ML18018A900

Response to Mark II Containment Request for Additional Information Dated 02/23/78

ML18018A900
Person / Time
Site:	Nine Mile Point
Issue date:	06/19/1978
From:	Sobon L General Electric Co
To:	Stolz J Office of Nuclear Reactor Regulation
References
NFN-244-78
Download: ML18018A900 (50)

Text

o4 SZ REGULATORY INFORMATION 13ISTRI BUTION SYSTE!'1

< RIDS)

DISTRIBUTION FGR INCOMING MATERIAL TGPREP REC:

STGLZ J F

NRC QRG:

SGBQN L J GEN ELEC DQCDATE: 06/19/78 DATE RCVD: 06/21/ 7 DOCTYPE:

LETTER NQTARIZEI3:

hlO SUB JECT:

RESPONSE

TQ NRC LTR DTD 02/23/78...

"CAGRSQ RELIEF VALVE LOAD TESTS BEHALF OF THE MARK II Gl/NERS GROUP.

COPIES RECEIVED LTR 1

ENCL 10 FORWARDING ADDL INFO GF REPT NEDM-20988>

TEXT PLANT" WHICH WAS SUBt'lITTED BY GE IJfl REVIEWER INITIAL:

X JM DISTRIBUTOR INITIAL:~

DISTRIBUTION OF Tt-tlS MATERIAL IS AS FOLLOWS TOPICAL RPTS 5 CORRESPONDENCE RE MARK II CGNTAIhlMENT (DISTRIBUTION CODE T002)

INTERNAL:

EXTERNAL:

REG FILE~~W/0 ENCL 50-322 FILE44LTR OhlLY 50-353 FILE+>LTR ONLY 50-367 FILE+4LTR OhlLY 5

FILE+<LTR ONLY 0-41 FILE+<LTR ONLY CTURAL ENG BR++W/2 ENCL AD FOR EhlG++N/2 ENCL I PELTIER+~W/ENCL J SNELL~~N/ENCL W KANE++W/ENCL D TIBBETTS++NfEhlCL D VA:-""ALLQ++LTRGl'lLY LNRO3 'CI lIEF~~+LTR ONLY AD FQR PLANT SYSTEMS+-:sLTR ONLY N PIKE~+LTR ONLY LPDR S

PORT JEFFERSON-PDR++W/ENCL POTTSTONN> PA PDR+~Nr ENCL EATAVIA>QH-PDR++W/ENCL CHESTERTON, Ihl-PDR4<N/ENCL OGLESBY IL-PDR++W/ENCL RICHLAND> NA-PDR++I1/ENCL OSWEGO> NY-PDR++W/ENCL TIC~~N/EhlCL NSI C>+W/ENCL

