ML18191A205

Forwards Response to NRC Comments of July 7, 1977 and February 23, 1978 on Caorso Test Plan

ML18191A205
Person / Time
Site:	Columbia
Issue date:	06/19/2018
From:	Sobon L General Electric Co
To:	Stolz J Office of Nuclear Reactor Regulation
References
NFN-244-78
Download: ML18191A205 (29)

Text

REGULATORY INFORMATION DISTRIBUTION SYSTEM <RIDS)

DISTRIBUTION FQR INCOMING MATERIAL TGPREP REC: STQLZ J F ORG: SGBQN L J DGCDATE: 06/19/7 NRC GEN ELEC DATE RCVD: 06/2if78 DGCTYPE: LETTER NOTARI ZED: NO COPIES RECEIVED SIJB JECT: LTR L ENCL 10 RESPONSE TG NRC LTR DTD 02/23/78... FORWARDING ADDL INFO OF REPT NEDM-20988>

"CAOliSO RELIEF VALVE LOAD TESTS TEXT PLANT" WHICH WAS SUBMITTED BY CE GIII BEHALF GF THE MARK II GIJNERS GROUP.

D1 STR IBUT IOhl QF Tl.}1S MATERIAL h

IS REVIEWER 1NIT IAL:

DISTRIBUTOR AS FOLLOWS ItlITIAL:

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TOPICAL RPTS 8c CORRESPONDENCE RE MARK II CGNTAIhlMENT (1)ISTRIBUTIGN CODE T002)

INTERNAL: REG FILE<<<<W/0 ENCL CENTRAL FILE<<<<W/2 ENCL 50-322 FILE<<<<LTR ONLY 50-352 FILE<<<<LTR ONLY 50-353 F1LE<<<<LTR ONLY 50-:.58 FILE<<<<LTP. Ql'lLY 50-367 FILE<<<<LTR GhlLY FILE<<<<LTR ONLY 50-374 FILE<<<<LTR ONLY FILE~<<LTR ONLY 50-410 FIL E<<<<LTR ONLY NRC PDR<<<<W/ENCL STRUCTURAL LNG BR<<<<W/2 ENCL CQNTAINI'IENT SYSTEMS<<<<W/7 EI'lCL AD FGR ENG<<<<W/2 ENCL C AhlDERSON<<<<W/EhICL I PEI TIER<<<<W/ENCL M D LYNCH<<<<W/ENCL J SNELL<<<<W/ENCL A BQURhl 1 A<<<<WfENCL W KANF<<<<W/ENCL S MINER<<<<W/ENCL D TI BBETTS<<<<W/EhICL DEPUTY DIR DPM<<<<LTR ONI Y D VASSALLG<<<<LTR ONLY LWR52 CHIEF<<<<LTP, ONLY LWR53 CIIIEF<<<<LTR ONLY , LWRNL CHIEF<<<<LTR ONLY AD FOR PLANT SYSTEMS<<<<LTR ONLY M RUSHBRGGK<<<<LTR ONLY W PIKE<<<<LTR GhlLY LWRN4 CHIEF<<<<LTR ONLY ih i l,,%

EXTERNAL: LPDR S PORT JEFFERSON-PDR<<<<W/ENCL PGTTSTQWN. PA-PDR<<<<W/ENCL BATAVIAiGH-PDR<<<<W/ENCL CHESTERTON> II l PDR<<<<'W/ENCL GGLESBYi IL-PDR<<<<W/ENCL RICHLAND WA-PDR<<<<IJ/ENCL GSWEGGi NY-PDR<<<<W/ENCL T I C<<<<W/ Et ICL Ulna..

NS I C<<<<W/ENCL hype

- - ACRS CAT <<<<W/16 ENCL DISTRIBUTION: LTR b5 ENCL 47 CONTROL NBR: 781 7~x01 ci3 SIZE: iP+22P

<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<<< THE END

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PROJECTS OIVISIO~

GENERAL ELECTRIC COMPANY, 175 CURTNER AVE., SAN JOSE, CALIFORNIA 95125 MFN-244-78 MC 681, (408) 925-3495 June 19, 1978 U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, 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 plus comments made by the staff at a July 7, 1977 meeting.

