ML20113G141

Forwards Addl Info on Applicability of GE Rept NEDE-24988-P on TMI Item II.D.1 Re Relief & Safety Valve Test Requirements

ML20113G141
Person / Time
Site:	Clinton
Issue date:	01/16/1985
From:	Spangenberg F ILLINOIS POWER CO.
To:	Schwencer A Office of Nuclear Reactor Regulation
References
TASK-2.D.1, TASK-TM U-0782, U-782, NUDOCS 8501240258
Download: ML20113G141 (11)

Text

f 1

., 3-U-0782 P04-85 01-16 )L 1A.120 ILLINDIS POWER COMPANY IP CLINTON POWER STAtloN, P.o. BOX 678. CLINTON. ILLINOIS 61727 January 16, 1985 Docket No. 50-461 Director of Nuclear Reactor Regulation Attention: Mr. A. Schwencer, Chief Licensing Branch No.2 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Clinton Power Station Unit 1 TMI Action Item II.D.1 Relief and Safety Valve Test Requirements

Dear Mr. Schwencer:

This is in response to your letter dated November 14, 1984 requesting additional information regarding TMI Action Item II.D.1, Relief and Safety Valve Test Requirements, specifically the applicability of GE report NEDE-24988-P, on safety relief valve testing, to the Clinton Power Station. Attached are Illinois Power's responses to your specific questions regarding this matter.

Please contact me should you have any questions regarding this response.

Sincerely y urs,

/

.0 ~

F. A. Sp nge rg j Directo -N lear Cicensing and C nfiguration Nuclear Station Engineering Attachment TLG/em cc: B. L. Siegel, NRC Clinton Licensing Project Manager NRC Resident Office Regional Administrator, Region III, USNRC Illinois Department of Nuclear Safety 2

h 8501240258 850116 /

PDR ADOCK 05000461 A A PDR FQ F l

Ill

P . _ .

=~0 m.

NRC QUESTION 1 The test program utilized a " rams head" discharge pipe configuration.

~

Most-plants utilize a " tee" quencher configuration at the end of the discharge line. Describe t . discharge. pipe configuration used at your plant and compare the anticipated loads on valve internals in the plant configuration to the measured loads in the test program. Discuss the impact of any differences in loads on valve operability.

RESPONSE TO QUESTION 1 The ' Safety / Relief Valve (SRV) discharge piping configuration at Clinton Power Station (CPS) utilizes a "X" quencher at the discharge pipe exit.

The average length of<the 16 SRV discharge lines (SRVDL) is 99' and the submergence length in the suppression pool is approximately 14'. The

-SRV test program utilized a ramshead at the discharge pipe exit, a pipe -

lengthLof: 112' and a submergence length of approximately 13'. Loads on

valve internals during the test program are larger than loads on valve internals in the CPS configuration for the following reasons

1. No dynamic mechanical load. originating at the "X" quencher is E transmitted to the valve in the CPS configuration because there is at least one anchor point between the valve and the "X" quencher.

'2.- The length of-SRV discharge line piping between the SRV and the first-elbow in the test facility was longer than the CPS piping, thereby resulting in a bounding dynamic mechanical load on the

valve in the test. program-due to the larger moment anan between the LSRV and the first elbow.- This length in the-test facility.is 12 ft

.and is a maximum of 11.7 ft in the plant configuration.-

3.  : Dynamic hydraulic' loads _ (backpressure) are. experienced by the valve

-internals in the CPS configuration. The backpressure loads may be either (i) transient backpressures occurring during valve actuation,:or (ii)-steady-state backpressures occurring'during

' steady-state flow following valve actuation.-

(a) The key parameters affecting-the transient backpressures are the fluid pressure = upstream of the valve, the valve opening -;

, time, the fluid inertia in'the submerged SRVDL'and'the SRVDL air volume. . Transient backpressures'increaue with higher upstream pressure, shorter valve opening times, greater line submergence,?and smaller SRVDL air volume.-'The transient backpressure inithaltest program was maximized-by utilizing an orifice plate in-the SRVDL to create a 35-40% backpressure 3- condition on the' valve internals, body bowl and discharge

- flange. This induced backpressure simulated the maximum ,

~ backpressure anticipated in the_ CPS SRVDLs. The maximum

- transient, backpressure occurs wAtla high pressure, steam flow conditions. The transient F3 *preJsure for the alternate E

shutdown cooling mode of 4Waret ton is always much less than the design for steam f;i- ad: tions because of the lower upstream pressure and *ta 2n ,6e valve opening time.

t 1

w -w- w. + +,e e,-+ .- ,-,-w. , - - , , - .m, .- ,

n.i (b) The steady-state backpressure in the test program was maximized by utilizing an orifice' plate in the SRVDL above the water level.and before the ramshead. The orifice was sized to produce a backpressure equal to or greater than that calculated-for any of the CPS SRVDL's.