=- ACRS CAT

- ~~~N/'16 ENCL CENTRAL FILE-~~W/2 ENCL 50-"52 FILE+<LTR ONLY 50-358 FILE++LTR ONLY 50-373 F ILE~~LTR ONLY 50-397 FILE+>LTR Ol'lLY NRC PDR~~N/ENCL CGNTAINtiENT SYSTEM ~~W/7 ENCL C ANDERSGN4~<N/EhlCL M D LYNCH++W/ENCL BOURhlIA~~N/ENCL MINER++A/ENCL DEPUTY DIR DPM+<LTR ONLY LWRlt2 CHIEF++LTF? ONLY LNRN1 CHIEF++LTR GhlLY RIJSHBRQGK++LTR ONLY I NR04 CHIEFS'~>LTR GhlLY DISTRIBlJTIGN:

LTR 65 ENCL 47 SIZE:

iP+22P CONTROL NBR 7, 1720193 THE EihlP gc~w~~~~qp~~~~~~

(

)

8 E N E R A L '

E L E O'T R I C GENERAL ELECTRIC COMPANY, 175 CURTNER AVE., SAN JOSE, CALIFORNIA95125 MC 681, (408) 925-3495 NUCLEAR ENERGY PROJECTS DIVISION MFN-244-78 June 19, 1978 U.

S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Hashington, D.

C.

20555 Attention:

Mr. John F. Stolz, Chief Light Hater Reactor Branch No.

1 Division of Project Management Gentlemen:

SUBJECT:

MARK II CONTAINMENT REQUEST FOR ADDITIONAL INFORMATION In response to your February 23, 1978 letter to Dr.

G.

G.

Sherwood, attached are ten (10) copies of additional information requested based on your review of Report NEDM-20988, "Caorso Relief Valve Load Tests-Test Plan," which was submitted to the NRC by GE on behalf of the Mark II Owners Group.

The information provided in the attachment to this letter addresses the February 23 request pIus comments made by the staff at a July 7, 1977 meeting.

The information in the attachment to this letter has'lso been discussed with the staff on several occasions using draft response material as a

basis, for,,that discussion.

Staff comments from those discussions have been incorporated.

This information is being incorporated into Adden-dum 2 to Revision 2 of Report NEDM-20988, which will be resubmitted to the NRC in its entirety in July 1978.

Your cooperation is requested in addressing all future requests for additional information on documents submitted by General Electric on behalf of the Mark II Owners Group directly to me rather than Dr. Sherwood.

Very truly

ours, L. J.

Sobon, Manager BHR Containment Licensing Containment Improvement Programs LJS: mh/1623 7

17ZOa<"-

Attachment cc:

C. J.

Anderson I. A. Pel tier H.

C.

Brinkmann R. L. Tedesco L. S. Gifford File: 3.2.7

+dd2 P

>ilr

~ (:

RESPONSE

TO NRC CGHMENZS OF JULY 7 ~ 1977 AND FEBRUARY 23 g 1978 (3N CK)RSO TEST PLAN NFC Ccxanent 1:

Effects of leaking SRV on quencher loads has been identified as an area of concern.

The Caozso test plan, however, does not addr'ess this concern.

It is our position that tests on a leaking SRV should be conducted.

Therefore, a test plan considering a leaking SRV should be provided for our revs.6wo

RESPONSE

Optional tests & measure the effects of leaking SRV an quencher loads have been incorporated into the Caorso test matrix as tests 41 through 44.

During the normal course of the Caorso tests, these tests willbe 4

performed as soon as it is deteztnined that valve A is leaking.

This detezminatian willbe made on the criteria that if temperature sensor T21 does not return to within 10'F of its pre-test reading, the valve is leaking.

Four discharge tests of valve A willbe performed under this canditian with all instrumentation channels recording data.

Following the cxmpletion of these tests, valve A willbe repaired or replaced and the 1

normal course of testing continued.

While the suggestion to implement controlled leaking SRV tests at Caorso is valid, this plan was not inplemented for two major reasons.

a)

Major in-plant hardware modifications would be required to implenent cantzolled lacing valve tests.

Such tests would require major site work to provide a source of controlled steam flow fram either an auxiliazy boiler or fran the reactor

~through rmdifications to the SRV and its piping.

Because the

'lant is not readily ananable to test operations, pa~cularly

.. fran the utilitypoint of vs, such an approach was not feasible.

Additionally, the difficultyin specifying a leak ratio, or series of leak rates such that a true leaky valve could be simulated, prohibited any attend to desi@ a deliberate leaky

NRC ('.cmnent 1 Page 2

valve mndition into the test plan.

Alternate approaches, as specified below, would be expected to give satisfactory, ifnot

delete, answers to the leaky valve question without impacting the plant hanhrue.

b)