The information in .the attachment,to,this letter has also 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 tu'evision. 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 ipformation 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 g7~01; BWR Containment Licensing Containment Improvement Programs LJS: mh/1623 Attachment cc: C. J. Anderson H. C. Brinkmann f L. S. Gi ford I. A. Peltier R. L. Tedesco File: 3.2.7

RESPCNSE TO NRC CGHMNZS QF QUIZ 7 J 1977 AND FEBRUARY 23 ~ 1978 (3N CM)RSO TEST PLAN NFC Chant 1:

Effects of leaking SRV on quencher loads has been identified as an area of cancexn. '%he Caorso test plan, hemmer, does not address this concern.

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

Therefore, a test plan mnsidering a leaking SRV should be provided for our revi6w RESPCNSE:

Optional tests bo measure the effects of leaking SRV on quencher loads have been incorporated into the Caorso test matrix as tests 41 through

44. Duz~g the normal course of the Caorso tests, these tests will be performed as soon as it is determined that valve A is leaking. This determination will be 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 will be performed under this condition with all instxenentation channels recording data. Following the ccmpletion of these tests, valve A will be repaired or replaced "and the normal course of testing continued.

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

a) Major in-plant h u'decare modifications would be required to

~ ~, *, implenent controlled leaking valve tests. Such tests would recpu.ze major site work to provide a source of controlle'd steam flnr fran either an auxiliary boiler or fran the reactor

-through modifications to the SRV and its piping. Because the

'plant is not readily anenable to test operations, particularly

.. fram the utility point of view, such an approach was not feasible.

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

NBC Cattnent 1 Page 2 valve condition into the test plan. Alternate approaches, as specified belch, would be expected to give satisfactory, if not complete, answers to the leaky valve'question without impacting .

the plant hardware.

b) As an alternative to controlled leaky valve tests, an optiona1 leds valve test series, as previously described, has been incorporated into the Caorso test matrix. Finally, because the px~y concern relative to the first. actuation of a leaky valve is its second pop like behavior, sufficient data regarding leaky valves will be obtained fram the second pop tests that will be conducted.

NEC Cament 2:

aha Caorso test matrix indicates that most of the parameters of interest will be repeated only once. ~ts performed either for ramshead (Quad cities and Nonticello) or for quencher (NEDE-11314-08) exhibited 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.'e reaxmend additional tests be conducted for first actuation of an SRV and subset actuatians of the sate SRV to dernnstrate repeat-ability. The nunber of tests should result in test data with statistical significance.

-.RESPCNSE:

Bctensive changes have been made to the Caorso test matrix fram the version trananitted to the NRC in revision 2 of the Caorso Test Plan (NEZEH-20988) dated December 1976. Attachnent (1) shcws the current test matrix as it appears in addendum 2 to revision 2 of the test plan issued in April 1978..

In tEe development of the test matrix, the following criteria has been the basis for the selection of tests to be conducted and the nurber of repetitions.

1) Tests which accarplish the nest irqmrtant cbjectives of the testing program are repeated met often in order to reduce the uncert-~ty in the major effects. These prUnary objectives would include confirmation of the follcwing:

a. Bubble pressures for first and subsepmnt actuations.

b. Air clearing bubble pressure attenuation with distance.

c. Discharge line pressure and water level transient.

To accanplish these objectives, 16 single valve first actuatians, se~ sequences of 5 actuations each (4 subsequent act~tions per sapxmce) and seven single valve actuations at reactor pressures ranging frcxn 50 to 1000 psia m.ll be canducted. These tests will not anly establish the data base for the primary objectives but will also test for the influence on bubble pressure of steam flawrate, which has been established as irqmrtant influent paraneter in

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

'Ihese include valve open tire, vacuum breaker size and quencher location. Should the effect of variations in these parameters be greater than is currently believed, this trend can also be estab'.shed with the az~nt nurser of tests pI-oned.

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 performance of retests should it becane apparent 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 concern was expressed that the nunber of multiple valve tests was insufficient for the purposes of verifying design basis nathcds. While that the cbserved loads for the multiple pop ~ it is believed will be sufficiently below design values to make rmre than the previous nmkmr of multiple valve tests unnecessary, an adc1Ltiana1 four valve test (see attach-ment (1) test 530) has been incorporated into the test matrix.

Therefore, the total nmrber 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 additicnal four valve tests should the cbserved pressure loads and data scatter be greater than expected (see attachment (1), test 545) .

NK Camrent 3:

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

fore, provide the value and +e basis for the selection of the tines for the Caorso tests.

