ML20049H977

Technical Report on D.C. Power Supplies in Nuclear Power Plants

ML20049H977
Person / Time
Issue date:	07/31/1977
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0305, NUREG-305, PB-269-338, NUDOCS 8203110080
Download: ML20049H977 (47)

Text

-

' e4 U-- --

D F

U.S. DEPARTMDIT OF COMMERCE National Technicallaformation Semce PB-269 338 Tecanica Report on 3.C.

Power Supa ies in Nuclear

%wer Plants l

1 Hudcar Regu!ctory Commission, Washington, O C Jul 77 c

3 0 770731 0305 R pop

-w-

.,_.c-,

jcar--w,,.n

=

7

[j.,

PB 269 338 l

h.

" ~.

NUREG-0305 fj

.1

!I

'l

<I t

T1 3

f.I l'

\\'

(

4 TECHNICAL REPORT

e

.i 4

i.

l ON i<l i

l r!

D.C. POWER SUPPLIES IN NUCLEAR POWER PLANTS 1

l 1

l A

i Office of Nuclear Reactor Regulation l

U. S. Nuclear Regulatory Commission etP9CCUCED BY NATIONAL TECHNICAL i

INFORMATION SERVICE

(

" ::'s%Vni;"'

i

- y,qj

~

~~

~ r~.

- ~ ~ _

J'

, \\

%I jjp j - ;- --

U!$;i

~4tt f s

.e l'"$j, i -1 s

.).,

. ' T.=

M -

h!

l

?. 15

~,*f i

a t

f

') -s 4

+ 3 'i

'I s

ld l

t.

2 sf-6

?. 1

! -i Available from

[-

National Technical Information Service y

Springfield. Virginia 22161 Printed Copy 54.00; Microffche $3.00 Price:

a The price of this dccument for requesters cutside k'.

of the North Anerican Contincnt can be cbtained E-from the National Technical Info w tion Service.

i 4

6 1

P 2

l -

f:

H.

1 i

i I.

I

,l

\\

-.gp

- - -. ~ ~ -

y-StBLIOGR APHIC D ATA

1. M cPort No.

2.

3. hec aree nt** Ac ce**ien.No.

1 5HEET Nt' REC-0305

s. keg ori pue

4. In 1. and s,beeile July 1977 Techn.ical Heport on D.C. Power supplies in Nuclear Power Plants t

....e r,.e m.n, or,. n.,....n e p..

,.A...r,,,

N..

io....,ee,,,o..

ca., No.

.. i.e r,.e-,n, or,.n....on s. me.nu u,e.,

'p Of fice of Nuclear Heactor Hegulation gi,,,,,,,,,,g,,,,,3,,.

Division of Operating Heactors/ Division of Systems Safety l'.S. Nuclear Hegulatory Cocunission Washir.gton, D.C. 20555

13. T yp t kepers & l'er. a

12. Sponsorens Or,ans,ation Name and AJJr. s t overeJ eg

>I Same as 9 Above 14

15. Supplementary Notes

16. A5teracts a

This report addressee. the reliability of the d.c. power supplies (batteries]

at operating nuclear power stations. It includes discussion of the NHC staf f's assessment of the likelihood and of the consequences of' a postulated failure of both d.c. power sup;ilies during normal operation of t. nuelcar power plant. It also discusses the NHC's plans for continuing review in this area.

5. A * ) E orJe 4*ta Doewerent Analysis, lle.llescrqtore I

\\

l 17b. IJentifier. 'Or en-I nJcJ Terms P

f ITs. C04A TI t ie t.t Geo.sr H.

w.e. sin. Inn

21. No..,a Pa n e.

3. A va.44N.ty w ate ment r

k a pee t J/

'* _ t !f f f !*

IU J

'e.

"* * * * *

• i 8 ' n 2 2. s'r.cc No Hestrictions on Distribution

[$

.N

[v t s uf f-,9 f)

Y...'....

• a==~'***

= * '. " i v..i.,* o i s.. s o, i s,.

r ii r,,,,.n,,,.. i.,

~...

f d

NUREG 0305 t

t I

t f

TECHNICAL REPORT ON

}..

D.C. POWER SUPPLIES IN NUCLEAR POWER PLANTS s

f I

i f

i

(

l i

i i

i t-i Manuv:ript Completed: June 1977 Date Published: July 1977 i*

Division of Operatmg Reactors Division of Systems Safety Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commmion Washington, D. C. 20555 l

hl I

. _ = --

9

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

i TABLE OF CONTENTS i

(

I.

INTRODUCT*0N AND SUM 1ARY

11. TECHNICAL CVERVIEW f[.

!!I. DISCUSSION OF TECHNICAL ISSUE I

IV. LIKELIHOOD OF LOSS OF ALL D.C. POWER si'r V.

CONCLUSIONS a$

I ATTACHMENT A - Wasoington Star Article of June 15, 1977 ATTACHMENT B - ACPS Consultant's Report 3

))

APPENDIX A o

r. ;

p.

t I'

d

• e

+

t r

i k

1' i

i l'

i e

7 I

l r

't, i.

l 4

til I

l 1

1 e

.=---.-*ew-,e-

-.m

---r.

--m w--

arc w

-m-

---

*,,..-re

,-g c,

3 n

.l-D.C. POWER SUPPt.!ES IN NUCLEAR POWER PLANTS l

5 1.

INTRODUCTION AND

SUMMARY

i Washington St_arE carried an article referring to The June 15, 1977 a

the flRC's investigation of a " safety flaw" in nuclear power plants.

The article is based on a reportU (dated April 12,1977) prepared by Mr. Epi,er, a consultant to the ACRS.

The issue discussed deals witn the reliability of the d.c. power 2

supplies at nuclear power stations. It is postulated that a sudden gross failure of the redundant d.c. power supplies cccurs during normal operation, and that this could lead to insufficient shutdown cooling of the reactor core. Mr. Ep

.tated that this potential for insufficient cooling is great enough to warrant consideration of prorrpt remedies.

The flRC staf f has had this tratter under review in conjunction with its reviews of generic safety issues, since receipt of Mr. Epler's April 12, 1977 letter. The stsff's views on this ratter are that:

1) Reliability of the d.c. power supply system is ar. important

+:

safety issue;

2) A quantitative reliability assessment of the d.c. p wer 2

supplies should be performed and any needed change in procedure or design implemented; and M See Attachrent A E See Attachrent B

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

1

) 1 1

3) Based on the staff's judgment of the low likelihood of j

occurrence of the postulated multiple failures, operating l

plants may continue to operate and new licenses may centinue 4

to be issued pending completion of the above mentioned quantitative reliability asses ment.

The specific scenario postulated by Mr. Epler is a. follows:

'4hile a nuclear power plant is cperating, one of two f

redundant d.c. power supply systems fails causing a reactor scram and subsequently cacsing loss of all offsite power. At this point, safe shutdown of the plant requires that the residual heat from the decay of radioactivity be removed from the reactor. Control of valve position and pumps needed to rernve residual heat af ter plant shutdown depends on availability of d.c. power supply, all remaining sources of d.c. power may be lost; and then, according to Mr. Eple r, continued ccoling of the reacter core cannot be assured.