As an alternative to mntzolled leaky valve tests, an optional leaky valve test series, as previously described, has been inm~rated into the Caorso test matrix.

Finally, because the priznax~ concern relative to the first actuation of a leaky valve is its semnd pop like behavior, sufficient data regarding leaky valves willbe obtained fram the secand pop tests that will be mnducted.

KC Cawnent 2:

Dha Caorso test matrix indicates that anst of the parameters of interest wiU. be repeated only once.

Tests performed either for ramshead (Quad cities and Wnticello) or for quencher (NEDE-11314-08) extdbited a great degree of data scatter.

Therefore, we believe that the current Caorso test matrix is insufficient to determine the repeatability of the test data.

We reocmmend additional tests be conducted for first actuation of an SRV and subseq~t actuatians of the sam SRV to denunstrate repeat-ability.

The nurrber of tests should result in test data with statistical significance.

.HESPCNSE:

Extensive changes have been made to the Caorso test matrix fram the version transmitted to the NRC in revision 2 of the Caorso Test Plan (NEDDY-20988) dated Demnber 1976.

Attachment (1) shmrs the current test matrix as it appears in addendum 2 to revision 2 of the test plan issued in April 1978,.

In the developement of the test matrix, the followiizg criteria has been the basis for the selection of tests to be conducted and the rxnrber of repetiticns.

1)

Tests which acccmplish the most inportant cbjectives of the testing program are repeated most often in order to reduce the uncert unity in the major effects.

These primary objectives would include confirmation of the follcwing:

a.

Bubble pressures for first and subsequent actuations.

b.

Air clearing bubble pressure

~t.te.equation with distance.

c.

Discharge line pressure and water level transient.

To aco:mplish these objectives, 16 single valve first actuations, sefjen sequences of 5 actuations each (4 subsequent act~tions per seymnce) and seven single valve actuations at reactor pressures ranging frcxn 50 to 1000 psia NU. be conducted.

Qhese tests wi11 not only establish the data base for the primary objectives but will also test for the influence on bubble pressure of steam flowrate, which has been established as izqmrtant influence paraneter in

2) Tests conducted for the purpose of m asuring the influence of variables which previous testing has shown to have only minor effects are conducted in numbers sufficient to establish this trend.

Mesc include valve open tine, vacuum breaker size and quencher location.

Should the effect of variations in these parameters be greater than is currently believed, this trerg can also be established with the current nmrber of tests planned.

While it is believed that the test matrix for the single valve tests is sufficient to acccmplish the objectives described above, provision is established for perfoxmance of retests should it becane app-mnt that repetitions of certain tests are'required

=(see item 23 of attach-ment (1)

)

During the 4/6/78 meetings with N. Su (NRC), the mncern was expressed that the nether of multiple valve tests was insufficient for the purposes of verifying design basis nethoch.

While it is believed that the cbserved loads for the multiple pop cases willbe sufficiently below design values to make aare than the previous nurrber of multiple valve tests, unnecessary, an additional four valve test (see attach-ment (1) test g30) has been incorporated into -the test matrix.

The~ore, the tcrtal nunber of planned multiple valve tests currently stands at two tm) mlve tests, one three valve,,test, five four valve tests and one eight valve test.

Further, a provision has been incorporated into the,test plan for up to five additional four valve tests should the cbsexved pressure loads and data scatter be grea".~-'

i, than expected (see attachment (1), test 445).

R

KC Camant 3:

Based an our evaluatian of previous SRV test data, we find that the SRV discharge time and duration between first actuation and sWsecpoent actuation influence the quencher load due to subsecpmnt actuation.

'Ihere.-

fore, provide the value and t4e basis for the selection of the tims for the Caorso tests.

RESPONSE

'Ihe SRV discharge times for the sequential pap tests (see attachment (1),

