _ _ . - . _ _ _ _ _ _ __ __ _ _ _ J

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ll1. LESSON OUTLINE: NOTES c) The proximity to DNB is the points '"-T" -

O "-003 where the curves become the closet & ^^}

(see point Z.)

Q 00tf

- DNBR = q" predicted (z) q" measured (z)

- DNBR must be equal to or greater than 1.3 to ensure that DNB does not occur in the core, d) Note that the DNBR is usually lowest i

in the upper regions of the core.

Y

% e) FactorsaffectingDNBRage Objective 8 Core local power (near core counterline)

RCS pressure (lower in upper core)

Core local Temperature (higher in upper core) .

Core flow rate (virtually constant)

d. Core Safety Limits Curve (Tech Specs) Objective 9 SR-3P-9M=004-
1) Curve provides operator with information f t ^ "f regarding proximity to DNB utilising the TP-005 l parameters previously discussed.

} 2) Power and Tave are two axes and a family of possible system pressure curves are

., T ,

  • m ',

J' ire graph is based on four loop M(,

therefore all the parameters ,

g .n 3 Ot 17 discussed are used to operating limits.

l a l~%% m e. .

3) Pressure Curve Description (2250 pois)

Three straight line segments I a) *

  • i (A,B,C) i 11

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I yo - c,.y - 34 col -o O ~L

-SR-LP-9M ill. LESSON OUTLINE: NOTES b) Segment A 1

- based on having no bulk boiling  !

at the core outlet.

- basis is OTAT and OPM trips i are only effective when core l 4T is proportional to power.

- line A is plot of Tave that ensures bulk core outlet Temperature is less than Tsat further ensuring no boiling is taking place and ,

AT is a true indication of power.  ;

l

- Obviously the line must slope i downward since Th diverges from Tave power increase.

c) Line Segment C

- generated by the W-3 correlation  !

that is it projects the DNBR  : _

limit,of 1.3 ,

  • Staying'below and'to the left of this curve ensures DNBR is 1.3 .
  • 1
  • Assumes no 'aoiling occurs in j

the core 1 l

- Curve C is more limiting than curve A at higher powers since  !

  • DNB is a localised concern I
  • DNB can occur even though bulk

' boiling is not occurring.

Wd lineSegmentB

})!' Y iJ fj~';' "'s-represents the point where the

" ' 4+q;'

L";f.. ' [ - limit on vapor quality at the

-> hottest fuel assembly outlet U'%'.'b' M , "

becomes more restrictive -

than the vapor content in the hot leg. ,

l

. 12 l.. _ _ _ _ _ _., _ .-_ _ ___ _

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Ill. LESSON OUTLINE: NOTES 1.e., assumes that significant Localized boiling can occur even though the total core outlet is subcooled.

- limit assumes 15 percent vapor content at core outlet to ensure DNB will not occur even with significant local vapor quality. 1

\

- Note that this limit is only l

found on the higher pressure '

curves (2250 psia and higher)

3. Peak Centerline Temperature (Limit 2) Objective 6 l
a. In addition to DNB, fuel temperature must be limited to prevent fuel melting.
b. Irradiated fuel melting temperature is approximately 5080'F. .: _

N 1) Irradiated Fuel melt temperature decreases by 58'F for every 10,000 burnup MTU

~

2) To protect fuel remaining in the core for three cycles limit is set con-servatively to 4700*F.
4. Susmary of thermal limits Objective 6
s. Two thermal limits of most concern

_  ? 1) DNBR 4Mpp l.3

  • Note: these limits apply only to

. 2) 700'F* accidents with pr .

4 relatively high bad t protections probability of V: occurrence

,I ,g17',4.. otection

~ W'gA(4 ,

D.'s) OTE trip in conjunction with -

b) Loss of flow trips and ,

c) Low pressure trips i

13

_ -__-_--._ ______.______ _._____.-___ _____ L _.___ _ _-_.__ _ _ _ _ _ _ _ - _ _ - _ . _ _ _ _ _ ._.._ ___._ _ _ _ _

." J -(f 36001 k 1

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l Ill. LESSON OUTLINE: NOTES

2) PCT protection a) High neutron flux trip is primary with backup by b) OPAr
c. Accident analysis ensures that these trips l anticipate and provide protection quickly I enough to prevent core damage.  !

