ML20009B438

Forwards Preliminary Draft SER Section 5.2.2 Re Overpressure Protection.Any Open Items Should Be Identified & Resolved Promptly.Comments Should Be Submitted within Two Wks

ML20009B438
Person / Time
Site:	Midland
Issue date:	07/02/1981
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Jackie Cook CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
References
NUDOCS 8107150448
Download: ML20009B438 (13)

Text

_ _

y ;,: w

. q g

I gy -

- t 1981 u,

~

4

. Docket Hos.:-

0 g

%y B

g%

~

Hr. J. W. Cook pb 2

Vice President L

Consumers Power Company

-d 1945 West Parnall Road 9

5 Jackson, flichigan 49201 m

Dear lir. Cook:

t

SUBJECT:

TRANSMITTAL OF PRELIMINARY SER DP/JT SECTION 5.2.2, MIDLAND PLANT, UNITS 1 AND 2 Enclosed for your review and comment is the preliminary draft section' of the NRC Staf f's Safety Evaluation Re, ort 5.2.2, "Over-pressure Protection."

Your attention is directed in particular to any open itens contained within this draft sections. A principal objective of this transmittal is to provide for timely identification and resolution of any aeditional analysis, missing information, clarifications or other work necessary to resolve outstanding imes.

Please contact the Staff's Project Manager regarding the need for any meccings and telephone conferences to this end.

Your coments, including schedules for completion of any further analyses or 4

other work associated with resolution of open items, are requested within tuo weeks of tnis letter.

Sincererly, Origtnal signed by Ipbert L T.d so Robert L. Tedesco, Assistant Director of Licensing Division of Licensing cc: See next page

Enclosure:

As; stated 8107150448 810702 PDR ADOCK 05000329 E

PDR

• See previous white ag. #42......

.^.P..:.'$.'..........

9L:W4 o,,,c.,

h..9.d;.e.........B..

.. m..... N.edes.9.- -.

.u,,,-.>

NN.....l.8.1.........ll.hl.N........?l Yl.8b-~~.