tests 5, 6, ll, 12, 13, 14) were selected to rreasure the effect. of.SRV dis-charge pipe t~erature on quencher loads.,This will.be aaxmplished by increnentally heating the pipe with 5 semnd sequential disch-mpes.

Based on the Mnticello ramshead test results, the last pop (and probably the next to last pap) on each sequential actuation test willbe at naximum pipe temperature.

Further, to determine if changes in the sub-sequent pop loads are due to taqmrature effects only, test 22, which naw starts with a 20 second blcwdawn and is followed by 5 secxzxi consecutive

pops, has been incorparated into the test matrix.

'Ihis serves the purpose of heating the discharge pipe to ma~nun pipe tesrperature for the semnd pop.

Ccmparisan of secand pop loads for test 22 with semnd pop loads for tests 5, 6, ll, 12, 13, 14, will allow detenihinatian of whether changes in subsecpmat pcp loads are due anly to pipe terperature effects (at a given water level).

Me effect of valve. closed times between actuatians willbe tested by c1osing the valve for 60 secands between actuatians in tests 5 and 11, and far about 10 seconds in tests 6 and 12.

In additian, per the informal discussions with N. Su of 4/6/78, tests with rv 2 seconds valve c1osed tines between consecutive pops has been incca~rated into the test matrix (see tests 13 and 14 of attadment (1)

).

The purpase of this test series is to obtain subsequent pop loads at maximum refload water level.

Priar to this, @eries, the data fnxn previous tests willbe reviewed ta determine the time of maximum water reflood.

'Valve closed times of less than 2 sec-onds are impracticable because of valve response limitations to apenning and closing signals.

(Valve respanse time fram open signal to full valve opening is ~'.2 seconds; response tizre frcm close signal to full va1ve closure ranges fram 1 to 2 seconds.)

NRC Qcrrment 4:

Page 4-6 states that a complete understanding of the subsequent actua-tion effect requires data on pool temperature in the vicinity of the

quencher, pipe tergerature and pressure following valve closure, flew.

rate of air through the vacuum breaker and dynamics of back flew of water.

Ne agree that the air t'enperature history inside the pipe could be important.

He+ever, insufficient information has been given in the test plan regarding the measurement of air temperature in the pipe.

Clarify what measurements or calculations willbe made to rronitor this temperature.

Response

The ta~erature history.of the saturated airsteam mixture inside the 1

SRV discharge line willbe measured using sensors Tl, T2, T3, and T27 (See attac?omnt (2) ), These sensors are designed to measure the ~esture of the environment in the discharge line, not the temperature of the discharge ~e wall.

'Jhe temperature of the saturated airsteam mixture can be verified using the rrethods described belch.

For the first actuation, the saturated air steam mixture tettperature can be apprcocizrated as being equal to the tanperature of the disdmrge pi~.

inner wall.

Gris temperature willbe measured on sensors T24, T25, and "T26 (see attachirant (2) ).

For sequential actuatians, the air mass will be measured as it enters the discharge line through the vacuum breaker.

Given this measured air mass,"the Kischarge line pressure and the total discharge lira ~-steam mixture volume at the end of the water reflood transient, the dmperature of the gas mixture can be calculated using 4

the ideal gas laws by assuming that the steam in the mixture is at saturation corresponding to the discbaype line pressure.

NK Ocsment 5:

The sensor failure rate was found to be quite high in the Wnticello Plant ramshead test program.

Sensors of'he sane manufacturer mxhl used in the Hanticello test. will also be used in the Caarso test.

In light of this experience, we believe that redundant instrumentation is needed in critical areas.

For instance, 'redundant sensars should be provided in the following locations:

a.

The vicinity of Quencher A and the place by which cmbined loads fram multiple SRV's actuation willbe determined.

b.

SRV line between elevation 51.612 and 45.770.

In addition, level

'probes should be added between Ll and L21.

~Res anse:

~

(a) Slight rearrangement of instrun~tation in the suppression pool was abolished since the issue of revisian 2 to the Caorso test plan'NEDDY-20988) dated Decerber 1976.

