NUREG-0651, Forwards Cost Benefit Analysis of Recommendation to Install Radiation Monitors Around Main Steam Lines of PWRs for Detection of Ruptured Steam Generator Tube.Review Conclusions to Be Used in Evaluation of NUREG-0651

Forwards Cost Benefit Analysis of Recommendation to Install Radiation Monitors Around Main Steam Lines of PWRs for Detection of Ruptured Steam Generator Tube.Review Conclusions to Be Used in Evaluation of NUREG-0651

ML19351F781
Person / Time
Issue date:	01/07/1981
From:	Lobel R, Vandermolen H Office of Nuclear Reactor Regulation
To:	Pam Baer Office of Nuclear Reactor Regulation
References
NUDOCS 8102200154
Download: ML19351F781 (10)

Text

-

f

'-. i t u a T/. T E s f.UCLE f.f' F.FC U L A TORY CC'..J.11SOOf; A:

'.O TON. D. C.1; bl 5 j

January 7,1981

'0TE TO:

R. Saer FROM:

R. Lobel H. VanderMolen

?J5 JECT:

COST SENEFIT ANALYSIS OF. RADIATION DETECTORS ON MAIN STEA!! LINES TO PROVIDE ADDED DIAGNOSTIC CAPABILITY FOR A P'4R OPERATOR DURING A STEAM GENERATOR TllBE RUPTURE As part of the review of NUREG-0651, " Evaluation of Steam Generator Tube Ructure Events," and as a preliminary test of the prioritization procedure being developed by our branch, we have evaluated a recommendation to put radiation nonitors around the main stean lines of PHRs in order to detect a ruptured steam generator tube. The conclusions of this work will be used in our evaluation of NUREG-0651. The details of the method are given in the Enclosure to this note.

)

p..

Richard Lobel i

1 i

.:,. Q, -f,

Han31d VanderMolen

Enclosure:

Evaluation of NUREG-0651 cc w/ enclosure:

M. Ernst I'

l 11

,/

7 i

~

$f q, N 9

Y 5 I 'ry -

t.

..,y[..

<Qi i

8102/00/59

9 ENCLO::RE INTROCUCTION NUREG-0651, " Evaluation of Steam Generator Tube Ruoture Events" reconmended additional radiation monitoring to help the reactor ocerator to determine tnat a steam generator tube rupture (SGTR) had occurred and to determine which steam generator had the faulted tube.

Af'.er investigating various alternatives, the most effective was determined to be one or nore radiation monitors around each steam line.

In order to determine the benefit of such a system in relation to the impact on the PWR licensees, the following study was done.

Our unit of measurement is the cost of the reconnendation to the licensee oer released curie (Ci) of radiation per reactor year (RY).

It therefore contains the impact on the licensee, the degree of protection afforded the oublic and the probability of the event frca which the public would be protected.

As a standard against which to measure, we chose the second train of ECCS which v.e calculated to have an impact of $1000/Ci/RY. The determination of this is given in the Appendix.

While the assumptions are very rouch, we feel they are reasonable and the results of the analysis provide a good basis for making our conclusion which is given in the next section.

CONCLUSION Ue conclude that the benefit to be gained by the installation of this system is not justified by the impact on the licensee.

ANALYSIS We assuce the rupture of the tubes in a steam generator such that the initial break flow is the most probable flow based on actual experience from the three actual cases. These flow rates are given below:

Point Beach 80 pgm Surry 125 gpm Prairie Island 390 gpm Average flow = 198 gpm s 200 gpm This is the initial break flow.

It will become less as the reactor is l

depressurized.

Ue assume the timing of an SGTR accident is based on the Cesign Basis Accident assumptions; i.e., it will take five m,inutes to trip the reactor,10 minutes to detennine that a SGTR has ocrurred and 15 minutes to isolate the faulted steam

~

genera to r.

The total time is therefore 30 minutes from initiation of the SGTR o stean generator isolatfor..

3 1

4 n

---.,,a e.

,._,,.,,,_m,,.

-m,,,.,.

-,,._,,e

,w

J I

If tne operator had this instrucentation, the time to trip the reactor would be tne same. Her.ever, the operator should be able to determine, within the trip ti.me, that he had a steam generator tube rupture accident. Therefore, no i

additional time is required for this determination. Once the reactor is tripped

• nd tne reactor coolant temoerature begins to decrease, it requires only a short tine to arrive at the no load temperature.

The coerator can then isolate the 3ulted steam generator. He requires no significant time to determine which

eam generator is faulted since he would cet a direct indicatisq from che new equ i pren t ~.

'4e allocated five minutes af ter trip to isolate the irs ~ ted steam generator.

These values are ccrpared in the Table below.

4 DBA STEAM GENERATOR TUBE RUPTURE Time Allocated for Tic.e with Reccemended CBA (min)

Instrumentation (nin)

Trip 5

5 Determining a SGTR 10 0

Isolate Faulted SG 15 5

Total 30 10 If we assume that the release would be taking place during the total time until steam generator isolation, the difference in time of,, lease would be 20 minutes.