The NRC staff's view is that the simultaneous and independent of redundant d.c. power supplies is so unlikely as to be failure l

incredibit and that their failure from a ccm:.' n event is judged to be low enough in likelihood thit adequate protection of the public health presently exists, but that additional technical l

i

{

e--

-m

7

_...._ studies over the next year should.ind will be performed to add y

confidence to this judgment. This view stems from the follcwing:

(1) the postulated scenario is highly unlikely; (2) the period of 5

vulnerability to the above cited single failure of the redundant d.c. power supply is limited, i.e., both the d.c. power supply failure initiating the scenario, and the second failure of the remaining source of d.c. power must occur within 30 seconds to

\\

defeat starting of the redundant 'ilesel and to acceptance of critical loads; and (3) the degree of vulnerability is mitigated L

substantially by the availability of alternethe measure for restoration of power or for removal of decay heat and of suf ficicret tire (at least an hour) for operaar implementation f

i of these alternative s.

Accordingly, while the NRC staff agrees that the issue is important and warrants a quantitative assessment of relishilities for the d.c. power supplies particularly with respect to cormn node failures, the staff does not agree that a nodification of the i

N presently assigned priority and schedule is warranted or that special licensing action on individual olants is warranted.

(

This report outlines the specific scenario pcstulated by Mr. Epler, provides technical background on desigr. of the d.c. power supply l

systems, and pro,ide', the.iRC staf f's view on the costulated I

accident scenario. It further delineates the basis for the staff's view on the likelihood of these postulated events and the safety siqqificance of such failures.

l i

l l

l l

l t

1 3

. : ~ --

7pm y sum 4

6

II. TECHNICAL OVERVIEW f

The d.c. power systems in a nuclear power plant are required to 6

be designed to engineered safety feature quality standards since l

they are necessary to provide control and motive pcwer to system',

in tt.e event of accidents or abnomal plant operations. Engineered f

safety features for nuclear power plants are designed and built to the most stringent criteria and requirementsN of any of the systems in the nuclear plant.

To assure a high reliability of these engineered safety features the regulations require that a single e

failure of equipment be tolerated with ut loss of function. This is in spite of the numerous controls, alarms, and surveillance tests which are leposed.

The NRC staf f's single failure criterion requires, as a minimu:'s, that electric power for on engineered safety feature be comprised of two redundant and independent divisions, eacn capable of providing the necessary plant protection functions durinn all nomal operating conditions and following various design basis accidents. The systems which remove decay heat from the reactor are designed to maet these r?quirements.

O ngineered safety features are required to meet all requirements Eof the General Design Criteria (Appendix A to 10 CFR Part 50) such as redundancy, protection from cuthquakes, floods, and tornadoes and the rigorous quality assurance requirements of l

Appendix 8 to 10 CFR Part 50.

l i

5-j Insofar as electric power requirements for safety systems are

)

concerned, each division includes an of fsite a.c. power connection, a standby emergency diesel generator a.c. power supply, and d.c.

l, power sources (a battery and battery charger). The battery charger 5-is powered by the independent offsite and onsite emergency a.c.

C power supplies and, by itself, can supply sufficient d.c. power i

without batteries to accomplish the shutdown cooling function (usually j

a spare [ third] charger is provided for use in the event of a charger failure).

II k

h Each d.. power supply normally provides control power within its 1

[

division of safety equipment for starting the diesel oenerator, for operating electrical circuit breakers, and for control of logic circuits associated with ele-trical safety equipment. Each battery also provides a.c. power t... x

c. to a.c. inverters to d

the safety instrument buses.

..me plant t..

'so use these d.c.

power supplies to control the switchyard circuit breabers. Fiqure 1 contains a simplified block diagram showing the various a.c. and d.C.

I y

components for a typical nuclear plant.

This simplified block diagram (Fiquee 1) and its associated line diagram (Figure 2) illustrate the relationship between the a.c.

and d.c. power systems of a typical two division safety system.

Given this typical safety system configuration, Mr. Epler's concerns stem from the following scenario:

y

! g

1) For rest plants, failure of one battery causes a reactor y

scram and sub',equent trip of the plant generator, prevents the starting of one diesel generator, and disables one train of the redundant systems to remove resid' sal heat;

2) It is then postulated that af ter the reactor scrams, the facility will lose offsite power as a consequence of the generator trip or due to some other independent cause; 3

s j

3) failure of the second d.c. power supply is then postulated l

to occur within the first thirty (30) seconds. (Ifthe second battery is available for the 30 seconds imediately folicwing loss of offsite power, then the associated diesel generator is designed to autonatically start, power the battery charger, and provide sufficient d.c. r)ower forresidualheatrenoval.);and

4) Given the postulated scenario outlined in items 1-2-3, the plar.t could be deprived of all imediate residual heat removal capability (the loss of offsite power means that the plant's norr.a1 non-safety cooling systems tre uravailable).

In assessing the p.t'.il.ited sequence two aspects are of critical importance: (1) tne likelibcod tha' this could happen at one of today's operating nuclear plants and (2) the consequence of this sequence. Although thE postulated scenario is extrenciy

- =

= -

=

2_-

l.

i

(

.7 t,

i I

unlikely, we believe that the frequency of singic battery degradation 3

experience is such as to warrant re-examination of the reliability t

of battery systems over the next year. In regard to the consequences.

even given the loss of imediate residual heat removal capability.

R nuclear plaats have sufficient tine (greater than one hour) within 1

f F

which to restore power or to manually align systems to add coolant to p

{;

the reactor to ensure adequete core cooling. The bases for this 1b

[

judgement are provided in Section III and IV of this report.

f,'

i

[

E t,

k i

1 I

E f

?

k i,

0 EO e

9 6

0

~

m.~

x.

t FIGURE 1

~

l TYPICAL TWO DIVISION SAFETY SYSTF,M i

i TRANSMISSIO'J LINES

?