Specifically, pressure transducers P36 and P37 were shifted fran their original locations near support column 8 to the positions shown an attachrent (3).

While this shift does not represent a direct backup for sensors P23 and P38, data frcm these sensors willbe sufficient to allow evaluation of multiple valve pop loads should either P23 ar P38'fail.

At the same tirre, should sensors P23 and P38 renain intact, sensors P36 and P38 will provide additional useful data which would;rat.be exact duplicates of readings fram P23 and P38.

(b) The spacing of level prtMs Ll and L12 is cansidered to be of minor importance for the follcaring reasons:

(1)

License data and GE rtedels indicate that the water column will not rise to sensor Ll during the natonal vacua breaker tests.

(2)'Ihe large nuNer of level prcbes bedim sensors Ll and L7 aller an evaluation of the velocity and acceleratian of the water column; this information, in turn, can be used ta determine approximately the maxunum level to which the water colunn rises during the reflacd transient.

NBC Cmnent 5

Page 2

(3)

Tenperature sensor T3 (see attadment (2) ) can be used to provide additional water level information between sensors Ll and L12 if recpu.red.

(4)

Qm pressure sensors P7 and PS (see attachrrent(5)

)

(0-25 psia) in the SRV line can also be used to measure water level. 'Ibis was done in the recent.'bhnticello bencher tests where the water level sensors and the inferred water level fran the lcm pressure readings cxxqared favorably..

With regard to the question of redunduncy for level prcbes betNeen elevations 51.612 and 45.770, Mngerature sensors T3, T4, T5, T27, and T28 can function as level sensors ifrequired and willprovide redunduncy for the level probes in case of failure.

NRC Ccmnent 6:

Submerged structure loads have been identified as a prim-~ design load for the Mark II containnent.

We believe that the analytical program indicated in the Mark II Owners Group naeting which was held on Febnmry 16 and 17, 1977, is insufficient to support the design loads for, suhnerged structure without expex~ntal data.

Therefore, we renx~mend that additional pressure sensors should be installed on support colmns and downoomers to measure the drag load during SRV operation.

~Bes nse:

Instrumentation which has,been installed at Caorso is expected to provide experimental data to s~rt the analytical quencher nadels relative to su1merged structure load caused by SRV actuations.

Measurements on two downcxxners and a support column adjacent to Quencher A willprovide the needed loads information.

The instrum ntation used, which is shown in attachnents 6,

7 and 8 consists of the follcwing:

1.

Strain gauges on vents Nos.

1 and 9.

2.

Pressure transducers on Support Column No.

7 and Vent No. 9.

GE believes the information obtained on pressure

loadings, supplemented by the strain gauge data, willprovide sufficient information to show that

'the analytical mxhls predict reasonably canservative~idads.

The in-plant, test data itself is not intended to provide a loading basis for suhrerged structure loads that can be directly or indirectly applied to define loads for other Mark II plants.

'?he tests are planned as a confixmatozy test to show the conservatism of the analytical nodels, which muld then be used to define loads for other structuvas (not instrurrented in Caorso) and for stnj~es in other Mark II plants..

S

NEC Qcrrtnent 7:

Provide the locations for pressure sensors Nos. 19, 23, 35, 36, and 37.

Be~e:

The locations of these sensors are sham in attachment (3).

Test 8

0 1

2 3

4 SVA SVA SVA SVA A

A A

A A

Test Type 1

'3~e1ve TABLE 6.1 TEST MATRIX Discharge Time sec.

5 5

5 2

10" VB 10" VB 10" VB 10" VB 10" VB CP$ NWL, CP, SP'

NWL, NWL, NWL CP, CP$

CP Initial Pi e Conditions 3

Valve Closed Time CVA sec Pi e Coolin tlrs.

>2

>2 501 502 503 504 505 601 602 603 604 605 SVA CVA CVA CVA CVA SVA CVA CVA CVA CVA SVA SVA A

A A

A A

A A

A A

A A

CP, WP, WP, HP, HP CP, WP, WP, HP, Hpi CP, CP

NWL, TWL, TWL$

TWL, NWL,

TWL, TWL,

TWL, TWL, 10 VB 10" VB 10" VB 10" VB 10" VB 10 VB 10" VB 10" VB 10" VB 10" VB Re uce Reduced VB VB 5

60 60 60 60 10 4 10(4) 10(4) 10(4)

>2

>2 2

>2 C7 Pl Pl O K U M 3; CA O ~

tO D O CO PO CO 9

10 1101