omy

- - - - ~ ~ ~ -

~~~

• " " ~ ~ " " " " " "

" " " " " " " " " ~

Nac roau sia po+n NRCM CNO OFFICIAL RECORD COPY -

~

umm mi-===

Q m

y,

+

1 DISTRI'1UTICM Docket File OIE(3)

LB#4 Rdg bcc: TERA /NSIC/ TIC DEisenhut L/PDR EAdensam NRC/PDR DHood ACRS (10) ffaj Mr. J. W. Cook co Vice President RVollmer Consumers Power Company 1945 West Parnall Road TMurleY Neld Jackson, "ichigan 49201 R

Dear Mr. Cook:

0 ELD

SUBJECT:

b TRANSMITTAL OF PRELIMINARY SER DRAFT SECTION 5.2.2, MIDLAND PLANT, UNITS 1 AND 2 Enclosedsf r y\\

a,*

our review and coment is the preliminary draft sectionVof the NRC Staff's Safety Evaluation Report 5.2.2, " Overpressure Protection."

Your attention is directed in p ticular to any open items contained within this draft section; A p'rincipi o jective of this transmittal is to provide

!W for timely identification'Eqd resolution of any additional analysis, missing information, clarifications or other work necessary to resolve outstanding issues.

Please contact the Staff's Project Manager regarding the need for any meetings and telephone conferences to this end.

'N Your comments, including schedules for completion of any further analyses or other work associated with resolution'of open items, are requested within two weeks of this letter.

\\

s Sin ererly, Original signed by Robert L Todasoo

\\

Robert L. Tedesco, Assistant Director of Licensing Division of Licensing cc: See next page

\\

Enclosure:

As stated

\\

bec: w/c enclosure

\\

W. Jensen G. Mazetis P. Check D. Hood w/ enclosure

\\

" ' " > DL:LB 4 DL : LE# A.........

. AD/L........

" "*"'I Diloodio.

...E Ado nsan..........RTodosco....

""> 6 /, u /,a.1......, 6 L../a.1,,,......g/,,.(a,1,,,,

n

w

cro w m n soin=cu ei c OFFICIAL RECORD COPY

• * '"o-32"24

,c?.

1.-

A e

s.

JULY 2 1561 '

! MIDLAND m

?.

U

- Mr. J. W. - Cook

' Vice :P res ident -

Consumers Power Cogany 1945 West Parnall Road 1

a hj

' Jackson, Michigan 49201 cc: ' Mi chael -I. Mi.11er, Esq..

'..Mr. Don vin' Farrowe,- Chief

~

i Ronald G. Zaririn, Esq.

Division of Radiological Health

' Alan S. Farnell, Esq.

Departrent of Public Health Isham, Lincoln & Beale 165 P.O. Box 330351

- Suite 4200 Lansing, Michigan 48909 1 First National Plaza

Chicago, Illinois 60603 William J. Scanlon, Esq.

2034 Pauline Boulevard Jares.E. Brunner, Esq._ _

Ann. Arbor, Michigan 48103 Consumers Power Company

. 212 West Michigan Aver.ue U.S.. Nuclear Regulatory Commission Jackson, Michigan 49201 Resident Inspectors Office Route 7 Myron M. Cherry, Esq.

Midland, Michigan 48640 1 IBM Plaza Chicago, Illincis 60611 Ms. Barbara Stamiris '

5795 N. River Ms. Mary Sinclair F-reeland, _ Michi gan 48623 5711 Summerset Drive Midland, Michigan 48640 Stewart H. Freeman Assistant Attorney General l

State of Michigan Environmental L

Protection Division 720 Law Building Lansing, Michigan 48913 Mr. Wendell Marshall

' Route 10

-Midland, Michigan 48640 Mr. Steve Gadler 2120 Carter Avenue St._ Paul, Minnesota 55108 w

w

t - h[. K [

- ENCLOSRE-a v.

,+

PRELIMINARY DRAFT SER FOR SECTION 5.2.2

t,

~

5.2.2 Overoressure Protection General Design Criterion 15 of Append'x A',10 CFR 50, requires that "The m.

Reactor Coolant System and associated auxiliary, control, and protection systems'shall' be designed with sufficient margin to assure that the

~

design conditions of the reactor coolant pressure boundary are not exceeded during any condition of nonnal operation, including anticipated operational occurrances."

Anticipated operational occurrences, as defined in Appendix A of 10 CFR 50, are "Those conditions of normal operation which are expected to occur one or more times during the life of the nuclear power unit and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine generator. set, isolation of the main condenser, and loss of w

a,ll offsite power."

Appendix G of 10 CFR 50 provides the fracture toughness requirements for restor pressure vessels under all conditions.

To assure that the Appenoix G limits of.the reactor coolant pressure boundary are.not exceeded during any anticipated operational occurrences, Technical Specification pressure-temperature limits are provided for operating the plant.

Over-pressure for the reactor coolant pressure boundary is accomplished by utilizing the two safety valves and one safety grade pilot operated relief ' valve manufactured by Dresse, Inc.which vent the top head of the r