r--------*

SWITCHYARD ---------g I

l l

DISCONNECT ^^ MAIN l

LINK l

\\

STEP-UP l

XFMR l

UNIT MW START UPMW l

l AUX

/

MAIN XFMR Ne y g

t N AN#XFMR GEN l

l 1

o y

o NON St.FETY DIESEL NON SAFETY l l

LOADS l g l

LOADS GENERATORS 7

~~

~~~

~

o I

u l

SAFETY

• j SAFETY I*

LOAD >

LOADS l

l

+

i l

l BATTERY BATTERY CHARGER CHARGER I

I I

o g

l u

k+

BATTEftY BATTEHY M

SAFETY DIVISION A SAFETY DIVISION B

./

NOTE: EITHER DIVISION CAN ALONE PROVIDE ALL SAFET/ FUNCTIONS

f--

3,,.,. :._,,._________________ m i,

i.

l

• */.,*.

i i

-m.

i.

l.

i l

i n

e

/

i r

i 1

i a

i f

i e

i l

l e,

t-9 g

.~_- 7 i

r e

l l:

l f...f...........,..._ _..(s.........f i.

i

3.. >. _ _ _ -.3.. _) _..,. _l ! _._ _ _I i

i I

I

)...) _ _ _ _'>...)

I I:

l I

i i

i l

l ll 1

l l.:::l u.l l

c.

i li l

m W

p i,

-- 8 I

A.i.'.T.A..

l

.A.i.'.7.A..

i i

t...,,

,, -. -s i,

(

ll

(

i r

i

.. 3... _ _ _ _p...)

3

>.. >....).___3...)_ -.

l ~.m..l l.l

[....,. j g.,.

Im n.

i ic i

i,

~,

I

==

::.==

i t

i i

1 f

i.

_, i

,! l,:

e i t

t 1l.

i

i i

e i is i

i,

d ll

)

) >

i:. i!

) >

i e

1 l

I 1

I i

,i l

l i

iI I.et.il I.=.1 ll li s 1.=1 1 rt..I e i ii i,

i I

l

{T.q 34,4 is ii w4 g ~w g

~

~. ~

i i;

.f a,....

FTl;i5-, i i

s ii i l 1 =l

!! l:

i l!

/

i

!i{,

5 i

ll 5'

55 e i l

i i ii nl. ~

.,. l l *,

• • • l l

l

..Tl.

Q

~

n...

l l

l l.-.._ _"-_4.:r...I p:.

i.l, L=.,I l..:4----!!,

A.]

i m

i.______________

,--t. ele i.____.__________;BD.

s 1.oma. Pr. T b.e.a $.e, &,.m.. g.rwaal L.e D.ny.=

i 1

8-f, III. DISCUSS 10ft_OF TECHNICAL ISSUE _

a Briefly stated, this issue is concerned with the circumstances i

i f

in plants (particularly PWRs), where failure of one of the two D r.,

power d.c. power supplies could cause a reactor scram.

f, supplies could cause a ructor scram and the subsequent loss of e

For such a postulated situation, nomal shutdown f

offsite power.

g

(,f the plant relies on availability of the auxiliary feedwater system in i%s, Jr the Reactor Core Isolation Cooling System in If it is further postulated BWPs for removal of decay heat.

th=t all a.c. power and the remaining d.c. rower supply fail, then removal of decay heat by comal methods riay not be possible.

I The complete loss of the d.c. power supply in a nuclear power

?

i plant for an extended period of time is a serious event.

Ho ever, when one ca.isiders the various pathways that such an event might take following this occurrence, one concludes that the cooling riechanisms and alternatives available to the reactor l

plant operators are sufficient to safeiy maintain cooling of the core until adequate d.c. power con be restored.

for the various scenarios involving loss of all d.c. poder, there will remain a sufficient amount of water inventory in steam l

ger:erators for PWRs and in the i? actor coolant systems for BWRs and pWRs to assure that the core will remain cooled for periods of tirne in excess of one hour. Appendix A describes i

k

the various systems effects for both BWRs and PWRs and sets i

forth the bases for these periods of time. It must be recognized that these calculations are neant to be representative only, and would vary between specific plants.

Ii if a reactor were to scram (as is expected for some plants) as a e

t' result of failure of one of the two d.c. power supplies, the H

remaining d.c. power supply wotld be expected to furnish the

{

necessary d.c. power. Ordinarily, offsite a.c. power would alsa

(

remain available even following a reactor scram. In this event, L

f<

the battery chargers would rectify a.c. from offsite power to supply c,

f power to the requisite d.c. loads.

L If the reactor scram were to result in a loss of offsite power.

L which is unlikely, and if the second d.c. 7.wr supply were to likewise fail prior to startup of its associated onsite diesel generator.

the operator would have to implement alternative measdre$ (depending on plant design) within a period of about one hour. Hwever, i' the l

onsite diesel generator associated with the second battery were to start and take on load as designed, f.c., within less than 30 reconds, then rectified a.c. power would be available through the battery charger to power the requisite d.c. loads. Thus the second d.c.

power supply would have to fail wii.hin 30 seconds of the loss of o

f tf' power event to realize a total loss of d.c. power. The probability of this event is exceedingly small.

1 1

I' I

~

t If of fsite power reclained available during the postulated battery failures, with consequent reactor scram ano turbine trip, the battery chargers in either division of the d.c. power supply would be avall-f able for furnishing the d.c. power (rectified a.c. power) be required l

for safe plant shutdown. Also, the plant's nonnal (non-safety) heat I

removal systems which are driven frce offsite power would still be available. The probability for loss of offsite power as a result of turbine trip is therefore an important consideration in esaluating the postulated scenario. The probo3111ty for this occurrence given 19 the Reacter Safety Study. WAS!!-1400, is one in one thousand. The a

sir.ultaneous loss of both d.c. p3wer supplies and offsite power is therefore an even more unlikely ever.t.

It is also very unlikely for r

offsite power to remain unavailable for periods in excess of one hour.

In sununary, experience during many events indicates that alternative methods are available to safely cool the reactor core even durir.g oagraded events. For example, breakers can be manually engaged by the opt. rating crew and valves can be manually opened and closed as i

re quired. As stated, suf ficient tirie is available to permit the reliance on operator action. When time is available, such as during the postulated sequences of events, the operator has access to a

.unber of alternative measures to safely cool the core, e.g., through the use of steam driven equipment.

To complete a discussion of postult.ted scanarios that could follow l

the loss of 4 d.c. battery, the coupling of the aforemeationed I

l l

a u

11 scenarios with various design basis accidents should be mentioned.

However, frequency of occurrence of a design basis accident, i

t occurring coincidently with a loss of all d.c. power, is much j

less than once in ten million per reactor year. As such, we l

believc that such hypothetical scenarios are not credible and i

l l

therefore need not be discussed further.

I 1,

b n

i d

i L

I 1

.I; t,

s

?

A e

d 6i i

f-i I

i t

l l

l I

I l

.t i l l

l IV.

LIKELIHOOD OF TOSS OF Att D.C. POWER j

The design of d.c. power systers in nuclear power plants consists of at least two separate power divisions. Each division has the i

capability tu satisfy the safety requirerents of the plant.

D.C.

systems for nuclear power plants, and in particular the batteries, are designed, installed, operated and maintained in accordance with the highest industry standards. During the life of a i

i nuclear power plant, batteries are under continual surveillance j