RIZPONSE:

We SRV discharge times for the sequential pop tests (see attachment (1),

tests 5, 6, ll, 12, 13, 14) were selected to neasure the effect. of.SRV dis-charge pipe temperature on quencher, loads. This will be acxxaplished by incrementally heating the pipe with 5 second sequential discharges.

Based on the Knticello ramshead test remits, the last pop (and prabably the next to last pop) on each sequential actuation test will be at maxirnurn pipe temperature. Further, to determine if changes in the sub-sequent pap loads are due to temperature effects only, test. 22, which new starts with a 20 second blcwdcam and is followed by 5 second cansecutive pops, has been incorImrated into the test matrix. 'Ihis serves the purpose of heating the discharge pipe to maximum pipe temperature for the second pop. Ccmparisan of secand pop loads for test 22 with second pop loads for tests 5, 6, ll, 12, 13, 14, will aller determination of whether changes in s~eqrmnt pep broads are due anly to pipe tenperature effects (at a given water level) .

, We effect of valve closed times betveml'~rations will be tested by closing the valve for 60 secands between actuatians in tests 5 and U.,

and for about 10 seconds in tests 6 and 12. Zn additian, per the informal discussions with N. Su of 4/6/78, tests with ro '2 ecands valve closed times between consecutive pops has been incorImrated into the test matrix (see tests 13 and 14 of attachment (1) ) . The purpose of this test series is to cbtaij subseqrmnt pep broads at maximum reload water level. Prior to this ~es, the data fran previous tests will be reviewed -to determine the time of maxUnum water reflood. 'Valve c1osed tixres of less than 2 sec-ands are impracticable because of valve respanse 1imitatians to openning and c1osing signals. (Valve response tine fram open signal to full valve r

opening is ~'.2 seconds; respanse time frcm close signal to full valve closure ranges fran 1 to 2 seconds.)

NRC Qxment 4:

Page 4-6 states that a cxxaplete understanding of the subsequent actua-tion effect requires data on pool temperature in the vicinity of the quencher, pipe tarparature and pressure following valve closure, flow.

rate of air through the vacuum breaker and dynanics of back flew of water. We agree that the air t'emperature history inside the pipe could be important. Hcwever, insufficient information has been given in the test plan regarding the measurement of air temperature in the pipe.

Clarify what measurements or calculations will be made to nanitor this temperature.

Res~>nse:

The temperature history. of the saturated airsteam mixture inside the SRV discharge line will be measured using sensors Tl, T2, T3, and T27 (See attachment (2) ). Cise sensors are designed to measure the temperature of the environment in the discharge line, not the taqperature of the discharge lme wall. 'Ihe temperature of the saturated airmteam mixture can be verified using the methods described belch.

For the first, actuation, the saturated air steam mixture tetrperature can be approximated as being equal to the temperature of the discharge pipe.

irnez ~Tall... This teaperature will be measured on sensors T24, T25, and T26 (see attachm nt (2) ) . For sequential actuations, the air mass will be measured as ~ enters the discharge line through the vacuum breaker.

Given this measured air mass,"the discharge iline pressure and.the total discharge line air-steam mixture volume at the end of the water ref lood transient, the terrperature of the gas aCuct~e can be calculated using the ideal gas laws by assuming +hat the steam in the mixture is at saturation corresponding to the discharge line pressure.

l(

NK Oorment 5:

The sensor failure rate was found to be quite high in the Mnticello Plant ramshead test program. Sensors of the, sane manufacturer mxhl used in the Mmticello test will also be used in the Caorso test. In light of this experience, we believe that redundant instrumentation is needed in critical areas. For instance, 'redundant sensors should be provided in the foll<wing locations:

a. The vicinity of Quencher A and the place by which cmbined loads from multiple SHV's actuation will be determined.

b. SRV line betw~ elevation 51.612 and 45.770. In addition, level Response: probes should be added between Ll and L21.

(a) Slight rearrangement of instrunentation in the suppression pool was acccmplished since the issue of revision 2 to the Caorso test dated Decenber 1976. Specifically, pressure transaucers plan'NEDDY-20988)

P36 and P37 were shifted frcm their original locations near support column 8 to the positions shown an attachnent (3). While this shift does not represent a,direct backup for sensors P23 and P38, data from these sensors will be sufficient to allow evaluatian of multiple valve pop

!! (!

loads should.either'P23 or P38 fail. At the~sane itive,-should sensors &23

~

and P38 remain intact, sensors P36 and P38 will provide additional useful data which would not be exact duplicates of reachngs fxom P23 and P38.