D. Accident Analysis Methodology

1. Initial step is to determine limits on plant parameters, which if exceeded could result in core l damage.
a. Once parameter limits are identified transients that could cause the limits to be exceeded may be identified.
b. The designer most also investigate and evaluate the response of the protection system - -

against those parameter limits and transients.

2. Sound accident analysis requires conservatism is estab Hshing initial plant conditions.
a. Limiting values of plant parameters are chosen that *ber:
1) Make the transient more severe.
2) They originate within their normal 7 operating ) k s.
b. Instrument errors used in accident analyses.
1) , ,z 1.02 (Calometric errors)

., 3 . . y'a .

t'l ' $) Tavire 4*F (whichever is most severe) 9,.s.'d,'_.,. p ,S W - Rod Control Deadband and 2.5'F ation error)

I 3)' MineEry Pressure 2 30 psi (whichever is moet severe) l c. Once the initial conditions are established l l the subsequent analysis must sho*w that the

14 l

l

t.o -(.9 - 3 600 t -C0 -C w w-lil. LESSON OUTLINE: NOTES

d. Operators must maintain the operating limits specified in technical specifications to ensure the validity of the accident analysis.
3. Reactor kinetics are also used in accident analyses.
a. The parameters used incorporate.
1) Moderator Coefficient (C(m) 1
2) Doppler Coefficient (o(D) l
3) Boron worth (otB)  ;
4) Delayed Neutron Fraction (Beff)
5) PromptNeutronGenerationTime(j*)
b. Again conservative valves of these parameters are used to maximize the severity on the transients to be analysed.
1) BOL vs EOL ,,

~

2) esp y HFP .
3) ARI vs ABO

~

(ARI refers to rods at the rod insertion limit) ,

c. Choosing the most limiting parameter may result in inconsistencies for core life considerations but it is desired for con-servatism.

TPM M

1) 4P-446 shows the Doppler-Power Coefficient SR-TP=9 M-006 Curve used in Safety Analysis. TF- 001 s .ry%

27.,2 1beisarve illustrates the concept

4. 3 'jo( h g conservative valves for physics n...S- $3 d (( ,3*,==
4. on Trip
a. Assumptions conservatively underestimate their beneficial effects of insertion. ,
b. Assumptions gAOf fr- #
1) Bottom skewed axial power shape assumes N delay in negative reactivity insertion.

l l

15 l

'.' - 00-Lf-](Co(~Q0~C

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l11. LESSON OUTLINE: NOTES

2) Maximum rod drop time of 2.2 seconds. ,

(Specified by tech specs) i l

E. Calculation of Trip Setpoints Objective 10 l

1. Trip Setpointo chosen by accident analysis allow for possible errors.
a. Once conservative trip valves for accident i protection are determined it is possible to work backwards, accounting for errors, to j obtain the trip setpoint specified in Tech.

Specs.

b.

77 OG

.l Example: Power Range Neutron Flux High Setpoint N l

Safety Analysis limit 118%

T7-007 j Process Measurement Accuracy j 1). Calorimetric Accuracy - 2% - '

2) Power Shape effect - SI .. _

4 l Calibration' error' - 0.5% .

Bistable Accuracy - 0.251 Technical Specification Allowable Setpoint 110%

Drift Allowance -11

N=4mm1 Trip Setpoint 109%

F. Relationship Between Safety Analysis and Technical Objective 12

Specifications.

l 1. FM , " ' demonstrates the ability of the n agt; to protect the public in the Mayy.-

' lye ~ TA'~c-

@ y formed assume worst case conditions.

j mb

% as prestausly discussed and, -

1

b. The plant is alway operated within the ,

limitations specified in Technical Specifi- -

cations. *

  • i l

i 1 16 l

i

i.o - t f- 3 6Cc(- co - C.

- ..m>=

Ill. LESSON OUTLINE: NOTES

2. Selected Tech Specs
a. Safety Limits curve TS. 2.1.1, Figure 2.1-1 M S" *" 077 ^,Cr4 (Previously discussed) p.coj
b. Limiting Safety System Settings - TS 2.2-1
1) Note the trip setpoint and allowable drift allowance.
2) Setpoints selected to prevent reactor core and RCS are prevented from exceeding their safety and transient operation.
3) In addition, the trips assist the ESFAS in mitigating the consequences of accidents.
c. Power Distribution Limits Section 3/.4-2

% D 7 1) l AFA(s)imits F (2.3 xassure upperaxial normalised bound limit of peaking q ,:

factor) is n'ot exceeded during normal .

operation and during xenon redistri-bution following a transient.