by the plant operators. Abnormal battery conditions are alanned" in the plant control room such that plaat operators are alerted before either of the d.c. battery rupplies is degraded, in addition, the specific gravity, temperature, voltage, and electrolyte level of the batterin' pilot cells are checked every week; each

ell of every battery is checked quarterly for level of electrolyte, specific gravity and cell voltage. Whenever any of these parameters is not within stringent limits as spacified in plant Technical Specifications,* the batteries at ? ide_ntified as being inoperable and must be recharged or replaced promptly or the plant is shutdown.

For exatple, the Standard Technical Specifications for batteries allow an ir. operable per!cd of t'so hour:, af ter which the nuclear power plant must be shutdown; and if the situation is not re, within six nours, the plant must be cooled down and depressurizeo.

• Technical Specifications are issued by the flRC as part of the Operating License for a plant. Operation of a plant in violation of the Technical Specifications is a violation of the license and could result in enforcement actiens by t.he l4RC.

l l

[~

1

< j

(

4 I

Periodically during the life of a plant, the station batteries are subjected tr, additional tests to monitor their capabilities.

A battery serv'ce test is performed at each refueling shutdown, and a battery discharge test is performed every five years. The 1

purpose of tiese tests is to confirm that the specified capacity i

and the capaJility of the batteries to provide power to engineered I

safety feature systems has not been degraded. These measures assure the aperability of the d.c. pnwer systems.

?

I In addition to the above surveillance requirements, many stringent I

l design considerations go into the wstallation cf the d.c. systea-and the station batteries. Batte.ry rooms and the batteries

(

themselves are designed not to fail during a major earthquake

)

(selsr.iic Category I design). In accordance with the Institute t

of Electrical and Electronic Engineers, Standard 308, the d.c.

System is required to be protected from "all natural phenomana" such as flooding, tornadoes, and hurricanes. In addition to failures that may be caused by natural phenotrena, this standard I

also requires that the d.c. system. including the batteries, be protected I

from several ottier postulated phenomena. Among the nortnally po tulated phenomena are whipping of broken pipes, m'5siles from srechinical failures in rotating machinery, and any single event which can cause r.;ultiple equipment malfunction. Further, the d.c.

power system is designed to retain its redundant functional capability with the complete loss of offsite power.

6 l

I l

.i

{ i Among the objectives of the staff's review of license applications is one of ensuring that the design of the d.c. system is not susceptible to comon failure modes. As a result, it is the staff's view that the d.c. batteries have been drsigned against significant comon mode failures.

The staff is also considering acts of sabotage and fires such as at Browns Ferry as potential comon failure nodes of batteries and other important safety equirment. In both thest. areas the l

staff is taking appropriate 3ction to reduce the likelihood of t

While I

these sources of comon mode failures to 7:ceptable levels.

l one can never conclusively prove that all such contributors to comon mode failure have been identified, the staff has been and will continue to expend considerable effort, analyzing operating experience and reviewing plant design considerations to increase our confidence in this rea.

Although the nuclear plant's d.c. power system is not required to be designed to perform its safety function following postulated accidents such as a loss-of-Coolant Accident when coupled with the sirruitaneous failure of all redundant and independent batteries, Since the d.c. power the battery chargers can provide d.c. power.

requirements for the normal safe shutdown scenario discussed herein are suh tantially less than those for postolated accidents, the a.c. powered battery charger provided in rest plant designs would

{

1 t ',

t 4

provide sufficient d.c. power to effect a safe plant shutdown, if I

either onsite,- t,ffsite a.c. power is available..

Pcwer retc. tors have experienced, on occasion the loss of a The battery and the loss of a single a.c. power subsystem.

1 -

staff is not aware of any case where an operating plant has experienced the complete loss of redundant d.c. power systems j

or the simultaneous less of redundant batteeies.

From the amount of operating experience that has been accumulated f

to date, we have concluced that the predominant s.auses for loss i

of d.c. power are the result of operator err.,r and incomplete i

indication of the statu of the d.c. system to the control coom operator.

In the flRC's Reactor Safety Study, WASH-1400 (fiUREG 75/014), the likelihood that a reactor facility would be without sufficient electrical power for equipment needed to safely shutdown the plant (even assumir.g the more demanding condition following a loss-of-coolant accident) is estimated to be between once in a hundred thousand to once in a million per year of plant operation.

These likelihoods already appropriately include consideration of 1 postulated loss of all d.c. oower (both battery subsystems).

We have no basis to modify these estimates at this time. These probability estimates do rot consider the fact that significant

j time (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or greater) is available to restore pcwer or provide alternate shutdown cooling and thus terminate the event.

I l

A probabi.istic analysis similar to that ustJ in the Reactor Safety Study (WASH-1400) has been performed to estimate the probability of not providing adequate shutdown cooling in a typical P"' caused

)

by unavailability of the d.c. power systec:s Ba.ed on this analysis, the battery failure sequences contribute much less than one percent of all the accident secuences shown in WASH-1400 tnat resulted in significant radiological consequences. We conclude, therefore, that the abnomal sequence identified by Mr. Epler is not a dominant ontributor to PE: risks previously considered in WASH-1400.

Operating experience to date has deronstrated a riumber of occasions (about five per year) when for brief periods of time, one of the two d.c. sources at a nuc1 car plant has t,een degraded or unavailable.

Evaluation of the duration and character of tNs.' events and the occurrence frequency shows that the WASH-1400 probability estimates cited above are not significantly affected by this recent experience.

The duration of the unavailability events (degradation or loss of one of two d.c. batteries) has typically been less than a few i

hours. This is because typical plant procedures require plant status checks during each work shif t and because control room alarms i

l l

.. [

iernediately identify degraded or unavailable conditions. Typical i

plant Technical Specifications require a plant to shutdown if an I

j.

unavailability event cannot be corrected within two hours.

To date, the unavailability events that have occurred could more i

