ML20097F991

Draft Chapter 6, Pressurized Thermal Shock Integrated Risk for Calvert Cliffs Unit 1 & Potential Mitigation Measures, of Pressurized Thermal Shock Evaluation of Calvert Cliffs Unit 1 Nuclear Power Plant

ML20097F991
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Site:	Calvert Cliffs
Issue date:	08/21/1984
From:	OAK RIDGE NATIONAL LABORATORY
To:	NRC
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ML20097F969	List: ML20097F984 ML20097F991
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR NUDOCS 8409190214
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Text

v A PRESSURIZED THERMAL SHOCK EVALUATION OF THE CALVERT Q.IFFS UNIT 1 NUQ. EAR POWER PLANT mo e,

,: i _~,

.: 2 List of Chapters b.i w LJ:

Chapter 1 Introduction Chapter 2 Calvert Cliffs Unit 1 Nuclear Power Plant System Description Chapter 3 Development of Overcooling Sequences for Calvert Cliffs Unit 1 Nuclear Power Plant Chapter 4 Thermal-Hydraulic Analysia of Potential Overcooling Transients Occurring at Calvert Cliffs Unit 1 Nuclear Power Plant Chapter 5 Probabilistic Fracture-Mechanics Analysis of Calvert Cliffs Unit 1 Sequences Chapter 6 PIS Integrated Risk for Calvert Cliffs Unit 1 and Potential Mitigation Measures 1

Chapter 7 Sensitivity and Uncertainty Analysis f Chapter 8 Summary and Conclusions 9

E w

4 k

P 8409190214 840828 s

PDR ADOCK 05000317 PDR P

i .'  :

6.0. PIS INTEGRA11!D RISE POR CALVERT CLIFFS UNIT 1 AND POIllNTIAL MITIGATION MEASURES T , c.

,,.;a

~

j.

6.1. Introduction L,, 3 6.2. Risk Integration 6.2.1. General Approach and Resnits 6.2.2. Dominant Risk Sequences 6.2.3. Relative Importance of Each Category of Sequences as Initiating Events 6.3. Effects of Potential Risk Reduction Measures 6.3.1. Introduction of a High Steam Generator Trip System 6.3.2. Reduction of Neutron Fluence Rate 6.3.3. Heating of the HPI Water 6.3.4. In-Service Inspection of Vessel 6.3.5. Annealing of the Vessel 6.3.6. Improvement in Operator Training 6.3.7. Maintaining RCP Operation During Secondary Side Overcooling Transient 6.3.8. Summary of the Effects of Potential Risk Reduction Measures i

I 4

I l

l l

LIST OF FIGURES l--

,,- a

'-L u 6.1. Risk associated with five dominant sequences.

6 .2 . Risk associated with each category of events.

6.3. Effects of potential risk reduction actions. (Note:

The top (solid) curve gives the total risk if no risk reduction (RR) actions are taken. The remaining carves show the total risk when specific risk raduction actions are taken.)

i k

i _ -_ . -

... 1 1-LIST OF TAB 128

~h !1 %:: a '

. 2 Ij

' 6 .1. Summary of risk intogratf on !l d[!'hh 6.2. Summary of risk vs EFPY, F,, and RTg for dominant risk sequences 4

1.

l l

i i

l.

i - . . _ ._ _ - _ _ _ _ _ _ _

CC-6el

, h, ,)..

y :.<f . , -

, - i nq ; . , J.*C~.y . m .

. ', 3 l 6.0. PIS INIEGRATED RISK FOR CALVERT (1,IFFS UNIT 1 AND POTENTIAL MITIGATION MEASURES 6.1. Introdnotion The preceding three chapters have outlined the procedures employed to esti-mate the three fundamental parameters (transient frequency, thermal-hydraulic history, and the conditional probability of vessel failure) required to quantify the PTS risi sesociated with a transient in Calvert 4 Cliffs Unit le This chapter discusses the means by which these three influences are integrated to yield an estimated frequency of vessel failure l

(through-wall crack penetration). Section 6.2 describes the risk integra-i f

tion process and identifies the dominant risk sequences, as well as the relative risk of different classes of transients, and Section 6.3 discusses the effects of potential corrective actions.

6.2. Risk Intearation f 6.2.1. General Approach and Results n

O i

i The frequency of a thronsh-the-wall crack associated with each sequence i'

identified in Chapter 3.0 is obtained by multiplying the sequence frequency by the appropriate conditional possibility of a through-the-wall crack I,.I Li presented in Chapter 5.0. The results of this exercise are presented in .g.