pressurizer. These valves discharge to~ the pressurizer quench tank through

. headers from the pressurizer.'

4

t i,,.,.

4, :

,.2 -

,~

n-

The pressurizer safety valvas are bellows-sealed, balanced [ind spring-loaded safety valve's which',t.y'e provided with auxiliary balancing pistons,in_ the.eyent of damage to, the bellows. The valves are, designed,,

fabricated, tested, installed,andcertifiedinaccord[ncewithSection

'III of.the~ Amer.ican Society of MechanicapEngiheers' Code #

~

~

for Class 1 components.-

Each of the two installed safety valves.has an installed rated capacity of 279,845 pounds mass per hour at three percent abou the' r 2500 pounds per square inch gauge relieving set point.

They are designed to prevent reactor coolant system pressure from exceeding 110 percent cf system

' design pressure.

The Pressurizer Pilot Operated Relief Valve (PORV) is sized to limit the Reactor System Pressure during load changes including a 1007. load rejec-tion, to a value less 'than the high-pressure reactor trip setpoint. The 1

PORV has a rated capacity of 101,643 pounds mass per hour and is set to open at 2260 psig which is approximately 100fsi below the reactor trip setpoint.

It may.be isolated either manually or automatically upon coincident indication of PORV not closed and low reactor system pressure by either of two Class 1E motor operated block valves installed ahead of

..the-relief valve.

Both thE PORV and block valves are ' Class A defined by Regulatory Guide l'.26, and are designed, fabricated.-tested, installed and certified by the requirements of the ASME code for Class 1 valves.

Preoperational testing will be performed to demonstrate the actuation of,

'the pressurizer safety and relief valves in accordance with the ASME code.

Experimenta1' verification of safety valves, PORV and block valves perfor-d-

g' d.

a 3,-_, _ _. l, y.

.s

~

mance'[rerequiredbytheiost,_THI-2 requirements _asdescribedinNUREG.0737__

subsection II.D.l.(see. Appendix * ).. Environmental qualification of the safety valves, PORV and block valves is discussed in Section 3.11

. of t'his SER.

3 The pressurizer quench tank has a volume of 1123 ft and.a design pressure of 100 psig. The tank is equipped with a rupture disk for overpressure

,. protection-with a discharge capacity greater than the combined safety and relief valve flow. The' applicant has calculated that the rupture disk will not be challenged by the mos't severe operational transient. The quench

~

tank serves no safety function.

Overoressure protection at Operatino Temperatures a.

The' pressure relief system is,designe5 to_ prevent overpressurization' 1

of the reactor coolant pressure boundary under the most severe tran-sients and limits the reactor pressure during normal operational transients. Theperformancec.apabilityofthesystemwasanal.pzed by the applicant. in accordance with the requirements of Section III, Article 9 of the ASME code.

Only one of the two spring loaded self-activated safety valves was assumed to functioil.

No credit w.'s taken for the electrically actuated PORV.

. The., Babcock and Wilcox topical report, BAW-10043, "Overpre'ssure Protection for Babcock and Wilcox Pressurized Water Reactors" was referenced as the basis for the design requirements of" the over-pressure protection system for the Midland units. This report con-cludes that the pressurizer safety valves provide sufficient redundant overpressure protection.

• LPM to add Appendix number.

T

\\O l...,...

y

~.

3 1

.g.

y.

~

The analyses presented ih BAW-10043 utilized input assumptions'which

.c are different1from those of the. Midland plant and utilizZ the CADD and A.

vs

. POWER-. TRAIN Computer Codes.

The staff's review'of CADD a'nd POWER TRAIN is discus. sed in Chapter 15-of this-SER. As a result of staff questions, the 3

applicant repeated the overpressure-calculations utilizing input assumptions

--~

applicable - to. Midland and-utilizing the,gADDS computer? code which was n:-

reviewediand. accepted by the NRC staff, for ATUS applications. The staff review J-of CADDS.is documented in NUREG-0460;" Anticipated Transients Without Scram for' Light Water Reactors, April 1978."' Tile CAT'"

code was found to' calcula'te conservative results for-ATWS applications ATWS. transient modeling involves t