i appropriate'y be considered minor degradation rather than failure, as none of the events resulted in total d.c. system failure and for many of the events, the " failed" battery could still have I

perfonned its intended funct;on. That is, the typical situation has been a degradation of capability relative to what is conservatively reoutred to deal with a design basis accident, i.e., a LOCA. The residual capability in the d.c. system during most of these unavailability events has been sufficient to accorrmodate the lesser loads associated with normal shutccwn scquences following a reactor scrae, e.g.,

Osidual heat renoval, with considerable nargin.

l l

i e-e

18 j

t V.

CONCLUSIONS

{

Emergency electric.. oower supplies, both a.c. and d.c. for

}

g i

nuclear puwer plants are important to safety. For this reason, l

l the electric power systems for operating nuclear plants and those plants under licensing reytcw have been required to pro <lde a high degree of reliability. It is this high reliability that provides confidence that sufficient safety margin exists against loss of all d.c. power for extended periods of time to allow an orderly examir ation of safety isst.es, such as this, f

However, because of the irpcrtance of the a.c. and d.c. power systems, the staff has been expending effort to review the i

i reliability of these systems and shall cc.iti.ue to do so in the future.

i

{

Some of the past efforts to upgrade the power systerts include the development of NRC Regulatory Guides and revision of industry standards. For example, Pegulatory Guide 1.81 does not allow snaring of d.c. power among multiple imits for newer plants; Requiatory Guide 1.47 requires improved indication of status of the d.c. system because of its irportance to the engineered safety feature systems. Generic concerns related to the power systems should now be resolved more expeditiously than in the past. It is important to note that the staff has, as a top priority Technic.a1 i

h i

t

+. -.s

,=

,3r.4v w +4

-%4+'3'*

. ~.~

,a 19 -

Activity, the cn-going task of evaluating the " Adequacy of Safety Related d.c. Power Supplies".

l l

It is also important to note that NRR's recent organizational restructuring provides a mechanism for continually receiving and reviewing all significant Licensee Event Reports, including if those related to the electrical systems of operating plants.

t As such, the Division of Operating Reactors is charged with the continuti reviev of all operating experience, including all reportable events such as failure and degradation of power systems, and has a mechanism for raising any significant failures i

t, to the attention of NRR managenent.

g In conclusion, we believe tnat the d.c. power supply system is j

receiving appropriate review and consideration by the staff and e

}

that no change in tle overall NRC a!proach to this issue or changes in priorities and schedules is appropriate at this time, i

i t

1 i

I t

l 1

e t

I

.6 ee 9

i James Earl Ray Series KGB Questions U.S. Writer Begins Today in The Star For 2nd Time in Prison

- * ~ ' * * *

- - ~ - '

~" ~

~ ~ ^ ~ ~ ' '

!Mi! TheWashingtonStar =

L r

• ~

125M.

N. Its

~=m WASHETON.D.C,W10 450AY JUNE 15.19U

'h*= 8 8 **' #.cu = =.=.

_ _ _jl l$c g=a_

t Nuclear Regulators Checking Safety Flaw

,u.

i N.~.'.'

i.n a..

, a,

u,nc o,,=.

x.i n *.

.m

.u..,

.u

..a~.a.,......,.

~ u ~." -..o.sa a ~.~ -- me..

.. a u

-=.--u.

--*~.

Tu~.

t..

<=.-..~a

.==.

~.

=.~*~a--

n- -'.=. i - a,.-u. ~~~~=

a.=~.,-

ea

-a a..,..,.-. u

,..c.n.,.., Jl w

...,....~~=a<

i~.s--e-~d-a- -~ --..o.--.-.-

, s.*

. w.

e.

._r...,..,.,

r.o.

. v1

--e

=

.g....

u...

.,, m., *~~ ~.e-1. s =

-r ~~.o- -- c. *. = '.** ** *i

... ~.w a

..c..n..

~.

~~ ===

>. u.

=c --

i.

, -~

.. a

• d m

...a...

a

.....i...-..a.

.~=s..=-

4u'll;;'.,*l ""~,.*.s.' "..".**. ;.'.,.t!;;."..'.a.*a.*-. :'"","!

• Cll,,.*'.,. a.!!.'.l.l:':.",a": ::: ", ".fr.*2*"..'.' ".-l.","".*,*.; O. c.t. n..wr.u

'-,.'L**lN:" M'il a

.u......,

....o I2;

.,n.....

.--.. c.

n 4,w,.,~~

1.,1. m.

m

.m.

w.na a...

2; l ns,

w,...n.ne..o e m o

te.. m.io. n n=,n.n

. i. i aa..

u,,m u m.:n.

g 55 2

e i

i i,

NUCLEAR Ommenmd Prem M i

}

i One et the teamrtue en es pleet, seeresse by es f

Caressas peese and 14sha Co wee Immad to beve "

been eos deve ey a member af *pareasur" er' ammaefecy4 stated immetams thes had been euer eened to the bemerisa.

several meenhers of to prtC"o *Adotamry ram-

{

manes en beeciar Aaseywards.* a panes of amasade essene enaca reveres Na C poltenes omre e =-as were anstarted by the sesta careams ensednes anc e rash of esaer romass hemory

.C - - esas

{

ef iases see d by pasas samaan e rure.

Due messes. Jesse testemas suswer band em>

etmar engtmeer dre the Teamspes %almy As$ etry, pesarted thes the harteries were tan. "the of as h-earted te as eheca the pases's enlevy staes" mesher. Dr. De vid tetremL e predement of sechner engemartsg at the Wateerway of Cahier maa at &me Anteles, and the prehteens of the sed beterry synien gave ham "the uneasy feeisag that that is port of the meaangs ed efeat s el&& sali the breves 7 erry armouse revissted" (Drewes Ferry a e mecasar pseer plast east Athens. Als, taas w*J aarteusy esmeded whee a fire etarted by a taaLa barned as esy e emet mut of the plaat a see centred satte e&a whack were haussed en the same team )

SM0f ftY AFTER Eberwebe and Otreet's setta eten e April, the fiSC permoed e repen treen L P. E pier. an Osa Esdae. Temas, reacter espert obe emocluded that the lamare of one bettery send l

trissar "meauple agespuent taahares" sad egnal as au,tomaut plaat snandsen, em ca wei.id put e sever strom se the penas opereser and en las re-mamang besary system.

Several recent pressene beee skusdy made as speearance wasta woned hhe+y casse me fauere ef basa bemerees." esas Epter,sec&and edt e het of re.

i test encadeeia Eater seassedes that each a taaamre

• • ende tiearly ha esseeerabae" and ergoed that e furthee *releaba%fy eaalysas" Demed endy resee tb

• amastessary de 'a y."

Assed as esplam Ge NRC*e teenamn. *.*.reeder ensd Lpler was a *%gnJy reelected" espen as the inald, and added thes the probeer, ens *eas that we'd better pay smentsee to." *"There are certae scenarios obert New saigLt ksee d.J!sulty castrob bag ceness valves." he addet asse gave a more deta4ard espleasuse of obst semid bappen if a suc6sar paaat on.aaesty bad sta estaade per er eaurte cut e# Dorsel settssears, he esad. eeund amur.4 en preendaag poest to a best.

res*

• eyews ensch weidd f===4 pumpeg osie, einesh Ge suckeer fuel care, carryiss '

easy beat and rehaving pressors. A secsad efe-l tem. the *es.orgeery core seebog sysus" would be et the re,ady. prepared to fluaa the care estaester if a p e breaa ansied etaw se tae f.rst sam 6-ing sysian.

A DOL *BLE battery feuere, eeeardang to Roma.

eeuld sulhty all tarse entens because tre betiery i

peset te needed to stan tae generenors and to operste the switches and estees anat contro! beta canhag systems Othout seehag, e austear feel care tsuid begie to everheat. melung tts esy e

through the reacts