(b) The spacing of level @xbes Ll and L12 is considered to be of minor iztportance for the follcwing reasons:

(1) Licensee data and GB nedels indicate. that the water column will not rise to sensor Ll during the nannal vacuum breaker tests.

(2). Me large n~Rer of level prcbes bet@em sensors Ll and L7 allm an evaluatian of the velocity and acceIeratian of the water calurm; this informaticn, in turn, can be used to deteanine approximately the mmcimum level to which the water calurtn rises

!. g ~

chu~g the ref lccd transient.

NBC Caraent 5 Page 2 (3) Temperature sensor T3 (see attachrrent (2) ) can be used to provide additional water level information between sensors if Ll and L12 required.

(4) Low pressure sensors P7 and PS (see attachrrent(5) ) (0-25 psia) in the SRV line can also be used to measure water level. 'This was done in the recant. Nonticello quencher tests where the water level sensors and the inferred water level fran the lm pressure readings cxnpared favorably..

With regard to the question of redunduncy for level prcbes between elevatians 51.612 and 45.770, temperature sensors T3, T4, T5, T27, and T28 can function as level sensors if m~uired and will provide redunduncy for .the level probes in case of failure.

NRC Conment 6:

Suhnerged structure loads have been identified as a pr'imary design load for the Mark II containnent. We believe that the analytical program indicated in the Mark II Owners Group neeting which was held on February 16 and 17, 1977, is insufficient to support the design loads for submerged structure without experinantal data. Therefore, we recxxmend that additional pressure sensors should be installed on support columns and downcomers to measure the drag load during SRV operation.

BesEonse:

Instnmentation which has been installed at Caorso is expected to provide experimental data to support the analytical quencher models relative to submerged structure load caused by SRV actuations.

Measurements on two downccmers and a support column adjacent to ~cher A will provide the needed loads information. The instzwnentation used, which is shown in attachments 6, 7 and 8 consists of the following:

1. Strain gauges on vents Nos. 1 arxl 9.

2. Pressure transducers on Support Col+an No. 7 and Vent No. 9.

GE believes the information cbtainM on pressure loadings, supplemented by the strain gauge data, will provide sufficient information to show that.

the analytical ncdels predict reasonably conservative loads. The in-plant structure loads that can be directly or indirectly applied to define II for other Mark plants. The tests are planned as a confirmatory test

~

test data itself is not intended to provide a loading basis for submerged to show the conservatism of the analytical nadels, which could then be used to define loads for other structures (not instrumented in Caorso) and for stru jtures in other Mark II plants.

,l L1

NK Ccmnmt 7:

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

Reste:

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

TABLE 6 1 TEST NAmZX (14)

Test Type Discharge Valve Closed Time Iaat $ 1 '3pa1ve Initial Pi e Conditions 3 Time sec. CVA sec Pi e Coolin mrs.

0 A CP$ NWL, 10'B 5 1 SVA A CP, NWL, 10" VB 5 >2 2 SVA A CP, NWL, 10" VB 5 >2 3 SVA A CP, NWL) 10" 'VB 2 >2 4 SVA A CP NWL 1O" .VB >2 501 SVA CP, NML, 10 VB 60 502 CVA A WP, TWI., 10" VB 60 503 CVA A WP, TWL, lO" VB 60 504 CVA A HP, TWL, 10" VB 60 505 CVA A HP TWL, 10" VB >2 601 SVA A CP, NWL, 10 VB 10 4 602 CVA, A WP, TWL, 10" VB- 10(4) m m 603 WP) TWL, 10" VE D C D m a-a CVA A lo(4) 604 CVA A HP, TWL, 10" VF lo(4) ~ M CA D D 605 CVA A HP, TWL, lo" vs >2 Z R O CO SVA A CP, NWL, Re uce 2 po 00 8" SVA A CP NWL Reduced >2 9 SVA CP) Reduced VB (6) >2 10 SVA CP Reduced VB 6 >2 1101 SVA A CP, NWL, Reduced VB (7) 60 CVA A WP, TWL, Reduced VB (7) 60 1803 CVA WP, TML, Reduced VB (7) 60 1104 CVA A HP, TWL, Reduced VB (7) 60 1105 CVA A HP TWL Reduced VB (7) >2 1201 SVA A CP, NWL, Reduced VB (7) 5 lo( )