Table 6.1 for two conditions: (1) 32 effective full power years (EFPY), or ENUT + 2e = 251*F, where Eg is the nil-ductility reference temperature, l ,,

and (2) th. roint in it.e wh.a n NUT + 2e = 270*F.* l,/

• The 170'F date are presented to preside informatten es plant risk when the reestas eriterie are reached. The maist weld screeslog estae wee seed rather them the 300*F sirsamferentist veld valse staae the eaatysis clearly ladleeted that the streamfereettet velde did met engstf t-eastly eestribete to the F15 risk. It shoold be seted that Calvert Citffs Omit A is met expoeted to reach the eersentas eriteria daring the present 9

i 11 sensed life of the pleet.

MsK5y$-c~-v- =-- g g.g "fE' 4M 374 Q M W.OCOS TF PM *.7-*7e 3,- w - . ,,- - Qf 5 7--^^.;

a , ,s-,-

i Tabis 6.1. Summary of risk integration 32 EFFT (RTg + 2e = 251'F) RTg + 2v = 270'F  !

Estimated Number Used __ Reak Segnesee Sequense . . . .Through-the-Well__

Segsease for Conditional Conditional Orderlag Conditional Threesh-the-Yell Sequesee Frogsemey Fettere Fallare Creek Fregsemey of Risk Fallare Creek Fregneasy Numbert (y,-1) Probabilityt Probability (yr-1) Due to PT5 Probab!!!ty (yr-1) 1.1 2.8E-4 1.1 3.0E-8 8.4E-12 1.2 3.7E-6 1.2 9.0E-8 3.3 E-13 1.3 3.8E-6 1.3 6.0E-7 2.3 E-12 4.9E-6 1.95-11 1.4 3.85-6 1.4 3.3E-6 1.3E-11 1.7E-S 6.5E-11 1.5 3.4 E-7 1.5 3.0E-S 1.0E-11 1,25-4 4.15-11 1.6 2.45-7 1.6 S.1E-S 1.28-11 1.7 7.75-9 1.7 1.9E-4 1.SE-12 4.8E-4 1.8* 3.7E-12 4.05-8 1.8 2.5E-4 1.0E-11 6.25-4 2.35-11 2.1 3.85-3 2.1 2.0E-7 7.6 E-10 $ 1.4E-6 S.3E-9 2.2 S.05-5 2.1 2.0E-7 1.0E-11 1.48-6 2.3 S.15-5 7.05-10 2.4 1.7E-S 8.7E-10 4 6.8E-5 2.4 S.15-5 3.$E-9 2.4 1.7E-S 8.7E-10 3 6.85-5 3.35-9 2.5 4.65-6 2.5 7.6E-6 3.58-11 3.25-5 2.6 1.3E-10 3.2E-6 2.6 8.2 E-6 2.6E-11 3.45-5 1.1E-10 2.7 8.7E-8 2.7 1.8E-4 1.6 E-11 4.SE-4 3.9E-11 2.8 1.C5-7 2.8 2.3 E-6 2.3E-13 1.1E-S 1.18-12 2.9* 5.05-7 2.7 1.8E-4 9.0E-11 8 4.55-4 3.1 2.35-10 8.SE-4 B** (I E-9 (8.SE-13 S.2 1.18-3 , _, 3 (IE-9 (1.1E-14 3.3 1.1E-S B (IE-9 (1.1E-14 3.4 1.1E-S E (IE-9 (1,1E-14 3.5 1.0E-6 3.5 8.0E-7 8.0E-13 3.6 7.0E-7 3.6 7. 2 E-6 S.0E-12 3.6E-5 2.55-11 3.7 S.SE-6 8 (IT-9 (S.SE-15 3.8 3.7E-6 B (IE-7 (3.7E-13 3.9 3.7 E-7 B (IE-7 (3.7E-14 3.10* 7.0E-7 3.10 6.7E-S 4.7E-11 9 4.1 1.1E-2 8 (IE-9 (1,1E-11 4.2 1.5 E-4 B (IE-9 (1.SE-13 "

4.3 1.SE-4 4.4 1.5E-4 B

B (IE-9 (IE-9 (1.SE-13 (1.SE-13 , ~m

).

4.5 1.35-S B (1E-7 (1.35-12 '" u "l]

l i

i

2 ."

• Table 6.1. (Contissed) 32 EFPY (RTg + 2e = 251'F) RTg + 2e = 270'F Transient .