• and the reactar o cement floor, as acustry emat,a eegeeers cau the "LD:sa eye-dree?." becaea no see reany knows bee far the me' tee. tediosctave twel eeut1 st inie tte earta "feers are combcnataans el circumstances that

{

sould get you tale very servoas troupie Up to ese the presret desits eas reasidered adequate." said Rosa. Neser p.asta. De sad. are assigned with esore reserve be.tery erstems.

Det any retrosetaee C&sese to des:ge etweria for older plants. be esp'aimat'. es ?;ee are goes. der-ab.e tasse becoves yed *e get to de at methode caDy and carelvby."

i l

l I

i l

l i

e

1, i

s 4e s,

./.'

UNITED STAT'E5 ATTACHMENT B i

l

[ ',.f.(7, i NUCLEAR REGULATORY COMMISSION l

~.// l AisvisORY COMMITTEE ON REACTOR SAFEGUARDS j

4 wasatsasGioas, o. C. 20$$$

%...../

l 4

l

(

5 i

I E. C. Case Acting Director

{

Office of Nuclear Reactor Regulatien

.'I RELIABILITY OF PWER StlPPLIES l

References:

i (1) Hemo from M. Bender, ACRS Chairman to L. V. Cossick, EDO, I

" Reliability of Power Supplies," dtd. March 15, 1977 l

(2) Hemo from R. F. Fraley, ACPS, to E. C. Case, NRR, "D.C.

System F411 ability," dtd. April 26, 1977 l

I The attached report frca Mr. E. P. Epler, ACRS consultant, provides information applicable to th6: reevaluation of D.C.

power supplies requested fri References (1) and (2) above.

8 F.

F,. Fraley Executive Director i

Attachment:

14tter from E. P. Epler to J. C. Ibersole

& D. Okrent dtd. April 12, 1977,

'e. D.C.

pcuer supply l

9d*

I R. Heiner.an, v/att.

R. Boyd, w/att.

h. l l

7-

-.p-

p.tct*.E:

l w."u. ymn,.=

' ' ' * * * ^ "

~

/17]/ l2 /D77 9

J.c. Ebersole 0.ORrenl-The rehi>bilit a f //,e n.c. posser suppl y

en,ne under dinu.ssion aJ /he taprit a meeIu'>q o/ /he ncns.

The single failure cellerion has been appIdel in Jhe

}

de. sign of Jhe balle/3 sejslem and ilis lhe stryulalo,,

posillon /hol reliabili1,so o blained, is adeguale.

l 3

l N suas sue,yesied howeru Jhai a rel&> bilti5anolsti y

s/>on.1d be pericemed.

The lollosoin3 are ins, tomnien i H is generath a7< ecol Jhal Jhe availa bi!il of D.C. is essen/ial k re,sidual heal removal, // is y

impor/ani Jhal Jhe deg>ee al lhis dependence be eslablished; ?f ilis indeed frue Jhol lhe loss of l>olk ballerie.s would invarial,),ar evill hi h proba bilti,

g q

teo d 10 tore mell, / hen./ h degree of dependence si,ould be branl and picci eh e.s/o blnherl. // wovid y

5 g

.lhen be reyv1<ed Jhal lhe D.C. supp y,which would be l

called upon for residual heal rernoral Jen June: pw year, ha<c on a<aila L;hl al tens) equ;raleni k

.Ihn1 e! Ihe neoc10.' shuidown s.j.s Jen s whic 1, is elonllenged no n> ore Jhon Jh<ee Jilnes per, eor, APRO.GTiT PD.DU

n. :.

1

// has been cbserved /In.ol As inos/ lininnces, when a.

single ballers; Imd failed, a scrous resulled iroon li,e failure, // svovid seem lherclore, JIon}.shov/d bola ballerie. fail, a scrorn wovle) invaria bl Jolloai. //

indeed il is fruc Jhal re.sidual />en) cos.id ml Le l

ronioved on / css of bolh isollerks, ihrn a scram resullds.; frein loss of holl, boileries wot,Id clearl

{

be Enlofernl>1e, // h /herefore mendalo.-

JhaJ tos.,

of heal reinoral capc-Isilil a nd renc4/ sevom he he Independen) y even/s, snode Jo The la1lure of bolk bnfleries is no) hi<g,1q l

imj>ro l> a lsle, Du rons; JIse period Aug 3I l'17't la War 5,/474, opp.<oxim ale 13 lwo y eo es, Ihere ware.ro reycel< ble occuirnisc e s relolo've k Jhe o. c. s u,,p h. T h e Jollowl.,x y

bn Hers; lailure.s ca u rred.

Dale n.po, led Le< nIan.

Ercu:l-

//or I? 19 74 Dresr}en 3 On llr* 'j Nound lo be In detjroderl condll ton riurinq re[veln'>q o ulatj e, caused by prior charger Jrovbles, Boller,) replaced.

Mo,- 12. 1976 v1 Yanlcee Da lle< s; ec tla g e losa.

charger failuie. 17ea c /o < a) power.spn> c ba lle,3 imonedialely connecled.

I B. 3 e eb w

l l

1

l

~

Nois 1176 fnlisades Dus voilarj e. dropped to lo v, ori I;ll ps np covs ed <hnry er breaher 4 (paras: lie )le/ i,s Iripa llent s hvidows, condllo'on, Jul 2G 1974 1]rewns Ferry 2 Cell pas) brolan n/l y

rivring clean lo,c;. fienclor in sheldown rondllron Isoweref.scrain signal wn.s ef eircra.ler).

July at 1976 (febinson 2 Lends removed from ballery sv10 k re a c lo/ c vilu't.nl, t

/176 Z ic n 2 W;lh r/Osel 7enern /cr in 3ej>0 19-opera l son -lhe los.1 oJ one Dt. hu.s preven /ed tirtu;-l brea hers leein renso vin <z ex<essir<>n.c.

ton Js.

As a resu i Jhe diesel generaloi wincliisp were des /raye</.

t Jerain re> vller),

Orc. 2 3 19 7C 6.jslcr creelt., buying et Ic's) o n.lh e M Y D.c. di> } rib uibn con lc/

olalion hollerq Jhe tun.s de rner.jh.ed wilh reockr al pweV, Shuldown yesvlled Ave; ll 1975 Cconer 2.

35nullaneov> oulcige of &

l bra fle>les. One.sdria.3 of fleowee and one slens,; el

.swilchhs.;.11alion baller its wes e brahy charged shnvlloneo us l,

y 6-4 i

6 Wu'I C/u.s gsJt>fou C gjt>jtv/Sa jarj pnae,ub I.ayaafouul l

f * * f T ' ' 9 l " " V ' ? Iis fu fc ugj goe jj>.sr,as '

O'f ZC iolir/ ll /' J I%!l 71 O!tPiU)bJ fer-( fons'if qvy 9u jjsv\\' ivnrap jrol ossAptu>sle eb-

)ulit I t ov2 9 stp ro i.u s usup u).pr

)gaivr:(*., ( o j dnrute suarJy 1.yot dJ's 7v.vy os yc pyd grsprujbJ>f fru jjuss\\

a su nta p r2>v ut' gfJ>su fO:ln>ot o>>assap 1v olc/ sox ltetu jJ(N mo y

$ avis,yte)(',no/gr oy y ga 1q,,leus sac)los baoJ" ffeJrJ n'aJJ rtelh a qu jjs/sI fojjns,Jr (rotnats/ pma t

t a.atpc ?.on;*( 94 )vfsbos?2ap uv tw :drairf z-70/fdA' f>(1ln/2t'

/U off'.l? f* ft'2 7u jjss,I fv;jn.st fteasa moja '

fv jjoJa mjtisj' gnuotsI ffejs dsitcif*tr. )utar,y sjeics6se s

• • "IP /JuP fo qocjj?ssI fou jnsa 90ruJ utasfoottteues uro'fO ftsa]/ olsIc3r,owJ m[ts)(' sne njp u ffJ> f qof t\\ gu ffattar.

/* jffJ )ol )'13 d'ofJDftoit $'?rfJtot*MltJrtfJrlJJj')Vnedf

[o sf fO ))1C3 }. O !t/f pggyt tegnj gjt02 j)' )UnrilI' flu llf4 tsuj,ajJ>f*oj,Ip?*

g* yno lJ 1.ojjub J tnl>?sro>s full>tf usup' ruo>> ytro es oresa' s u n t J p fu lln< >. o j q o jis q v pjs u ye v t.! >>r.

l lT'S,

J. A. chi >tg e/ ><as lovnJ Jo se ope <nJo'>3 nJ lui! cn,soet/q Lecavse oi mullijte tenkage pnllis le grau,,d i1,rovg h s of s init bene e.s _

t Lvs lie brenhe/ nhi<h runneel.s lhe Lo

4. */he I,nn,)3 hotlesies and Jherch3 v o tales l1, depen.dente of jse Jwo D. c..s v,,j>Ile.r -

If Jhc failure of one of La ballerie.s wntd a.J cause renclos shvidowes. In,4 wacId be delerled and

\\

Jhe bn flery reJu ned to. service in ten locur.s, Jhon lhe (<>tiv?e role of to 'per yeo/ wov/d yield on ennentin btliI of lo"per yeo/, T/>is woul,/ nioke eress n.,n ', b.,ll)ery nInjosi ortepin ble. TI>e indep >ndon y e/d n n l

Jodvec cl bolk bolleries would 4 hen vnn tnila b;Ill)la cl ort).p o r y e n v' eu l,d A el /c' b l

en-lirel

. salts As we hoic.seen howere/; lhe y av.se.s ieni/c/ shuldowrs in;Iuie oI ene bnflers, vsunIl r

hoc Jirof e renmisuh3 and residval heal navsl be reinored v>lng l

haflerug, and a mid geneint con /u.skn re.s vili;,,

f,c;..,

l mulliple eyvipmen) Inilures. Fus/ bro: sco cen t medinnii no s bare aliends; anode n n nyp eovanc e uho'cl. we.vid lik.elq l

muse Jhe Inilu, e al holh bn fle, Tes, l

Jo,,,n,o<,,

l If ll is roo fien>.~,1 ft,,/ !he D.C..suj, ply s essenlInl i

t B. /-

I to rearnun s nerlf ressocial,4/sht 4/ot' do!! awing ynoscip/ts i