The~ differences in the line configuration between CPS and the test

- program as discussed above result in the loads on the valve internals for'the test facility which bound the actual CPS loads. An additional consideration.in the selection of the ramshead for the test facility was to allow more direct measurement of the thrust load in the final pipe segment. Utilization of a "X" quencher in the test program would have

- required quencher supports that would unnecessarily obscure accurate measurement of the pipe thrust loads. For the reasons stated above, differences between the SRVDL configurations in CPS and the test facility will not have any adverse effect on SRV operability at CPS relative to the test facility.

k i

2 y

NRC QUESTION 2 The test' configuration utilized no spring hangers as pipe supports.

Plant specific configurations do use spring hangers in conjunction with snubber and rigid supports. Describe the safety relief valve pipe supports used at your_ plant and compare the anticipated loads on valve internals for the plant pipe supports to the measured loads in the test

-program. Describe the impact of any differences in loads on valve operability.

RESPONSE TO QUESTION 2

'The CPS safety-relief valve discharge lines (SRVDL's) are supported by a combination of snubbers, rigid supports, and spring hangers. The locations of snubbers and rigid supports at CPS are such that the location of such supports in the BWR generic test facility is prototypical, i.e., in each case (at CPS and the test facility) there are supports near each change of direction in the pipe routing.

Additionally, each SRVDL at CPS has only 1, 2 or 3 spring hangers, all of which are located in the drywell. The spring hangers, snubbers, and rigid. supports were designed to accommodate combinations of loads resulting from piping dead weight, thermal conditions, seismic and suppression pool hydrodynamic events, and high pressure steam discharge transients.

The dynamic load effects on the piping and supports of the test facility due to the water discharge event (the alternate shutdown' cooling mode) were found to be significantly lower than corresponding loads resulting from the high pressure steam discharge event. As stated in NEDE-24988-P, this finding is considered generic to all BWR's since'the test facility was designed to be prototypical of the features pertinent

_to this issue.

During the water discharge transient there will be significantly lower dynamic loads acting on the snubbers and rigid supports than during the steam discharge transient. This will offset the, increase in'the dead-load on these supports due to the weight.of the water during the r

alternate shutdown cooling mode of operation. Therefore, design f

adequacy of the. snubbers and rigid supports is assured as.they-are

designed:for the larger steam discharge. transient loads.

l-This question addresses the. design adequacy of the spring hangers with

-respect to the weight of the~ water during the liquid discharge

' transient.- Due to the nature of the design of spring hangers there will be-little increased load on'the_ spring hangers because of-the water-loadings.1 The increased loads. are mostly taken'by the adjacent rigid vertical hangers. These rigid hangers are designed for combined loads that include the high pressure. steam discharge transient. These design l- basis combined loads will bound-the combined loading (including increased weight) expected during the liquid. discharge event.

Therefore, it is believed that sufficient margin exists in the CPS L

pipingisystem design to adequately offset the' increased dead load on the spring hangers 11n an unpinned' condition'due to a water filled condition.

.Furthermore, the effect of:the water dead weight load does not affect the ability of SRVs to open to establish the alternate shutdown cooling

~

I  : path'since the loads occur in the SRVDL only after valve opening.

[ 3

s. -

NRC QUESTION 3 The purpose of the test. program was to determine valve performance under conditions anticipated to be encountered in the plants. Describe the

. events and anticipated conditions at the plant for which the valves are _

. required to operate and compare these plant conditions to the conditions ht the' test program. Describe the plant features assumed in the event evaluations used'to scope the test program and compare them to the features at your plant. For example, describe high level trips to prevent water-from entering the steam lines under high pressure

. operating conditions as assumed in the test-event and compare them to

-trips used at your plant.

RESPONSE TO NRC QUESTION 3 The purpose of the SRV test program was to demonstrate that the Safety

. Relief Valve (SRV) will open and reclose under all expected flow Jconditions. The expected valve operating conditions were determined through the use of analyses of accidents and anticipated operational 1 occurrences referenced in Regulatory Guide 1.70, Revision 2. Single failures were applied to these analyses so that the dynamic forces on

-the safety and relief _ valves would be maximized. Test pressures were the highest predict 2d by conventional safety analysis procedures. The LBWR Owners Group, in their enclosure to the September 17, 1980 letter from D. B. Waters to R. H. Vollmer, identified 13 events which may result in liquid or two-phase SRV inlet flow that would maximize the dynamic forces on the safety relief valve. These events were identified by evaluating the initial events described in Regulatory Guide 1.70, Revision'2, with and without the-additional conservatism of a single-active _ component failure or operator error. postulated in the event-sequence. It was concluded from-this evaluation that the alternate

- shutdown cooling mode is the only expected event which will result in J11guid at the valve inlet. . Consequently,'this_was the event simulated in the SRV_ test program. This conclusion and_the. test results applicable to CPS are discussed below.