+02 1103 1104 1105 1201 1202 1203 1204 1205 1301 1302 1303 1304 1305 SVA SVA SVA CVA CVA CVA CVA SVA CVA CVA CVA CVA SVA CVA CVA CVA CVA A

A A

A A

A A

A A

A A

A A

A A

A Cpi CP CP, WP, WP ~

HP, HP CP ~

WP, WP ~

HP, HP CP ~

WP, WP, Hpi HP,

NWL, TWL,

TWL, TWL, TWL

NWL, TWL,

TWL, TWL$

TWL

NWL, TWL,

TWL, TWL,

TWL, Reduced VB (6)

Reduced VB 6

Reduced VB (7)

Reduced VB (7)

Reduced VB (7)

Reduced VB (7)

Reduced VB (7)

Reduced VB (7)

Reduced VB (7)

Reduced VB (7)

Reduced VB (7)

Reduced VB 7) 10 VB 10" VB 10" VB 10" VB 10" VB 5

5 5

5 5

60 60 60 60 10(4) 10(4) 10(4) 10(4) 2 (15) 2 (15) 2 (15)

-2 (15) 15

>2

>2

>2

>2

>2 (i)

Test Type Test

() (1) (2) (10) (13) Velve Initial Pi e Conditions (3)

Discharge Time sec.

Valve Closed Time CVA Sec Pi e Coolin

.Hrs.

1401 1402 1403 1404 1405 j SVA CVg CVA CVA VA A

A A

A A

CP,

NWL, WP ~ TWL~

WP ~ TWL/

HP ~ TWLt Hp TWL 10" VB 10" VB 10" VB 10'3 VB 10" VB 2 (15) 2 (15) 2 (15) 2 (15)

>2 15

.SVA F

CPt NWL/ 10" VB 16 SVA F

CP NWL 10" VB

>2

>2 17 18 19 20 2201 2202 2203 2204 SVA E

CP, NWL, 10" VB SVA E

CP NWL 10" VB SVA U

CP ~ NWL, 10" VB SVA U

CP NWL 10" VB SVA A

CP NWL 10" VB CP NWL 10" VB HP, TWL, 10" VB HP~ TWLt 10" VB HP TWL 10 VB SVA CVA CVA CVA CVA Retest of Tests 1 Throu h as Re uired 20 2o(s) 5 5

5 5

lo(4) lo(4) lo (4) lo (4)

>2

>2

>2

>2

>,2

>2 Pl (Tl D < D fTl C/l D

e-)

h)

O O

X 20 CO Co 24

?7 28 29 rLVA A,F A~F~E~U A~F~E~U A~F~E~U A F E U cp, NwL. lo" vB NW 1

VB VB CP NWL 10" VB CP NWL 10" VB 5,1O(9) 5 10 9 5 10 15 9 5,10,15,20(9) 5 ilo,l5,20 (9) 5,10,15,20(9) 5 10 15 20 (9)

>2

>2

>2

>2

>2

>2

>2 31 32 B C DL CP, NWL, 10" VB AtBtD~H CP f NWLf 10" VB K ri 5,10,15,20(9) 5 glotl5t20g25/

30 35 40 9

>2

>2

0

Test Type Test N(1) (2) (10) (13)

Valve 33 30 pAaTReactcr a

Pres (10)

SVA 34 100 psia Reactor A

Pres (10)

SVA 35 200 psia Reactor A

Pres (10)

SVA 36 400 psia Reactor

.A Pres (10)

SVA 37 600 psia Reactor A

Pres (10)

SVA 38 800 psia Reactor A

Pres (10)

SVA SVA Initial Pi e Conditions (3)

CP p NWL3 10" VB CP 3

%'TLp 10" VB CP,,

10"

<, 2 2

2

> 2

> 2 Optional Optional 0 tional Valve Closed Time CVA,Sec Pi e Coolin Hrs.

2 a

rn m

U C/l O ~

h)

O O

X to 00 M

CO 4401 SVA (13) 4402 CVA 4403 CVA 4404 CVA

>4b LV, NWL, 10" LVt TWL3 10 LV, TWL, 10" LVp TWL, 10" r.v Twr.

10" A

A A

A 45 Retests of Tests 24 Throu h 41 as Re uired lo (4) lo (4) lo (4) lo (4) lo (4)

> 2 Optional Optional Optional Optional Optional

NEDM 20988 REVISION 2 ADDENDUM 2 Notes:

(1)

(2)

(3)

(4)

SVA = Single valve actuation CVA = Consecutive valve 'actuation Reactor power level to be as follows at the beginninq of each test:

Reactor Power Level water leg transient is to be based on water leg transient data from prior Test No.

1 through 22, 4lthrough 44 25-100%

24 and 25 30-95%

26 40-90%

27,28,29,30,31 50-85%

32 50-85%

0, 33'throuah-39

-Determined at site 40 50-100% power CP

=

Cool pipe WP

=

Warm pipe HP

=

Hot, pipe NWL =

Normal water level TWL =

Transient water level VB

=

Vacuum breaker size (Only one VB in use on line A, both VB in use on Subse uent Actuation - Transient Water Level all other lines)

Predetermined valve closed time to be the minimum time reauired to assure that the water leg has returned to a steady water level with oscillations of less than

+-

1 foot about this steady value.

Evaluation of the predicted test runs.

(5)

Vacuum breaker butterfly valve at 25-29'90's full open), but at same settin (6)

Vacuum breaker butterfly valve at 9-14', but same setting for LhtII NsB.

(7)

Test Nos.

11 and 12 are to be run with the smallest vacuum breaker size tested in tests 7 to,10 which resulted, in the highest, water level overshoot in the i>s.

SRV line of lesst."sa..'.

';'Ok ft. (4,6 M).

(8)

Predetermined valve open time to be the minimum time required to assure that the discharge line has reached a steady state temperature.

This time will

. be. the, time at which both T21 and T23 are constant to within 2'F (1'C)

'ise per second in Test 21 I

(9)

The valves are closed sequentially to avoid the possibility of a scram.

(10)

Initial pressure at safety/relief valves to be 950 psia (66.8 Kg/cm2) to 1000 psia (70.4 Kg/cm2) except during Test Nos.

33 throuqh 39.

Initial -.

tdhperature in the suppression pool at the start of each test will be 88 to 85'F (26.7 to 29.4'C) for all multiple valve tests and 80-90'F