Estimated Number Used Segsence Rank Sequence Segsease for Conditional Conditional lThrough-the-Tall Ordering Conditional hrough-the-Tall Begneste Fregeeney Failure Fa!!ste 8 Creek Frequency of Risk Fallare Creek Frogneast Numbes* (yr-23 Probabilityt Probability (yr-1) Due to FT5 Probability (yr-1) i 4.6 9.58-6 4.6 2.0E-7 1.95-12 1.05-6 9.55-12 4.7 5.0E-5 B (IE-7 (5.0E-12 4.8 6.55-7 B (1E-7 (6.5E-14 4.9 6.15-7 B (1E-6 (6.7E-13 ,

4.10 5.0s-6 B (1E-6 (5.0E-12 4.11 7.25-5 B (IE-7 (7.2Fr12 5 4.12 9.7E-7 B (1E-7 (9.7E-14 4.13* 6.25-6 4.13 6.0E-6 3.7E-11 10 ,

5.i] 5.4 5 (1E-12 (5.4E-12 .

5.2 4.65-2 3 (IE-11 (4.6E-13  ! , _ ~

5.3 1.25-3 5 (IE-11 (1.3E-14 5.4 2.3E-3 5 (IE-11 (2.3 E-14 5.5 6.25-5 B (1E-11 (6.0E-16 ,

5.6 1.0E-2 8 (IFr11 (1.0E-13 ise 5.7 7.65-4 5 (IE-11 (7.65-15 5.8 1.5E-4 5 (IE-11 (1.5E-15 5.9 4.15-5 B (1E-11 (4.1E-16 5.10 1.55-4 B (IFr11 (1.5E-15 .

5.11 1.0E-5 B (IE-11 (1.0E-16 5.12 2.5E-6 B (1Fr11 (2.5E-17 5.13 7.05-7 5 (IE-11 (7.0E-18 5.14 1.5E-4 8 (IE-11 (1.5E-15 5.15 1.0E-5 B (IE-11 (1.0E-16 5.16 2.5 E-6 B (IE-11 (2.5E-17 5.17 6.8E-7 5 (IE-11 (6.8E-18 5.18 1.5E-4 8 (IE-11 (1.5E-15 ..

5.19 1.0E-5 B (IE-11 (1.0E-16 . . -

5.20 2.5E-6 B (IE-11 (2.5E-17 i.

5.21 3.7E-5 8 (IE-11 (3.7E-16 J~

i q=a t

s I +

3 - 4 Table 6.1. (Contissed)

~

32'EFFT (RTNUT + 2e = 251'F) RT[ 2e = 270'F ,

Estimated Number Used Sequence Rank Sequence Segsease for Conditional Conditional Through-the-Well Ordering Conditieset Throssh-the-Telt Beguesee Frequemey Fallare Fallare 3 Creek Frequency of Risk Fattare Creek Freguesey Numbert (yr-1) Probabilityt Probability (yr-1) Dne to PTS Frobab!!!ty (yr-1) 5.22 2.65-6 3 (1E-11 (2.6 E-17 5.23 5.05-7 D (IE-9 (5.05-16 5.24 1.85-7 B (15-9 (1.8E-16 5.25 9.05-6 B (IE-10 (9.0E-16 .

5.26 6.6E-7 5 (IE-7 (6.67!-14  !

5.27 3.35-7 B (IE-7 (1.3E-14 5.28 4.4E-4 8 (IFr11 (4.4 E-15 5.29 8.0E-6 5 (IE-11 (8.0E-17 5.30 2.05-6 8 (1E-11 (2.0E-17 5.31 6.85-2 B (IE-11 (6.8E-13

• 5.32 9.0E-4 5 (IFr11 (9.08-15 5.33 9.05-4 3 (IE-11 (9.0E-15 5.34 9.05-4 3 (15-11 (9.05-15 g  !

5.35 6.0E-5 8 (1E-11 (6.0E-16 *

, 5.36 3.4E-3 8 (IE-9 (3.4 E-12 5.37 2.0E-6 3 (IE-9 (2.05-15 5.38 1.05-6 3.6 7.2 E-6 7.2E-12 5.39* 2.0E-6 B (1E-9 (2.0E-15 5.40* 3.3E-6 B (1E-8 (3.3E-14 5.41* 4.6E-5 8 (IE-9 (4.6E-14 5.42* 9.0E-5 B (IE-8 (9.0E-13 J 5.43* 5.8E-5 4.13 6.0E-6 3.55-10 7 4.1 2.8E-4 3 (IE-9 (2.8E-13 I 6.2 1.08-5 3 (IE-9 (1.0E-14 1 6.3 1.4E-2 k (IE-9 (1.4 E-11 6.4 1.3 E-4 3 (IE-9 (1.3 E-13 6.5 2.6 E-6 B (IE-9 (2.6E-15 g.-N/-nq  ;

6.6 1.1E-5 3 (1E-9 (1.1E-14 j N 5] \

v. m

c. 1 i _ .

l

)

.i e

? . .