2) F (s) - Heat Flux HCF and F =

E0thalpy. Rise HCF limits enAle a) Design limits on peak local power density and minimus DNBR are not exceeded, and, b) In the evest of a LOCA, the peak fuel clad temperature will not exceed the 2200'F acceptance

., criteria limits.

, &c ., ' . . v, radial power distribution are (gp ~.

$- design 7, w valves.

l

4) DNB Parameters a) Limits assure that each of the -

parameters are maintai'ned within  ;

the normal steady state envelope of operation assumed in the accident analysis.

l 17

l O - C.(#. .3,b,60

_ , ' - C0 4 lil. LESSON OUTLINE: NOTES b) Limits will ensure DNBR 1.3 in maintained

- Tave 590*F

- Pressure (Par) 2219 psig.

- RCS Total Flow 382,800 gym

3. If the calculations for the accidents of chapter 15 show unacceptable results, the following options are available.
a. Change the RP or ESF systems trip setpoints.
b. Charge tech spec limits.
1) Changing allowable power shape (F (s) or 1J .
2) Lowering RTP
3) Changing DNB, parameters.

C. Relationship of ESF Systems to Safety Analysis Objective 13

1. RPS will provide protection for most plant accidents.
2. For those accidents where the RPS is inadequate, the ESF Systems function to:
a. Prevent or limit the level of core damage, and t
b. Prevent or minimise the release of radioactive materials to the atmosphere.
3. ECCS Acceptance Criteria; 10CFR50.46 Objective 11 l the criteria upon which the s.4*( N eapabilities of the ESF systems are -

1

. E taget..~  !

if.SW dk-b 4,$ W 4riteria are:

1 I

s Q e A- . \

1) Peak cladding Temp 2200*F

- Prevents loss of clad int,egrity f l

2) Cladding oxidation is limited to 17 I

percent of existing clad thickness l l

- Prevents excessive loss of local clad i strength and ductility. I i

18

." uo- te. 3sec t - co -f re ir 077 111. LESSON OUTLINE: NOTES

3) g generation is limited to 1 percent f the amount that would be generated if r'L A all the Zircaloy in the active core region would react.

This prevents:

1 a) Accumulation of explosive gas mixture in containment and, b) Non-condensible gas accumulation in the RCS that could result in a loss of core cooling.

4. Core remains intact and coolable.

- Prevent cladding and fuel failures from blocking coolant changes.

5. Core remains in a configuration which can be cooled for a long period of time. -

t

- Ensures that Decay Heat Removal can be continued - -

to prevent further core damage.

H. 10CFR Limits

, 1. 10CFR20 deals with normal occupational or nonoccupa- j tional exposures.

a. 10CFR20 also gives release limits for various types of radionuclides.

{

b. 10CFR20 limits will not be exceeded by Condition II eventa
2. 10CFR100 limits deal with reactor require-ments.

a.# Ldadas. boundaries of exclusion zone r?.1 sa d. Eu g

~

lation zone.

4,%. .y b C hipeesse es for an exclusive area are

  • . 'V . 18mt no more than 25 REN whole body

and'300 to the thyroid from iodine for -

the first two hours following the accident.

c. Exposure limits for a low population zone are!

the esse except they are for the* entire ~

duration of the transient.

19 I

co - W - 3f.cca - CO C '

e -u m:

Ill. LESSON OUTLINE: NOTES I. Operators Role in Plant Safety

1. To understand the influence of the operator on plant response during an accident, one must i understand plant conditions at the start of the i

event.