The break flow is not constant dering this time since the RCS pressure is decreasing. Ne assumed a break. ,1 rate as shown below.

200 - ' '.

i Break Flow Rate i

1 2

3 i

(gpm) i t

0

's<

Timetmin) 5 10 30 For the DBA case, the flow would be the Areas 1, 2 and 3.

• !ith the recommended instrumentation, we assume the flow is Areas 1 and 2.

Area 1 = (200 cal)(5 min) = 1000 gal I

m1n l

l l

Area 2 = 1/2(200 cal)(5 min) = 500 gal min _

Area 3 = 1/2(200)(25) - Area 2 = 2000 gal I

i

._,~,...._y,,,,,,-.-...,-,,,._,,_,m_.-.,...

-.,,,,..,__..._.-_,.,,,-___.-_.---.m,

..,m.-.-,,--,-..---

The difference in break ficw is then a = 3500 gal - 1500 gal

= 2000 gal Exxon has done an an? lysis of the release rates from a SGTR accident for several different cases. These are:

Case 1 - Coolcr. activity equivalent to 0.12% failed fuel prior to accident.

No failed fuel during the accident.

Case 2 - Coolant activity = luCi/gm (Technical Specification limit)

Case 3 - Coolant activity = 60aci/gm (Iodine Spike Technical Specification '_imit)

Table 1 gives the results of their analy:is in terms of curies released / minute for six different scenarios. Only two of these scenarios are significant for our consideration:

Scenarios 1 and 4 Scenario 1 would apply to the case of i

offsite power available and is a release to the condenser (and condenser air ejector).

Scenario 4 is a release through the ADVs and corresponds to the case of loss of offsite power.

The break flow rate assumed for these analyse 5 was 1000 gpm.

We can convert our calculated total flow into an equivalent time to be used to calculate curies released from this table.

t'9 = Equivalent time =

Total flow throuoh break (inminutes)

(in callons)

Flow rate used to calculate release rate (1000 gallons / minute) 2000 callons 1000 gallons ninute 2 minutes

=

e can now calculate the curies / reactor year.

Data show that the frequency of losing offsite power is 10-3 per reactor year.

There have been three eteam generator tube rupture events in aporoximately 300 reactor years (Pt!R) of oceration giving a frequency of 10-2 per reactor year. Then, from the scenarios of Table 1, for Case 1,

'Akers, D. N., et al, " Assessment of_ Fission Product Monitors for FMR Atmos;heric Steam Dump Ef fluents," Exxon Nuclear Ccrpany Idaho, Inc.,'NUREG/CR-iE64, ENIC3-1 1952

-. - _, ~_

J 5

TM 1

TOTAL RADIONUC :DE RELEASE ?ATES U TUBE STEAM GENERATOR

,:2j Fai;ed F:lel 4

Release Ratcs Ci/?.in

• b;c l i de s Case?'c.= er I

8 1

2 i

3 5

5 4

.I i

i i

.i i

~

) Halcgens 2.2(-5)

! 3.7(-3) 2.2(-2) 3..(-3) 3.7(-2) 1.9(-3)

Neule Gas 2.9(*l) 3.6(+1) 7.2(+1) 7.2(0) 7.2(+1) 3.2(-1)

I I

Cs. R' NR

! 9.3(-5) 5.9( 4)

9. i(-5 )

9.3( 4)

! 4.9(-5) 0*.cer t

i Nuclices NR

! 7.9(-5) 4.7( 4) 7.9(-5) 7.9( 4)

! 4.0(-5)

?.R-Not Rel eased J

1.0 tCi/9n OEI Release Rates Ciir.in Nuclides Case 'n.rter i

i 1

2 3

4 5

5 i

Halegens 5.9(-5) 1.0(-2) 5.9(-2) 9.9(-2) 9.9(-2)

.7(-3)

.;coie 1

Gas 7.2[+1)

! 9.0(-1) 1.S(+2)

1. f+1) 1.S(+2)

! 3.1(+1)

I I

Cs. Rb.

NR

2. 5 ( -4 )

1.5(-3) 2.5( 4) 2.5(-3) 1.2( 4)

Otner 1 Nuclides NR

! 2.0( 4) 1.1(-3) 2.0(-4) 2.0(-3)

!- 1.0(-4)

';R-Not Releasec 60 uCi/gm CEI Release Rates Ci/ min

! Nuclides Case.'w.ter 1

2 3

4 1

5 6

1

! Ha gens 3.7(-3)

! 5.1( -1) 3.7(0) 5.!(-1) 6.l(0)

! 2.9(-1) i Nccle l

1 Gas 4.4 ( +3 )-

! 5.5(+3) 1.1( +4 )

1.1(+3) 1.l('4)

! 4.9(*3) i I Cs. Rb.