4 ,~ 8 Tante 6.1. (Coettased)

. _1 32 EFF7 (RTg + 2e = 251*F) --- RTNDT + 2e = 27 0*F i

Transient Istimated Number Used Sequence Rent Seguence Sequence for Conditione! Cendittoast i Through-the-Well Ordering Conditional Thressh-the-Tall Segmenee Frequeney Failure Failure Craek Frequemey of Rish Failure Creek Frequeesy Numbert (yr-1) Probabilityt Probability (yr~1) Due to PTS Probability (yr-1) 6.7 1.8E-4 6.8 3.6E-6 3 (IE-9 (3.68-15 6.9 8.0E-7 6.10 9.05-6 i 6.11 3.0E-5 , B (IE-9 (3.0E-14 -

6.12 2.4E-7 3 (1E-9 (2.4Fr16 6.13 3.7Rr7 6.14 2.28-6 5 (1E-9 (2.2E-15 1 6.15 4.4E-7 8 (IE-9 4.4E-16 i

6.16 1.25-7 3 (IE-9 1.2E-16 . ,-

6.17 6.0E-7 5 (IE-9 6.0B-16 4

6.18 1.05-4 8 (IE-9 1.0E-13  ; O i 6.19 1.0E-6 6.19 SE-6 5.0R-12

.  ?

m 7.1 1.05-3 B (IE-9 (1.08-12 i w

7.2 9.05-6 3 (1Fr7 (9.0Fr13 1

7.3 4.35-7 3 (1E-7 (4.35-14 7.4 1.2E-3 8 (IE-7 (1.2 Fe12 7.5 3.3E-7 8 (IE-7 (3.3E-14 7.6 6.05-7 B (IE-7 (6.0E-14 7.7 2.0E-6 8 (1Fe7 (2.05-13 7.8 1.5E-7 3 (IE-7 (1.5E-14 7.9* 2.0Er7 3 (IE-7 (2.05-14 8.1 1.0E-3 8.1 4.0E-7 4.08-10 6 2.2 Fe6 2.25-9 8.2 3.05-4 8.2 1.5E-4 4.5 E-8 1 2.9E-4 8.75-8 8.3 5.05-6 8.3 5.9E-3 3.0E-8 2 8.05-3 4.05-8 8.4 2.5 E-2 3 (1E-10 (2.5E-12

.... ..ei .eg.e.e... bE 3 t

5ee chapter 3 for definition of seguemee ===bers.

((Q

.i.8 ..ie.1.u.. ..e4 r. ti. .eg.e.e..

ig r.-

g-s~d 1

I

- - _= -

. .  ; CC-6.6 ,: .

- e 6y8;!!j.. j As noted in Chapter 4.0, a limited number (12) of event sequences were cal-culated in detail using the LANL thermal-hydraulic analysis code TRAC.

' These segaences in turn served as a basis for estimating the thermal-hydraulic histories of approximately 115 sequences. Fracture mechanics failure probatilities were assigned to each sequence from one of the fol-loving three data . sources presented in Chapter 5.0:

(1) Direct Analysis of Sequence - If the minimum temperature of the sequence dropped below 350*F and the sequence did not fall into Category 2 below, a specific fracture-mechanics calculation was performed for that sequence. The conditional vessel failure pro-bability reported in Chapter 5.0 for the specific calculation 1.

i

used in Table 6.1 and the sequence number is repeated in column 3 to indicate that the numbers presented are based on specific cal-

culations for that sequence.

i (2) Assignment of Value from a Separate Sequence - In Chapter 4.0 and l

Appendix J, several sequences were identified as having essen-

tially the same thermal-hydraulic profiles as another sequence.

! In this case a fracture-mechanics calculation was performed for i only one sequence and the same failure probability was assigned I

j to the other sequences in the group. In Table 6.1 the case l

i number of the calculated sequence is listed in column 3 to iden-i

{ tify it as representing the sequence listed in column 1.

i

'I l

- ,,-.,..v- . ,- , ,,,-- -

. . D M P ra-CC-6.7 Q'. ,Y / l O di.2 0 1 (3) Value Obtained from a Bounding Calculation - Many of the over 100 sequences involved relatively minor cooling of the primary sys-tom. Rather than perform a separate calculation for each of 1

these sequences, a series of bounding calculations were per- i fonned. As discussed in Chapter 5.0, these bonading calculations assumed a step decrease in temperature along with full pressure.