the prediction of the reactor system overpressure produced by prtimary to secondary heat flow nismatch which is the same phenomena involved.in the verification of safety valves sizing.

The non-eauilibrium pressurizer model of the CADDS code provides for a conservative predict' ion of reactor coolant pressure and the staff concludes"thh5 the CADDS computer code is an appropriate method for analyses of safety valve sizing.

NUREG-0460 contains a committnent by B&W to verify CADDS using experinental plant startup data.

We require that the details. of this progran be submitted for P,idland.

input assumptions were utili:ed in these calculations as follows:

Positivemoderato,rcoefficient(+5.0x160 ae/e*F) a.

b.

Small Doppler coefficient '(-1.46 x 10' t.e/ e* F) c.

Trip setpoints at. maximum tolerances

'd.

No. turbine bypass or power operated atmospheric ' vent valve actuated

.e.

No ICS action; and no reactor runback feature f.. No. direct reactor trip assumed to occur from any non-RPS signal with the exception of the loss of station power in which

( i, +,....

~ ~

., ~

n

+.

i

(

l control rods are tripped by the loss of power to the CRDMs g.

No pressurizer pilot operated relief or spray actuation h.

Pressurizer safety valves assumed to have 3% accumulation i.

Only one pressurizer safety _ valve is modeled, representing Thalf _ of the actual installed capacity.

The transient producing the highest reactor system pressure was found to be a control rod withdrawal from low power which produced a peak reactor

. system pressure of 2720 psig at the pump discharge.

The ASME ' code-oermits reactor system pressure up to 110% of design valve or 2750 psig.

Thus the ASME code requirement is met by the Midland design.

.We note however that the calculations describeY above do not meet the requirements of Standard Review Plan 5.2.2 which require that the reactor scram'is initiated by the second safety-grade signal from the-reactor protection system. We require that this assumption be considered kq in the applicant's safety valve sizing calculations.

In particular, we require that the aoolicant show tfiat the calculations are I

conservative for antic,ipated transients assuming failure o'f the first g.

i~

reactor trip, but with full credit for both spring actuated safety valves.

We will report on resolution of this area in a supplement to this report, e

'b.

Overpressure Protection at low Temperature For the Midland plants, the equipment providing overpressure protection of the reactor vessel at low temperature is both redundant and diverse.

Protection,is provided by. a three setpoint PORV, the Decay Heat Removal System.(DHRS) relief valve, administrative control.and operator action.

t

~

,,7 t

s.

Added protection is afforded by the plant design which'orovides for.

~

the presence of a steam bubble or a nitrogen blanket in the pressurizer at al's times.

f i.N The applicant has analy ed draign overpressure..eventis as a function

/

of temperature and compared the maximum pressure produced to the Appendix G limit curve. The Ninparision is reproduced here as

Figure 5.2.2.1.*

y

-The evaluation included t.edvertent actuation of an RFI train, inad-vertent opening of a Core Flood Tank discharge valve, inadvertent discharge of nitrogen to the pressurizer, failure of a makeup control vhlve, actuation of all pressurizer heaters, failure of the Decay Heat Removal System and starting of a reactor coolant pump. The most severeoverpressureconditionatlowtemp$raturewasfoundtobepro-duced by inadvertent HPI operation.Rowever, even for this condition the Appendix G limit

• were calculated not to be exceeded.

For' temperatures between 280*F and 33D*F, the PORV setpoint is set at 550 psig.

Alanas are provided to indicate to the operator improper

~

PORV setting, block valve closure or increasing primary system pressure.

The design flow rate of the PORV at 550 psig is above the maximum single train HP1 flow rate.

At reactor system temperatunts below 330'F, the Appendix G limits could be exceeded in the event of PORV failure without operator action. The applicant has presented the results of an analysis indicating that 10 minutes are available for operator action to terminate the transient in the event of inadvertent HPI operation after actuation of a high pressure alarm.

Additional pro-tection is provided by locking out one of the two HPI trains below a

• Note to LPM - Awaiting approval by MTEB.