1202 CVA A MP, TWL, Reduced VB (7) 5 10 (4) 1203 CVA A WP, TWL, Reduced VB (7) 5 10(4) 1204 CVA A HP) TWL, Reduced VB (7) 5 10(4) 1205 CVA A HP TMI. Reduced VB 7) 5 >2 1301 SVA A CP) NWL, 10 VB 2 (15) 1302 CVA A WP, TML, 1O" VB 2 (15) 1303 CVA A WP, TWL, 1O" VB 2 (15) 1304 CVA A HP) TWL) 10" VB -2 (15) 1305 CVA A HP, TWL, 10" VB 15 >2

~ ~7ÃCP+lr:r.) T (z)

Test Type Discharge Valve Closed Time Test N(1)(2)(10)(13)Valve Initial Pi e Conditions (3) Time Sec. CVA Sec Pi e Coolin .Hrs.

1401 j SVA A CP~ NWL, 10" VB 2 (15) 1402 CVR A WP ~ TWL, 10" VB 2 (15) 1403 CVA A WP I 'TWLt 10" VB 2 (15) 1404 CVA A HP ~ TWL~ 10" VB 2 (15) 1405 CVA A HP TWL 10" VB >2 15 ~

SVA F CP, NWL, 10" VB >2 16 SVA F CP NWL 10I! VB >2 17 SVA E CP i NWL, 10" VB >2 18 SVA E CP NWL 10" VB >2 19 SVA U CP~ NWL, 10" VB >2 20 SVA U CP NWL 10" VB >2 NWL 10" VB >2 'o iW 2201 2202 SVA SVA CVA A

A A

CP CPi NWLi 10" VB 20 20 (8) 5 lo(4) lo (4) o (

%7 m m m I-I C/)

a 2203 CVA A HPi TWL, 10" VB 5 lo (4) O O 2204 CVA A HP ~ TWL ~ 10" VB 5 lo (4) M ~ D) 00 CVA A HP TWL 10" VB 5 >2 Retest of Tests 1 Throu h as Re uired A,F CP ~ NWL, 10" VB 5,10 (9) >2 1 "VB 5 10 9 >2 I~ >2 5 10 15 9 W7 HVA 'A~F~E~U CP~ NWL~ 10" VB 5,10,15,20(9) >2 28 MVA A,F,E,U 5lot 15 t20 (9) >2

>2 29 HVA AIF~EtU 5,10,15i20 (9)

MVA AFEU CP NWL 10" VB 5 10 15 20 (9) >2 31 BCDL CP, NWL, 10" VB 5 tlo 15 ~20 (9) >2 32 AfBJDgH CPg NWLg 10 VB 5/10 f 15, 20,25/ >2 K L 30 35 40 9

Test Type Discharge Valve Closed Time Test N(l)(2)(10)(13) Valve Initial Pi e Conditions (3) Time Sec. CVA,Sec Pi e Coolin Hrs.

33 50 psia~Reactor A CP~ NWL, 10" VB 20 2 Pres (10) SVA 34 100 psia Reactor A CP, NWL, 10" VB 20 2 Pres (10) SVA 35 200 psia Reactor A CP, NWL, 10" VB 20 2 Pres (10) SVA 36 400 psia Reactor A CP ~ tAJL ~ 10" VB 20 2 Pres (10) SVA 37 600 psia Reactor A CP, NWL, 10" VB 20 2 Pres (10) SVA 3S 800 psia Reactor A CP, NWL, 10" VB 20 >, 2 Pres (10) SVA SVA B d CP NWL 10" NWL 1 ii VB 20 ll 12 2

2 C7 (

X7 O Pl Pl 41 SVA (13) A LV~ NWL ~ 10" VB > 2 Optional O ~

Ch h)

O to O

42 SVA A LV, NWL~ 10" VB 2 Optional SV Z,V NWL 10"- VB > 2 0 tieeal OO OO 4401 SVA (13) =A LV~ NWL~ 10" VB 10 (4) Optional 4402 CVA "A LVq TWL~ 10" VB 10 (4) Optional 4403 CVA A LV~ TWL~ 10" VB 10 (4) Optional 4404 CVA A LV, TWL, 10" VB 10 (4) Optional

>4b LV TWL 10" VB 10 (4) 2 Optional 45 Retests of Tests 24 Throu h 41 as Re uired

NEDM 20988 REVISION 2 ADDENDUM 2 Notes:

(1) SVA = Single valve actuation CVA = Consecutive valve 'actuation (2) Reactor power level to be as follows at the beginninq of each test:

Test No. Reactor Power Level 1 through 22, 41through 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 (3) CP = Cool pipe WP = Harm 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 (4) Subse uent Actuation - Transient Water Level 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 water leg transient is to be based on water leg transient data from prior test runs.