IN FSAR. it is generally assumed: l

a. Plant is operating inside the allowed control band.
b. Plant is at steady state conditions prior

{ to the start of the transient.

c. Plant is operating at the most limiting conditions and parameters as allowed by Technical Specifications.
2. The FSAR Chapter 15 is essentially a worst case 1' accident study assuming the most limiting valves - l for plant operating parameters.
a. The operator, must realize however, the con-servatism built into the FSAR analysis. .
b. The operator must understand that transients l, based on "best estimate" parsmaters and initial conditions may proceed differently than those that actually taka place in the plant.
3. Role of Operator i
a. Normal Operation
1) Flant Operators in an assumed band with -SR.3D 4PfHW8

~

significant margin to the various reactor M

... tion trip setpoints. (740$ 1

'f'

? Q.4

'+'

arms alert the operator to J(g.fi

i developing off-normal conditions ,

..,* (i.e., control system failure) '

b) ~ The operator acts as backup to plant control systems to ensure the first two assumptions discussed in ,

1.1. a thru e are true. .

20 t

e

- -----e--,y-r--+.-~=----we-- , ~ , , . _ , - re,---,

_ (o - Q - Acc i-CC d 5-u-e ^a-111. LESSON OUTLINE: NOTES

2) Plant operation is allowable in mode 1, with certain deviations ao long as they are within the bounds of Technical Specifications.

i The deviations include su~ch items as:  !

a) RCS leakage l

b) RCS activity c) Reactor Power Distribution d) Safeguards components out of service.

3) If the operator allows the plant to operate outside the bounds of Technical Specifications and therefore out of the bounds of accident analysis, certain key initiating events could result in a potentially damaging situation. ~

a) Example: If the plant is operated at -

100% power with several steam .

generat'or safety valves out of .

Service, a complete loss of load event could result in plant

, overpressurization.

b) This event then goes from one that may have allowed a fairly rapid plant restant, to one that results in significant downtime to repair any damage and to perform required analyses to verify the safety of plant startup.

b. Key issues being presented.

A ltb))'j the plant to operate with 1 equipment lineups or outside

-J,,3pr 4., the boundaries of Technical Specif1-T des,Mg* '

~ emetens, the

- -

  • ebdesecident operatorand analyses, mayendanger invalidate the health and safety of the general -

public.

2) The operators most important function then is to combat the unexp'ected.
  • 21

, . a 3 s e.en - C O 4

=- .N7 7 lil. LESSON OUTLINE: NOTES a) Even if the plant is operated as designed, and all Tech Specs and administrative limits are observed, a chance exists that any given initiating event may degrade into an accident scenario not analyzed for in the FSAR.

b) In these instances, operator action is required to mitigate reactor core damage. l

c. The operator must, above all, be able to apply his knowledge of the principles and Fundamentals of plant operation to evaluate a sequence of events different fr.

t

1) Past industry events clearly demonstrate the need for this skill.

2). The remainder of this program, will deal -

with operator and equipment response to e mitigate the effects of Core Damage. , ,

~

III. SIDDIARY .

Review Lesson Objectives.

O S

V em

% d D.?i:;.

,. .4 ,

  • ? n 1

22 i

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TRAINING MATERIAL ROUTZNG L O - < p.3sc o n c,_

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k N q ..;m gg s e pMATER REVISOR S. 6mdm /

REASON FOR REVISION:

MAJOR REVISION DUE TO ERRORS OR OMISSIONS.

REVISION DUE TO CHANGES IN EQUIPMENT.

X REVISION DUE TO CHANCES IN PROCEDURES /0PERATING INSTRUCTIONS OR POLICY.

DESIRE ADDITIONAL GRAPHICS / HANDOUTS FOR THIS TRAINING MATERIAL OTHER COMMITTMENTS: TES k NO DRAFTING REQUEST FILLED OUT. TES NO X DATE SUBMITTED FOR SUPERVISOR'S REVIEW 7/21/f4 REVIEW SAT UNSAT APFROVAL SIGNAT /kHw1 DATE '7

          • DATE NEED E FOR EW FROM TTFDG d *****

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LIBRARY CLERE: 677)

DATE TO TYP DG 7U/

DATE TO DRAFTUG A//M DATE FROM TTFING DATE FROM DRAFTUG A/ / A j COMPILED MAftRTALs TO INSTRUCTOR FOR REVIEW. DATE 1

l USTRUCTOR REVIENs TTFING SAT UNSAT DRAFTING SAT UNSAT l REWORR nETT . ,

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SUPERTISOR E @ . TTFING - SAT UNSAT i UNSAT

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LIBRARIAN RECEIFT FOR INCLUSION /0UTDATING OF FILES. DATE 1 I

FILE WORK DONE DATE 1

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