g.

r

+

.,

+.

v :

xc i.

y;,

reactor. system' temperature of 330 F..

The computer code used in this.

~

. analysis (DYSID) has'not been previously reviewed by the NRC staff

-g Land we require-a'dditional justification for the use of this code end.

. input assumptions. The 10 minute warning alarm and alarms and inter-locks designed to. prevent improper.PORV and. block valve settings are

. discussed in.section 7 of this SER.

At temperatures below 280*F,.the FORV setpoint is reduced to 360.psig and the isolation valves connecting the decay heat removal system to the primary system are opened. Operatiori of the Decay Heat Removal System (DHRS) provides the. capability to continue cooldown of the

-reactor system and provides additional pressure relief capability by

's means of the DHPS relief valve. The PORV and the DHRS relief valve provide' redundant means.of relieving the additional inventory that would'be added to the primary system in the event of inadvertent HPI operation down to a reactor coolant temperature of 160 F.

At this temperature both trains of HPI are locked out.

The DHR5 relief valve is an ASME Class II seismic component'.

The DHRS relief. valve will be tested for operability as specified in the ASME Ccde Section XI.

c.. Seconde y System Overoressure Protection Overprissure Protection for the secondary system is provided by 8 spring

-loaded safety valves each main sceam line upstream of the Main Steam isolation valve.

The valves are designed to open at staggered setpoints, the lowest ~being the secondary system design pressure of 1050 psig and the

' highest at S: above that value. The combined steam relief capacity of the safety valves in 21% greater than the design main steam flow rate at full 1.

'";' i f,,

3; y

?

6'.

-8

The' applicant has analized the secondary system response to'a complete loss' of loa'd from design overpower-using the POWERjTRAIN computer code oq and has--cetermined that the 110% overp/ressure limit pf the ASME code J

would.~not~be exceeded..Our review of the POWER TRAIN code is discussed' in'S'ection 15 of this SER.

Y Conclusion With the exception of the above noted staff concerns,we conclude that the Reactor System and. secondary system of the Midland plants will be adequately protected from overpressure at both low and operating temperatures and that the design meets the requ_irements of General Design Criterion 15 of Appendix A, 10 CFR 50.

1" s

e e

J b

I

v.

y n-.

mm 3

WI 1

gz 2O 9

f 3l l

op c.w o

>I l

e<

C e

o eo 2-10 le mC t-O 5 E >Cl 2 q H<

z u_I s

<jo Ol o a

CO C O

elo Cl o mw c.

C e n

w c.

H w d CO w

mN ElI E Nl N

F

o. O w

EH! ED oZ D

c C

I mO ml O Xe o

ma ea w g wlg O

5 f

e Eo ElO n"

i 1

^

l oon

\\

l l

o I

e l6 W

lis a

E Ie E

P.

[c ro lo 0

wZ C

a.

l 5

[ c.

N d

w' 2

u O

o C::

G e

E E

~W C

o m3 a

w=

co

<5 o <>

e

'w M

% +5 qaW wzi d

~

!!s

-O C 5D wE h'=

mis R;

>O C

a a Oc OZ G

a<

m58 l

L se' N

o e

r<e o

o o

e i

zs=

I I

I I

o 1

o 8

8 E

o m

o.

o S.

8 i

u n

l

\\

l DISd - S.saJd sod l

g-n?

<~

~

e,.

~f: < - >;

~

3 + ;,. <;._,

~

z

/,

RSB Open' Items

. g, Midland FSAR Section 5.2.2 1). Second Safety Grade Reactor Trip.

./

Showthat.theoverpressurecalculationsarecon'$ev#ative for antici-

')

~

j.

' pated transients assuming failure of the first reactor trip but with g

full credit for both spring ggtuated safety _ valves.

,c h-2); Analytical Models-

'o

~)

Provide a complete description of the' DYSID code used to analize a

inadvertent HPI~ operation.

Includeallequations,assumpiionsand experimental verification.

Provide comparisions between the DYSID r

code and the CADD code which has been : approved by the NRC staff. Discuss the conservatism of the imput assumptions used in the Midland DYSID analysis, in particular those for pressurizer leveh.

b).- The POWER TRAIN code. utilized for overpressure analyses has not been approved by the NRC staff.

An additional information request was sent to B&W April 4, 1979, letter to J. Taylor, E&W frem S. Varga, NRC, " Review of Topical Report EAW-10070.

Provide this rbquested

~

information-with a discussion of applicability to Midland.

.c)

Provide the details of the CADDS verification program using riant

~

.startup data disc'ussed on page'XIV-39 of NUREG-0460.

q 6

I y

i J '

,. ~.., _ _.,,

m-...- _.,.