A bounding calculation resnit was used to represent a sequence if: (1) the minimum temperature for the sequence was greater o

than 350*F and (2) the use of a bounding calculation did not lead to a significant contribution to the total estimated plant risk due to PTS events. The use of a bounding calculation was con-sidered to be an over-estimation of the risk and thus the proba-bilities entered in Table 6.1 for these sequences are preceded by a " < " sign. The use of a bounding calculation for a sequence is indicated by the letter "B" in column 3 of the table.

The total plant risk due to PTS is obtained by saaming the individual estimated risks associated with each sequence or residual group as presented in Table 6.1. This total risk value was determined to be ~g z i

~I per reactor year (RY) at 32 EFPY and ~1.4 x 10-7 per reactor year when 10 the limiting weld reaches an EIWor + 2. value of 270*F.

ss 6.2.2. Dominant Risk Sequences k A review of the rank ordering of the individual sequence risks given in

[

h Table 6.1 shows that the total plant risk due to PTS is dominated by five h

I seguences (2.1, 2.3, 2.4, 8.2, and 8.3). These sequences represent I

t i

L- -- . . _ _ - - - - . _ . , - _ _ _ _ _ . - - , . _ . _ , - - _ .

CC-6.8 '

f~5

• 3 J L%N' t 'l dj approximately 97% cif the total plant risk due to PTS at 32 EFPY as deter- /.y mined by this study. The risk associated with each of the five transients hi l

'.x "

is presented in Table 6.2 and plotted in Figure 6.1 as a function of Eg. ' l lf

/

It is interesting to note that as Eg increases, the relative contribu-l tion to the total risk from the LOCAs which result in loop flow stagnation (as in sequences 8.2 and 8.3) decreases, while the relative contribution due to sus 11 steam-line breaks (as in sequences 2.1, 2.3, and 2.4) increases. In the following paragraphs each sequence is discussed with

~

respect to thermal-hydraulic characteristics, fenuency of occurrence, con-t X ditional failure probability and relative change with increasing g values.

'N .

Sequence 8.2 Sequence 8.2 is basically a small-break LOCA with a loss of natural circa-lation. This stagnation condition can be achieved by several means but.

would appear most frequently to be due to the occurrence of the small-break LOCA at a hot 0% power condition (lotr core decay heati. In Chapter 4.0 it was assumed that tLis scquence would Jead to loop stagnation. Since this assnaption led to a dominant sequence, it was necessary to cotually perform the calculation of the thernal-hydraulic properties for this sequence.

(SeeresultsofTRACcfalculationsinAppendixF.) The IRAC calculation confirmed the previous assumption and loop flow stagnation was predicted to occur.within a few hvadred seconds after event initistion.

The downconer temper'st area calculated by TRAC for sequence 8.2 were some-vhat higher than those calculated by Theophanons and presented in

~-.e.e , - - - -.. . . - . , , - - , -

-w-e '.,me,wwwn- nv ve -

--w*+ -- , - - - - ---- ---- *.----~,w--

- ->m-- - --,c-w---e -----e ~=w

! CC-6.9 m, n r ._pl

- d' L Table 6.2. Summary of risk vs EFPY, F , and RTNDT for dominant risk sequences  ;

EFFT 9.2 16.8 24.4 32 41.2 53 .0 F,, 10 19 n/cm2 1.52 3.03 4.55 6.06 7.88 10.24 RTg 4 2cr, 'C* 79 99 112 122 132 143 Sequence -

Number +-4'*- -= mo , - -

Through-the-Tall Crack Frequency (yr-1)

.m-e. -wee -m*

2.1 7.6E-10 5.3E-9 2.3E-8 2.3 1.0E-11t 1.5E-10 8.5E-10 3.5E-9 1.2E-8 2.4 1.0E-11 1.5E-10 8.5E-10 3.5E-9 1.2E-8 8.2 1.5E-10 3.6E-9 1.8E-8 4.5E-8 8.7E-8 1.8E-7 8.3 1.8E-9 9.5E-9 2.0E-8 3.0E-8 4.0E-8 5.0E-8 Total 2.0E-9 1.3E-8 3.8E-8 8.0E-8 1.4E-7 3.0E-7

• Temperature headings in 'F are 174, 210, 233, 252, 270, and 289, respectively.

t Read: 1.0 x 10-11 O

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, l' '

\

f

='

CC-6.10 p' . . .,

, 1, r,.rt- a 1

I i ,,

'h ff k if b ' h

- t RTNDT

• to ("C) 79 99 112 122 131 143

' b

~ ;

l 1 I I & I

Sequence

No.