! NR

2.5(-4) 1.5(-3) 2.5( 4) 2.5(-3) 1.2( 4)

I J:.ne r i Nuclides NR 2.0( 4) 1.1(-3) 2.0( 4) 2.0(-3) 1.C( 4)

NR *ict Releasec i

r D

S

-Q-D oj o

7 ao

i

Cji,

-2 SGTR [2 x 29 Ci (1-10-3) + 2 x 7.2 Ci (10-3)), g,gga,

_ jg 3Y RY SGTR SGTR For Case 2 (luCi/gm) we must include the probability of reaching this value of activity in the primary coolant. A rough estimate is that between 5 to 10%

of all P' irs will reach the limit every year. Me will use the higher probability of 0.1 Der-reactor year.

h=(10-2S

)(0.1) [2 x 144 SGfR(1-10 ) + 2 x 20 (10 )ShfR3

= 0.237 For Case 3 we need the probability of reaching the short tern Technical Scecifi-cation limit of 60uCi/gn. '4e assume that a spike of 60 times the steady activity occurs 5" of the reactor year. But to achieve the limit, we would have to already be at the steady activity limit of luCi/gm.

In addition, most spikes occur due to changes of power or pressure in the downward direction (associated with reactor trips, for example). We assume only 20% of the spikes occur with a power increase so that the reactor would still be at 100% power i

at the beginning of the SGTR accident.

Therefore, for Case 3:

1 h=(10-2)(0.1)(5 x 10-2)(.2) x

[2 x 4400 (1-10-3) + 2 x 1100 (10-3)3 4

.088 We assume the cost of each detector (purchase, design of supporting system at i

the plant and installation) is $50000.

Assume a factor of 3 for safety grade.

Therefore, each detector is $150000.

Assume two detectors / loop (a 1 out of 2 loaic)

Therefore, cost per 1000 is $300000.

Table 2 gives the results in terms of our proposed units of $/Ci/RY.

i l

l

. TABLE 2 COST OF STEA'i LI!1E RADIATIOi M0!i! TORS Ci/RY S/Ci/RY /1000

=

.622 a.82 x 10' These numbers are to be compared with $1000/Ci/RY for a second train of ECCS.

The system is therefore not providing an efficient reduction in curies released for safety dollars expended.

AP P E *.., : X CALCULATION OF STANCARD OF COMPARISON FOR UNITS UF $/Ci/RY As a standard against which to measure, we chose the second train of ECCS.

To calculate the cost-effectiveness of this train in a rigorous manner, it would be necessary to re-calculate the accident sequence prcbabilities in WASH-1400 using an appropriately higher value for the ECCS failure probability. Hcwever, tnis is far too extensive a calculation to be prav.ical for our purposes.

Instead, we need Tables VI 2-1, VI 3-1, and V 3-14 of WASH-1400 to form a rough estimate.

The nine PWR accident sequences have probabilities and consequences as given in the attached Table.

Categories 1 through 7 involve core melts.

In Categories 8 and 9, the ECCS works and the core does not melt.

The probability of a LOCA ("A" and "Sl" in WASH-1400 language) is 4 x 10j per reactor year. Thus, the ECCS can be considered to prevent almost 4 x 10 core melts ge* reactor year. However, the curies released by a core melt can vary widely. Me can calculate an " average" core melt release as follows:

6 Average release = { 1 (Probability of Category n)(release of Category n) n=

e (Probability of Category n) n=1 8

= 1.9 x 10 Ci The removal of one ECCS train would increase the probabilities of most of the individual accident sequences in each of these categories.

In this rough approximation, we assume that all seven core melt categories will be precor-tionally increased.

5 Examination of Table V 3-14 implies that the ECCS failure rate in larca LOCAs is on the order of 3 x 10-2 per event. Therefore, the success rate is:

1 - 3 x 10-2, 97; If one ECCS train is removed, the success rate becomes 1 - (3 x 10-2)1/2

= 82.7%

Therefore, the second ECCS train prevents 97% - 82.7% = 14.3% of the 4 x 10-4 LCCAs per reactor year.

This means that the second train prevents.

(.143)[4 x 10-4)](1.9 x 108 4

Ci) = 1.09 x 10 Ci/RY frcn beina released, as a rough approximation.

1 7

2-The cat. of an ECCS train is approximately 10 million dollars.

It is tnerefore straight forward 'a calculate a cost / benefit ratio of 7

10

=90 3

1.09 x 104 Ci/RY 0

Ci/RY Ci/RY ww

=

8 e

.-mn

_.ew-g'

. _.. ~..

- - - ~ _. - _. _ _. _

1 r

i '

i l

TABLE i

Release Category:

1 2

3 4

5 6

7 8

9 4

Probability,(RY)'I

.9 x 10 8 x 10-6 4 x 10-6 5 x 10-7 7 x 10-7 6 x 10-6 4 x 10-5 4 x 10-5 4 x 10-4

-7 8

-5 Curies released (x 10 )

11.52 9.27 5.18 2.82 1.27 1.04 2.08 x 10-2 8 x 10-3 10 J

l t

?

i

!i r

4 l

r I

t I

1 f

A i

4 L

i i

9 l

f s

g e

1