1 The BWR Owners Group identified 13 events by evaluating the initiating

events described-in Regulatory Guide'1.70, Revision.2, with the additional conservatism of a single active component' failure or operator L

error postulated'in the events. sequence. These events and the

plant-specific features that mitigate these events, are summarized in L

Table 1. Of these=13 events, only 11 are applicable to CPS because of L.

its design and specific plant configuration. Two. events, namely Events if3 and #11'are not applicable to CPS, because Clinton does not have a-

~

- High Pressure Core Injection -(HPCI) system.

p ,

• 4

.L ..

For the 11 remaining events, the CPS specific features, such as trip logic, power supplies, instrument line configuration, alarms and operator actions, have been compared to the base case analysis presented in the BWR Owners Group submittal of September 17, 1980. The comparison has demonstrated that the base case analysis is applicable to CPS because the base case analysis includes plant features which are already present in the CPS design. For these-events, Table 1 shows what CPS

specific features are included in the base case analyses presented in the BWR Owners Group submittal of September 17, 1980. It is seen from Table 1, that most plant features assumed in the event evaluation are also existing features in CPS. All features included in this base case analysis are similar to plant features in the CPS design or do not have a negative effect upon this comparison.

JThe SRV inlet fluid conditions tested in the BWR Owners Group SRV test program, as documented in NEDE-24988-P, are 15* to 50*F subcooled liquid at 20 psig to 250 psig. Event #7, the alternate shutdown cooling mode of operation, is the only expected event which will result in liquid or

-two-phase fluid at the SRV inlet. Consequently, this event was staulated in the BWR SRV test program. At CPS, this event involves flow Lof subcooled water at an RPV pressure from 13.8 psig (Minimum Alternate Shutdown Cooling Pressure) to 130.0 psig (Maximum Alternate Shutdown Cooling; Pressure). SRV actuation / operation with liquid flow under these conditions is acceptable since at lower pressures the associated loads are.less.-

As discussed above, the BWR Owners Group. evaluated transients including single active failures that would maximize the dynamic forces on the safety relief valves. As a result of this evaluat'7n, the alternate shutdown cooling mode is the only expected event involving liquid or two-phase flow. Consequently this event was tested in the BWR SRV test.

program. The fluid conditions and flow conditions tested in the BWR

.0wners Group test program conservatively envelope the CPS plant-specific

' fluid conditions expected for the alternate shutdown cooling mode of operation.

[1 L

i 5

1AllLE 1 - EVLill5 LVALUAIED  !

6 i m

i i 5 . i x u i .. 3 3 3 4

% 3 3 3 h i' .3  ; .% .% .% 38  %

a +

5  ;% e O' M' 0' M c '.7 . h E

%' . 19 MS iS e AW. o S S S : .t $

i 'O  ? a5 JA oA +> l'> * % 8 u'A 4A .25 i C "' m

. *$ $ ,- S ,

- - E3 $ i$ ca$. b E EC EM EU EJ 3YE L JM 40 JU 40E 4%

22 i Ai. AM Ai AA E55 s! SM Si mi SS& S .":

o s m m

. C U C E E E E E E C C C C liigh Water Level 7 Alarm .

liigh Drywell Pressure Alarm X/ X/ X / X/ X/

/S M /S /S /s FW Level 8 Trip .

RCIC Level 8 Trip ilPCS Level 8 Trip IIPCI Level 8 Trip 2 2 2 X X X X X X X X X llPCI/S and RCIC Initiation on l Low Water Level S S NAl S .

S S .S .

S S X X X X X llPCI/S initiation on High Drywell Pressure S-S S NA 1 S X

RCIC Initiation on High Drywell Pressure NA3

. I e4

L.

Lost m m MJ8 J2840 5333 'yS1 EI# x x acaa3 aogeaado ,

• aano g333 7 *$saada0 'VSS ZI# x

~

aanite3 dlJ1 81 I3dH <

'I3dH 'VSS II# x aantted dial 81 sadH

'S3dH 'VES OI#

aanttej dpi 81 3IDH m

'3I38 'V85 6# . x 350 MJ8 1SW B#

algetteneun ucp::ns unop2nys

'Suttoo3 usep2nqS 4tV 1#

o 2 ' ads E pH OIDH'%uatsural 9#

$ aantge3 d iJ1 81 53dH m

'S3dH quassueal S#

=

W

" aani5e3 d!J1 81 3I33 m

'3IDH guatsuaal tr# x

~ ~

aantle3 dtJ1 81 I3dH <

d 'I3dH tuatsueal ry E

m m m "tiej 5a8 ssaJd 2# x x y aani1e3 d p1 81 M3 m m m

• • t!P3 *1uo3 M3 I# x x x c

o n .