~(26.7 - 32.3'C) for all singl'e valve tests except Test No.40 (see Note 11).

6-3B

Notes:

(Cont'd)

NEDM 20988 REVISION 2 ADDENDUM 2 0

(11)

Prior to the test, the pool shall be cooled to within 5'F of service water temperature or for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, whichever is shorter.

(12)

Extended SRV blowdown to continue until the hiqhest reading from the in-plant monitoring system reaches 101.5'F (38.6'C) or for the predicted time for the bulk pool temperature to reach 101.5'F, whichever is shorter.

(13)

Optional Leaking SRY test Tests 41 through 44 will be performed if it is determined that SRV "A" is leaking.

b)

SRV "A" will be determined to be leakina if pipe temperature sensor T21 does not return to within 10'F of its pre-test reading (5.6'C) within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

(14) It is desirable to obtain a record of any unscheduled conditions (such as loss of site power and containment isolation) which results in relief valve discharges which h'eat the suppression pool to temperature above 48.9'C (120'F).

All instrumentation which is utilized for these tests (see section 6.1) should be actuated and the data recorded continuously at all times that the pool temperatures exceed 48.9'C (120'F) and a

relief valve is open.

(15)

Tests 13 and 14 to be conducted only if it has been determined from tests 1 through 4 that the maximum water level overshoot in the discharge line is ( 15.0 FT

( 4.6 M) from SRV closure.

(16)

This note applies to maximum positive and maximum negative bubble pressures.

If one or more of the four four-valve tests exceeds the mean plus one standard deviation from all the SYAs, run two additional four-valve tests.

If these six four-valve tests have an average pressure

.exceeding the average four-valve predicto.d ppepsure from the empirical

'"model, OR if the maximum pressure from the s'ix four-valve tests exceeds the predicted value plus two-thirds of the difference of the 90-90 design value and the predicted value from the empirical model, run three additional four-valve tests, for a total of nine such tests.

(Maximum pressures at the highest real-time pool boundary pressure sensors are to be used as bubble pressures.

The empirical model is to be evaluated at actual Zaorso test conditions.)

6-3C

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-20988 REYISION 2 F?6.

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9 T

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Causes 400 Im belov braciag hub

-',bootes Pour unisexual gauchos to be located at 90 intervals around circumference on 2 selected dovacomer rents (see Fig. 5 <<A for location).

Pairs o. "iametrically opposed strain gauges to be connec.ed to read moments.

Orientation of gauges relative to containment axes to be recorded, S.C. Mos.51,52,53, 54 on Vent No.

1 S

C 'Sos 55 '6,57'8 on Vent No 9

5-21

NE~ 20988 REYl S lON 2

PI~. 5-5" %PPL bY= S"=:i C.

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41 P

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Kl, 40,700 Notes Preaeure aeneora to be oriented alaag radial line to quencher "4" 5-17

(,

(l e> +

~ '

NIAGARAMOHAWKPOWER CORPORATION NIAGARA '

MOHAWK 300 ERIE BOULEVARD. WEST SYRACUSE. N. Y. I3202 April 28; 1978 Office of Inspection and Enforcement Region I'ttn:Mr. R. T. Garison, Chief Reactor Construction and Engineering Support Branch U.

S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pennsylvania 19406

Dear Mr. Garison:

'Re:

Nine Mile Poigt-U 2

Docket NOI. 50-410

'n November 1, 1977 you were notified of our continuing effort to evaluate the criteria for stress relief design for a safety related tunnel wall for Nine Mile Point Unit 2.

The rock stress evaluation which is described in the report entitled "Geologic Investigation for Nine Mile Point Unit 2" is now complete and provides much of the basic data required for this evaluation.

My April 28, 1978 letter to Mr. Case transmits this report.

The specific design requirements for stress relief are now being resolved.

Design information will be provided in July, 1978.

Very truly yours, NIAGARA MOHAWK POWER CORPORATION N/

.+~X"(v~.1 +)y Yr(4

'Gerald~K.

Rhode, Vice President System Proj.ect Management