., H Total i 8

, 8.2 i

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CP e

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a-  :

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t, i

aL

-bw w

s. - ,

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I A,  ; / .i2, c

j

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I -

1. . O.-,

I o

E' Ii .

I $ -

2.3 + 2.4 6

" i b

~

E

~/

6 W

0.0 10.0 20.0 30.0 40.0 $0.0 60.0 EFPY i

0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 l

Fluence Figure 6.1. Risk associated with five dominant sequences.

if 1

. . . CC -6.11 hg

• ,0. ,

1 Chapter 4.0. However, the TRAC analysts have pointed out that TRAC cannot corrsetly account for the reverse flow and stratification conditions expected when' HPI water flows into a stagnated cold leg. As a result, it was assaned that TRAC would over predict the downconer temperstare, and the l temperature profile provided by Theophanons was taken to be the best esti-mate of temperature conditions for this transient.

The cooldown process for this transient is dominated by the constant inflow of relatively cold HPI water into the stagnated cold loops. The minimum temperature is 125'F and it occurs at the 2-hour analysis time limit. The temperature will continue to slowly drop beyond the 2-hour time period, but an increase in the failure probability at times greater than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is not expected.

4 Sequence 8.3 i

The principal difference between sequences 8.3 and 8.2 is a difference in i

pressure during the latter part of the transient. In sequence 8.3 the LOCA i

event is terminated by isolation of the break. Due to the nature of this event, no credit was taken for controlling the repressurization sad thus the system quickly reaches a high pressure condition. The minimum 1

l l I s

> l 3  ;

i

• CC-t .12 f N* l temperature for this sequence is essentially the same as that for sequence 8.2, but the final pressure is considerably higher than that for sequence 8.2.

i l

The event frequency determined for this sequence is almost two orders of magnitude smaller than the event frequency for sequence 8.2, but the higher  ;

pressure results in a conditional failure probability increase of almost a l f actor of 30 at 32 EFPY.

i Sequence 2.1 l

This sequence is a sus 11 steam-line break at hot 0% pov3r, and it has the l

highest event frequency of the five dominant sequences. However, the severity of the transient is substantially less than that for sequence 8.2 or for sequence 8.3. The minimum temperature for sequence 2.1 is ~250*F, which is to be compared with a minimum temperature of ~125'F for sequence 8.2. Thus, the conditional failure probability is lower. However, as the BINDT value increases, the conditional failure probability increases much more rapidly than it does for sequence 8.2 or sequence 8.3. For situations involving very high EINnT **l***e it 18 Perceived that this sequence could become the dominant transient.

Sequence 2.3 i

This sequence is also a steam-line break at hot 0% power, the principal difference between this sequence and sequence 2.1 being that this sequence

i l

. . CC-6.13 ['3n  !

O .

(2.3) has the additional fallate of the operator not controlling the repressurization. The additional failure reduces the event frequency by about two orders of magnitude; however, the effects of this f ailure produce a much more severe trancient due to the increased repressaritation rate (minimum temperature is the sene as segnance 2.1). This results in an increase in the conditional failure probability of two orders of magnitude over that for sequence 2.1. Thus the integrated risk associated with tran-sients 2.1 and 2.3 are approximately the same.

Sequence 2.4 4

In the analyses performed in Chapters 3.0 and 4.0, sequences 2.3 and 2.4 were treated es identical sequences. The only difference between them is I that sequence 2.4 includes the additional failure of the operator not con-trolling the auxiliary feedwater flow to the steam generator on the intact steam line. This additional failure was determined to have little effect

on the thermal-hydraulic conditions in the downconer region, and, as noted f in Tables 6.1 and 6.2, the PTS risks for the two sequences are the same.

4 6.2.3. Relative Importance of Each Category of Sequences as i

Initiating Events

{

i i

l In the previous section the individual dominant sequences were identified i

l and discussed. In this section resnits are presented for categories of sequences. Eight initiating event categories have been developed in provi-ons sections. These categories are:

a il i

L_ - - - _ _____ . . - . - , . - . , - - -_. .- ._ _ - - - -

CC-6e14

• f: ,

- i . ,. ~~ ?

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i (1) Large main eteam-line break at hot 0% power.