(5) Vacuum breaker butterfly valve at 25-29'90's full open), but at same settin (6) Yacuum breaker butterfly valve at 9-14', but same setting for PhtlPQsB.

(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 SRV line of less than 15.0 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 .T2~1 and T23 are constant to within 2;F,(l'C) rise per second in Test 21'9)

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 through 39e 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 526.7 - 32.3'C) for all singl'e valve tests except Test No.40 (see Note 11).

6-3B

NEDM 20988 REVISION 2 ADDENDUM 2 Notes: (cont'd)

(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 readinq 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 SRV 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 conditionsrelief (such as loss of site power and containment isolation) which results in valve discharges which heat 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 overshoot in the discharqe tests 1 through 4 that the maximum water level

(

line is 15 .6 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 averaqe four-valve predicted pressure from the empirical model, OR if the maximum pressure two-thirds from the six four-valve tests exceeds of the difference of the 90-90 the predicted value plus design value and the predicted value from the empirical model, run three addi'.i,onal four-valve tests, for a total of nine such tests. (Maximum pressures at the highest real-time pool boundary,p-">,sure sensors are to be used as bubble pressures. The empirical model is to be evaluated at actual Caorso test conditions.)

6-3C

NEDH-20988 REY I S ION 2 FIG. 5-2A PiPE AND QUENCHER TEMPERATURE SENSORS t.c. 5d.95&

23 26

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a7OC 4 50 I9's3 O00 1

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D T Col. Crt Wo,1 Seo also Fig, 5 28 Temp. sensors oa Ar Ho. 1 should be placed @ideal between folio&ay colm'f holes:

1 2 5 6 9 ~ 10 12 1) 16 ~ '17 Moo 3 JO0 4

' ))<o5 Era Zo, 1 No+ 2 Xoto1 Tl thrcMyh T10, and T27, T28 located $ ns$ de pfp$ ny or quenche T21, 22, 23 located on outer surface of o3pe T24, 25, 26 located near inner wall of p1pe 5-10

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NEON-20988 REVISION 2 F'G. 5 gA I".V.L S'tlSCRS 12 L

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CZ4

/ 9/4 9

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Era Ko. 1 6

Last ooluaa of holes

{Col, 17)

NEDN-20988 REVISION 2 C

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hot>> 0+v>>elva vari

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950 Do. 1 T0 km Zo.

ZZ 4P. 70 No, $

Zoo 4 5~6~4,5 Are Zo.

so~2

'otee~ ill Ionsors S locate" i,=sile yipiaz or qvonoher All diatancea in am.

'59

ZEN-20988 REVIStON 2 FIG, 5 5"- "C 'ZGGIZt Vc,.'iT PP= S~ ".

v~SOPS vatez lerel (46/

<CjooA Zotoi Pressure tranaduoers to be oriented on radial line to guenoher "l" S

NEDH-20988 REVIS?ON 2

?~ Js CHUG"- LCCnTIC. iGCOKER V"-K'S Drnc 1g EL 4g7p Causes 400 ma belov braciag hub Fou? uniaxial gLUgos to be located &t 90 intervals around circumference on 2 selected dovncomer vents (see Fig. 5-5k for location) . Pairs of diametrically opposed strain gauges to be connected to read moments. Orientation of gauges relative to containment axes to be recorded.

S.CD Mos.51,52,5'4 on Vent Ho.

S.C, Hos ~ 55 ~ 56 57 ~ 58 on Vent No ~ 9

~

5-21

NET-20988 REY I S ION 2 CO'CRe

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P,'59,40 41,42

'~ater Level i9 41 p P 94 2/

40 P Zl, 40.700

<4004 bootes Preeessre ssenaors to be oriente4 alan@

raCiaL Line to qssencher "4" 5-17