applainhib k Renclos Jh ldowr, S..sl.eno s wo uld be =

l nypin able h, Jhc D. C. Jaj>pIq.

.l l

J. &hulp Jepn/nliin of safel.; anr} conlrol.

Rensedu Rensore frorn j!,e haliesics, dedot oled la res$lani hei:1 reinoral,all lenlutes s hotIn could ta vs.~. o r le a d to. re o cle.v sc rain-2 Pr *> < fo le Prolecl..'rs ay.sless.s shnl/ be dednaled o

and used 4r no purpcse o{iser lhan puolec/ dot.

Ronred

.Ronsore o f) porn 3llA. hads,

y 3 P_r__p1ciple - Redunr}a n} cl> nan els, ot lrnins, slooll be I h' efent/enl.

llerne 1 Reinore lhe bus die brenh w whnh is,uiles 3

Ihc Jailure of sclh ho lifvies.

'l ?rfIeiple

'I~/s e 3yslen s 3 lao ll reres l No a se(e fond llicos on loss ol fowef Rensidq Mo remeds; is possil:!e. As a inin uis un u

.lhe chesuje/ shoufr>I br Jit tr] 30 os lo on 1[on(2 e./hr nete] lcr.Ils e naso Pire13[re poclrtli n,nlunh is /hr princij,lr en vse of clon er;e< una > niin b,'Iils,.

//nt >r; runde lirese toriet/s*ns ;/ nov/d be nyproprin}r lo ash. whelher Jbc 0J. 5ys/ ens aonilaLllll svo vl,} Le s 'I i

_m_

I

eorr,juntable la lion) ul lhe Reaclo/.Wlrlo uin S alens y

inn derj u nlr.

I sho'c lo is ilsell dreiner) 4>, alon e be s

i i

//1s </ew lionI./!>e.sd><,le in;luie c<1lerionis inndegunie la ns.s vee a relia ble D.C. Jujyh. /J /s z

nis.o r/eo / Jt.n l n relinl>;l;/3 anoiq.rss of 11 e pinwl 1

l s.jslens usovid yield no uselvl adal;lionni inforin,ikn ani.' wovIr} eciull in unn e ces.s o <9 delo y I

4, I

i 1

{$)

l' i

l l

i

(

i i

i APPENDIX A l

6 i

Analysis of Light Water Reactor Response to Complete Loss of AC/DC Power With Scram l

t n

i e

Revision 1 June 23,1977 t

i e

6 I

I d

k

//

9 I

4-/

i

)

introduction The purpose of this appendix is to sumarize the results of calculations i

to determine the effects of a complete loss of all electrical power on a Light Water Reactor. A tvoical boiling water reactor and a pressurized water reactor were analyzed to determine the amotnt of time from the I

initiation of a complete loss of all electrical power until the retctor core midplane is uncovered. This criterion was chosen since no fuel or systems failures would be expected before reaching this condition and

{

since restoration of power and/or other approp 4 ate action, including the possible implementation of steam driven pumps that require no electric power, would result in returning the reactor to a safe and stable condition.

Accident Scenario Following loss of all power accompanied by reactor trip and turbine trip, decay heat would be generated in the reactor core at a rate of 7% of full power imediately af ter shutdown, and decaying to 1.9% at 15 minutes,1.3%

at I hour, and 1.1% at two hours following shutdown.

For BWRs, the main condenser which provides the normal heat sink would not be available. The reactor coulant system, typically operating at 1025 psia saturation conditions (548'F), would then risr. in temperature to generate saturated steam at 556*F, corresponding to typical steam pressure relief valve setpoints of 1100 psia. Recirculating coolant is normally only 41'F subcooled, so that the slightly loor temperature metal mass of the reactor vessel and recirculation piping systems providas only a small heat sink. As system pressure increases above safety valve setpoints, spring loaded safety valves would dump steam from the reactor coolant system to the containment suppression pool water and would rescat as the pressure subsided. This process Meg,mg e.

-=. -

2 Would continue until a normal or emeroency heat stak were mada availahla t.n the Fuel claddino temocratures would not risa AnnrPCiAhly Ahnve the reactor system.

coolant temoerature until the core hoof r.; *.a hacnme uncovared, a tw' nhesh froth in the upoer oart of the enre would nenvida adeounte coolino (Saced na conservative heat transfer calculations usino Dool boilino heat transferl until the enalant licaid level droceed to the core midolane. This crocess is exoected to ornvide adeouate time (at least one hour) to make normal and/or ome-cency heat sinks s

available-to the system.

For PWP.s. secondary coolant would be isolated in the st-3.m generators and would provide a heat sink until the steam generators toiled dry, exhausting to the atmosphere through safety relief valves. The primary coolant would then begin to beat up from typical temperatures of 554'F it. the cold leg and i

605'F at the core outlet ard in the hot leg. The pressu-izer, containing the i

steam bubble and the primary system safety valves, is initially at 647'F.

The primary coolant heatup rate would be slowed by heat losses to the metal r

mass of the reactor vessel and piping. The liquid ccolaat would initially expand due to heating and would cocpress the steam bubble ;o raise system pressure to the relief valve setpoint (2500 psia). The steam bubble would t

f then be discharged and the pressurizer would be filled with water. The

{

safety valves would alternately cpen und reseat dependent on system pressure i

due to the continuing water expansion and neriodic water disc'.arge processes.

While much of the discharge through the safety valves would be in the subcooled water condition when leaving the system, considerable flashir.g within the pressurizer would occur prior to resesting of the safety valves after each opening. Boiling within the system removes latent heat.from system coolant and thereby delays the time req 9frf3 for system hea":,.

JpS AAd a.e' h.

a Ma

-M

,-y y--,

e

.. -._ ~.. - - -

- -., _ i 9

-y m

r A After the primary coolant system is heated to 2500 psia satt.re. tion temperatur (66B'F).t the core outlet, & steam bubble would begin to fora in the reactor vessel upper head and this would be:ome the high pressure poir.t ir. the system.

I At this time flashing steam would predominate the discharge at ti.e lower pressure location of t.ie safety relief valves. Stee liquid disch ge would continue until the hot leg leading to the pressurizer and the pressurizer became filled with steam. Subseqtent discharge wauld be stea.n w:th negligible entrained liquid. M-for the Bh9, steam release wou'd contit.ue and adequate core cooling would be tr,aintained until the reactor coolant liquid level dropped to the core tcMolane. This process would take at least two hnnes dich should be sufficient tt ce to ake nonnal and/or emeraency heat sinks available to th system before that condition is reached.

Method effalysis This analysis is based on i non-mechanistic loss of all electrical power.

{

That is, no spe:ific event or series of events has been postulated which would Icad to a complete loss of all electrical power. The analysis asstnes that all a.c. and 1.c. power is lost simultaneously (onsite, offsite end eeeroency t

u power). The analysis also assumes a turaine trip, reactor trip and complete loss of feedwater floe at the initiation of the event.

The Boiling Water Reactor chosen for this study was the Hope Creek reactor j

j (Public Service Gas and Electric Company) since this is a large Bh'R for which i

detailed information was irr.ediately available. This reactor is ennsirtered to be

{