(2) Small main steam-line break at hot 0% power.

(3) Large main steam-line break at in11 power.

(4) Small main steam-line break at in11 power.

(5) Small-break LOCA (<0.016 f2 t ) at in11 power.

t (6) Small-break LOCA (<0.016 ft2 ) at hot 0% power.*

/Y g ,

(7) Small-break LOCA (>0.016 ft2 and <0.05 ft 2) at in11 power.

(8) Steam generator overfeed.

I The risk associated with each of these eight categories is plotted in Fig- [, -

are 6.2, along with that for an additional category (No. 9) that includes 11 residual groups.

6.3. Effects 21 Potential Elsi Reduction Measures i

1 2

The effects of potential mitigating actions were examined as a part of this l study. ' Ibis section is not intended as a list of recommendations but is provided to give information on the relative value of actions which could be taken prcvided a need to reduce the integrated risk due to PTS is iden-i tified.

nte setegory has previeesty been defined as ses11-break 18CAe which lead to leep stagnettee. Simee this estesery ese foemd to be dealasted by esall-break IACAs et het St power, the estegory title was enaaged to better describe the eageesees within the estesery.

' i lmetades mais feedvetor everfeed events (stish are the eely reester trip

segesesse that de met fall tote one of the other event estegories) ples r es.iii.,, fe.d..te, .e, reed e,ent e.

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Figure 6.2. Risk associated with each category of events.

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In the prossarized thermal shock evaluation of the Oconee plant,1,,,,, j reduction measures were examined:

(1) Limitation on primary systen repressurization, (2) Introduction of a high stesa generator trip system, (3) Reduction of neutron fluence-rate, (4) Heating of the HPI water, (5) In-service inspection of vessel, (6) Annesling of vessel, and (7) Improvement of operator training.

Limiting repressurization was not examined in this study for Calvert Cliffs Unit 1 since the low head HPI systen already slows the repressurization, sad the practicality of introducing an automatic restraint on repressuriza-tion is not clear. The other six measures were examined for Calvert Cliffs Unit 1. In addition, one other risk reduction action, that of maintaining RCP operation during a secondary side overcooling transient, was examined.

nose seven corrective actions are discussed below.

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li 6.3.1 Introduction of a High Steam Generator Trip System Calvert Clif fs Unit 1 does not have a system that automatically terminates feedwater flow when a designated high steam generator level is reached.

The principal effect of such a system would be early termination of an overfeed event. In the thermal-hydraulic analysis performed for this study, no credit was taken for termination of feedflow for the overfeed events. Thus feedflow continued until there was insufficient water in the hot well to maintain flow. Under this assumptfor, the maximum overfeed condition is obtained; however, the consequence of this maximum overfeed was negligible. Thus the introduction of a high steam generator trip of feedwater pumps would have no effect on risk reduction.

6.3.2. Reduction of Neutron Fluence Rate The benefits obtained from reducing the neutron fluence rate in the vessel wall by factors of 2, 4, and 8 were evaluated. Since fluence has a cumula-tive impact on the vessel EINDT value, reducing the fluence rate will retard the effective rate of aging. This can have a significant effect on risk reduction. It was found that the fluence rate reduction factors of 2, 4, L. . 8 resulted in risk reduction factors of approximately 3, 11, and 27, respectively, at 32 EFPY.

6.3.3. Heating of the HPI Water l

l In the Ocones analysis it was determined that heating the HPI water would provide only a maall risk reduction since the vent valves ensured that the

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CC-6.18 L- . . .

warm water would always be mixed with colder HPI water before reaching the vessel vall.

For Calvert Cliffs Unit 1 the situation is substantially different. Since the plant does not have vent valves, the dominant risk sequences 8.2 and 8.3 are greatly impacted by the temperature of the RPI water. A 40*F increase in the RPI water temperature would translate to a 30*F warmer downconer temperature at the 2-hour time period. H is 30*F warmer downco-mer temperature decreased the conditional failure probabilities associated with sequences 8.2 and 8.3 by factors of 10 and 2.5, respectively, at 41 '

EFPY (Q + 2a = 270*F) . This resulted in a total risk reduction f.,* tor of 3.8 at 41 EFPY.

6.3.4. In-Service Inspection of Vessel In the Oconee analysis1 it was assumed that in-service inspection would reveal 90% or 99% of the surface flaws with depths equal to or greater than 6 mm. It was farther assumed that all flaws found would be repaired. If before the in-service inspection, no calculated f ailures were attributed to initial flaws with depths less than 6 mm, then the 90% and 99% inspection would reduce the conditional probability of failure, P(F E), by factors of 0.1 and 0.01, respectively, his assumption led to an overall reduction in the probability of vessel failure by about a factor 2 at 32 EFPY. he reduction factor was limited by the fact that the very shallow flaws which would not be detected or repaired actually make a significant contribution to the total probability of vessel failure.