U

• e 3 E 3  ;

u i 5 5  ! 5 5 m E e g s c~ s 2

=

3  : .

3. Ho

• M k & h a M du mE t t T*

• m o 5* 8A '

5 a a i 2 2

me wu n o 2 -

O' 03 8 9 2 e 8* 8 8

== ue a

= = a

.5 . ..

S3H01V33 1NVid E* I E E E 5U 5 55 5b 2@

• a a A Et 8"

= ua *15 x* r . a. a at es "e B 5 .3 . .2 2E ao 't U S E E5 E E4 3 ,8  ? & E 5 *E E E~ .

i 8 l i

~

LCSI m m

• MJ8 Jaa^0 S333 'V87 EI# x x x acaa3 aogeaado ,

m

Jano S333 T ssaada0 'V85 ZI# x aanttej d J1  ! 87 I3dH

'I3dH 'V8S II# x aantiv3 d J1  ! 81 S3dH "

'S3dH 'V85 OI# x santpe3 dpal 31 3I33 m

'3IDH 'VES 6# x 350 1J8 1SW 8#

atqetgeAeun uopanS unop2nyS

'Suttoo3 umcpingS ity l.#

e

• ads

• pH 3I38 quassueal 9#

3 santieg dial 87 53dH m 'S3dH 2uassueJ1 S#

=

y aanitej d ial'g1 3I33

'3I38 quatsueal yg aanigej d!J1 81 I3dH

[ 'I3dH quatsueal g gy ,

M .

W tieg Say ssaJd 2# x x x 3

• e aantp3 dial Si ng m og-u a 2uo3 na t#

a:

- o eU

~

a; e eu

$ .5 3 a -

- 2

c e >

A 8 a h  %

5 322 5 h G a = 8'S

% b G S S 3 5 85"

- - = x = = a , ,,q e 5 8 5 5 5 E 3 hhe

. g . . . g 4 #%s S3Hn1V33 1NVid E E 6- 2 E O 5-o OEE M Mu M M M* M ' '

S u u3 u u u u u-3 Se 3 3 30 3 3*

E M We 4 M Ma W M3 s E a a a a a a a u

4. :.

3 Footnotes to Table 1

1. Not Applicable because this initiating event can not occur.

Clinton does not have an HPCI.

L 2. Not Applicable. .An HPCI level 8 trip is not required to make the test results applicable because CPS does not have an HPCI.

3. Not Applicable because CPS does not have RCIC Initiation on High Drywell Pressure. This does not effect the applicability of the test results to the CPS SRVs, because lack of this feature can not cause liquid flow through the SRVs.

9

i NRC QUESTION 4 Describe how the values of. valve Cy 's in-report NEDE-24988-P will be used at your plant. Show that the methodology used in the test program to determine the valve Cy will be consistent with the application at your plant.

RESPONSE TO NRC QUESTION 4 i The flow coefficient, C y, for the Dikkers safety relief valve (SRV) i utilized at CPS was determined in the generic SRV test program (NEDE-24988-P). The average flow coefficient calculated from the test results for the Dikkers SRV is reported in Table 5.2-1 of NEDE-24988-P.

This test value has been used by Illinois Power to confirm that the liquid' discharge flow capacity of the Dikkers SRV's will be sufficient to: remove core decay heat when injecting into the reactor pressure vessel (RPV) in the. alternate shutdown cooling mode. The Cy value determined in the SRV test demonstrates that the CPS SRV's are capable of returning the flow injected by the RHR or core spray pumps to the suppression pool.

If it were necessary for the operator to place CPS in the alternate shutdown cooling mode,,he would assure that adequate core cooling was being provided by monitoring the following parameters: RHR or core spray flow rate, reactor vessel pressure and reactor vessel temperature.

The flow coefficient for the Dikkers SRV reported in NEDE-24988-P was -

determined from the SRV flow rate when the valve inlet was pressurized to approximately 250 psig. .The valve flow rate was measured with the supply line_ flow venturi upstream of the steam chest. The C y for the valve was calculated using the nominal measured pressure differential-between the valve inlet (steam chest) and 3' downstream of the valve and the corresponding measured flowrate. Furthermore, the test conditions and test configuration were representative of CPS conditions for the

alternate' shutdown cooling mode, e.g. pressure upstream of the valve, fluid temperature, friction losses and liquid flowrate. Therefore,'the repo'rted Cy values are appropriate for application to CPS.

5 10