typical of Bh7's for the analysis of the loss of all electrical oower. This analysi is 'aased on an energy balance between the core decay heat and the energy removal

)

)

i t

E

~ ~

a

I i

from the reactor by heating and vaporizing the reactor coolant. The

. analysis wts carried out to the point where decay hert had raised the temperature of all of the reactor coolant and the retal in the reactor coolant system to saturation temperature at 1100 psia and had discharged thrcugh safety relief valves that emount of the vaporized reactor coolant to reduce the reactor coolant liquid level to the core midplane.

The Pressurized Water Reactor chosen for this study was the Oconee 1 reactor which is a Pabcock and Wilcox reactor. This reactor was chosen because it includes the once through steam generator design which has less liould mass in the steam generator than other designs. This reactor is therefore believed to be a limiting type for a PWR. This analysis included calculation of the i

j point at which decay heat had raised the temperature of the reactor coolant to saturated conditions at 2500 psia (primary safety valve setting) and caused discharge of the pressurizer steam bubble and subcooled liquid to acconnodate liquid expansion during heating. At this time, 52 minutes after reactor trip, it was assumed that fit.shing occurs within the system and saturated steam was discharged at the safety valve tr. accommodate decay heat addition until the reactor c Slant licuid level had dropped to the core cidplane.

In addition to the decay heat nccessary to completely vaporize the initial mass of water in the steam generator, the decay heat necessary to raise the temoerature of the metal of the reactor coolant system to the saturation temperature at 2500 psia was included.

An additional PWR calculation was performed with a more c:nservative assumption on the pressurizer liquid discharge. In this calculation it was assumed t$at the liquid discharge through the pressurizer safety valves continued until a i

W --

m.

jf-S-steam bubble formed at the core exit and had become large enough to fill the reacter vessel upper head, the reactor hot leg, pressurizer surge line and the pressurizer. This conservative calculation results in th2 liquid level reaching the core midplane 19 minutes sooner than results reported in Table II.

l t

1 In each analysis the decay heat values were taken from the ANS decay heat standard curve.

Recovery from the Accident

}

Recovery from the accident woul,i normally involve restoration of a.c. power from nomal offsite or emergency onsite diesel generator sources accompanied by restoration of d.C. power via battery chargers for the d.c. power supply system.

For a B'.P., nonnal feedwater system pumps cculd then te employed to supply make-up to the reactor coolant system until normal reactor coolant inventory for the shutdown condition is restored and normal decay heat removal systems are placed in operation. Steam relief could be accomplished via turbine bypass to the main condenser.

For a PWR, the feedwater system could then be activated and the stehn generators would act as condensers to relieve primary steam which may have accumulated at the reactor outlet. Borated make-up water would be supplied by the charging pumps which are de:igned to deliver crolant against pressure greater than that where pressure relief occurs.

Emergency coolant injection systems are also available if needed for supply of l

high pressure make-up coolant for BWRs. Emergency systems would be usec' if 1

1 n.

n.

~--...s..,.......

the diesel generators were the only available a,c. power source. The main i

condenser would then not be available due to lack of power to the circulating water pumps, and there would be no power available for norwal feedwater pumps.

The Reactor Core Isolation Cooling (RCIC) system and the High Pressure Coolant Injection (HPCI) systems, operating on steam driven feedwater oumos would then be available for both coolant injection and the heat sink for most BWRs until pressure has been reduced to the point where the Residual Heat Removal System can be placed in operation. (Alternateisolationcoolingsystemsare available on early BWR designs.)

For PWRs, primary coolant injection would be accomplisned via the charging pumps; feedwater to the steam generators would be provided by the a9Filiary (emergency) feedwater systems. When the primary pressure has been reduced to approximately 1500 psia, the high pressure injection system would be used to supplement the charging flow into the primary system. The heat sink during i

{

the recovery would be provided by discharge of secondary steam to the i

2 atmosphere and by decay heat removal systens af ter the prirary pressure has been reduced to the appropriate level (approximately 300 psia).

Results The results of the above calculations are as follcws: For the Hope Creek BWR (3293 M'n.) the time for the reactor coolant level to reach the core l

midplane is calculated to ie 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 6 minutes. For the Oconee 1 PWR (2570 MWt) the time for the reactor coolant level to reach the core midplane i

is calculated to be 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 10 minutes. A detailed tabulation of I

3 calculation inputs and results is provided in Tables.I & II.

i f.

1 L-

~..

. -., - -. +

J l

l 1

l l

~

1 TABLE I BWR (3293 MWt)

Iaftfal Reactor Coolant Mass (Ibs.)

549,000 t

9 f

Initial Reactor Coolant Energy (Btu)

.29 x 10 6

i Reactor Vessel Metal Mass (1bs.)

1.5 x 10 3

Reactor Coolant System Volume (ft )

21,140 Time for Liquid Level at Core Midplane

~

(min.)

66.*

i 9

Reactor Coolant Energy at 66 min. (Btu)

.12 x 10 Total Mass Discharge (1bs.)

322.000 For the case where offsite power is lost simultaneous with a i

reactor scram, this pararreter will be reduced to 58 minutes due to the stored energy in the reactor fuel.

i i

I 1

i

. - - - ~ ~ ~

,.,-m,.,

.y_,. - -..

I 3 +.. :

c,- v.

P,.

~

}l@ y

-[,

. zggLE II

- ~ >

5 z

4

~

a s-

,e p,

,n PWR (2570 MWt) Calculations _

m ;. ' '

.f [.

a.,

-,m

.3 JJP yg,'if.

1 O ; KM,. - D ~

1 s

'i j M%rm.Qy'4

-9

,.:.,(,,

L

c-

,g

,e 3'-

~ ' '

s.

s 3

"- y [g 5.086' x 10 4l 5

,,, g

.\\.

~ Initial Primary Coolant Mass (1bs.). -

=a-,

s 9h. L,7....!,

,' N

- g

+

jf0

. Primary Coolant Initial Energy (Btu) 3.5 x 10

t.

,sw 3

,(

6 e.er 5.11 x 10

$f

-../

.P.amary System Metal Mass (1bs.)

+

i. Secondary Coolant Inventory (1bs.)

110,000

?

i c

g 4

1 Q

Secondary' Coolant Initial Energy (Btu) 1.17 x 10 1

7, 12094.

' 'j -. "

g 3

., Volume of Primary System (f t )

,)

((

. Time to Empty Steam Generators (min.)

19

.a.

,-. Time to Heat System to 2500 psia

^

ST Sat. Temp.(min.)

'I 3;

m.

5'

~

f

.' t Mass of Subcooled Water Discharge (1bs.)' 1.05 ). 10 Time' for Liquid Level at Core Midplane 130. *

(min.)

5 3.37 x 10 Total Mass Discharge (1bs.)

~

f.-

+

9 For the case wtiere offsite power is lost simultaneous with a 4

reactor scram, this pirameter will be reduced to 121 minutes

~-,

due to the stored energy in the reactor fuel.

I.

.~

a s

.s e

4 b

^

' & g,g gy

.y g

-.AA

- O