1 __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

1 CC- 6.19 _

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Since the Oconee analysis was performed, many questions have been raised i

concerning the efficiency of flaw detection methodologies

• used and the practicality of repairing flaws. As a result, this explicit analysis was not performed for Calvert Cliffs Unit 1. However, a review of the dominant sequences reveals a distribution of failures with respect to flaw depth

) which is similar to that observed for Oconee. Thus under the same assnap-tions as used in the Oconee analysis, a factor of 2 reduction in vessel failure probability due to identification and repair of flaws wonid not appear to be unreasonable.-

1 4

l 6.3.5. Annealing of the Vessel Annealing of the vessel will restore the fracture toughness of the vessel 3

material, effectively cancelling the effects of neutron finance. 'llie extent of recovery will depend on the chemistry of the vessel material, the time-temperature characteristics of the annealing procedure, and the number of times the vessel is annealed. If it is assumed that full recovery of the vessel is achieved, a reduction of 1 to 2 orders of magnitude of the risk relative to that at 41 EFPY may be possible. However, further annealing won 1d be required on some periodic basis if this measure is to l l prevent regrowth of the risk. It shon1d be noted that the feasibility of i in place vessel annealing was not addressed in sufficient detail by this study to assare the effectiveness and practicality of this measure.

+

{ 6.3.6. Improvement in Operator Training 1

Operator training was not directly addressed as a variable in this study,

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l , ,i L _u_-s bat it was indirectly examined as part of the human factors evaluation of operator actions. In situations requiring relatively rapid response (<10 min), training would be considered to be a dominant inf1sence on the suc-cess or fallare of the action. However, since the large steam generators and low head HPI system at Calvert Cliffs Unit 1 appear to spread out the time available for the operator to perform the important actions with respect to PTS, it does not appear that increased training would greatly affect the integrated risk due to PTS at Calvert Cliffs Unit 1.

However, two items should be pointed out which do not greatly impact the risk at 32 or 41 EFPY but which are associated with training and could have some impact under different conditions (at much higher values of RINDI **

much higher frequencies of low decay heat, etc.):

(1) A good portion of the probability associated with the failure of the operator to control pressure with respect temperature during an overcooling event was attributed to the written procedures.

Very little guidance other than a simple caution was provided to the operators. This does not mean that a series of procedure steps are necessary to address the issue. One or possibly two j well worded procedure steps could reduce potential confusion.

i l

(2) A review of the dominant sequences reveals that almost all of the j

risk is associated with events occurring at low decay heat. In l our review of the training program it did not appear that the l

l special significance of low decay heat was emphasized. This does l

not mean that training should ignore the potential for a PTS l

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event in any operational mode. Dat the special potential of a PTS consequence should be recognized for any event which ocents at a low decay heat condition.

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1 6.3.7. Maintaining RCP Operation Daring Secondary Side Overcooling I Transient It has been mentioned at several times in this report that the staff of Baltimore Gas and Electric is considering a change in criteria for tripping the reactor coolant pumps. The present procedures require tripping the pumps whenever safety injection it actuated. The new procedures would ,

require tripping only two of four pumps upon safety injection actuation, with the tripping of the renaining two pumps in the case of a LOCA or loss-of power event.

i

! 71o principal effect of this procedure will be to ensure forced circulation during all steam-line break and overfeed events. Based on a LANL TRAC i

analysis, this could lead to a downconer temperature that is higher by as l

! much as 100*F for excess steam-line flow events occurring at low decay i

! heat.

Then the value of BIgg7 + 2e is less than 270'F, the risk reduction due to this procedure change would be negligible since the secondary side events l contribute little to the overall risk. However, when BINDT + 2a increases f beyond 270*F the small steam-line break at hot 0% power becomes a larger l

and larger contribution to the total risk. By leaving two pumps in opera-

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magnitude.

6.3.8. Sannary of the Effects of Potential Risk Reduction Nessures Of the seven potential risk reduction nessures discussed in the previous sections, only four were found to actually have a significant potential for actual risk reduction. These four actions were:

(1) finance reduction.

(2) heating of HPI water, (3) vessel annealing, (4) change of pump trip philosophy. ~

The effects of these nessures are graphically presented in Figure 6.3.

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