GO2-80-279, Responds to Re Actions Taken to Preclude Cracking of Jet Pump Holddown Beams.Reducing Tension Preload to 25 Kips on Beams Provides Adequate long-term Solution

Responds to Re Actions Taken to Preclude Cracking of Jet Pump Holddown Beams.Reducing Tension Preload to 25 Kips on Beams Provides Adequate long-term Solution

ML17276A181
Person / Time
Site:	Columbia
Issue date:	12/04/1980
From:	Bouchey G WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Tedesco R Office of Nuclear Reactor Regulation
Shared Package
ML17276A177	List: GO2-80-279, Responds to Re Actions Taken to Preclude Cracking of Jet Pump Holddown Beams.Reducing Tension Preload to 25 Kips on Beams Provides Adequate long-term Solution GO2-81-442, Forwards Minutes from Mechanical Engineering Branch 810929-30 & 1001 Meetings in Woodbury,Ny & 811016 Preservice Insp Meeting in Washington,Dc.Agenda Items from Both Meetings & Subsequent Resolution Encl ML17276A179 ML17276A180
References
GO2-80-279, NUDOCS 8111090527
Download: ML17276A181 (48)

Text

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December ~, 1980 G02<0-279 Docket No. 50-397

&. R. L. Tedesco, Assistant Dfrector, Licensfng Ofvfsfon of Licensing U. S. Nuclear Regulatory Cottmfssf'on washington, D.C.

20555 Cear Hr. Tedesco:

Subject:

CRACKING OF BMR JET PlNP H L G

B Ref.:

Letter, R. L. Tedesco to N. 0.

nd, same sugect, dated August 0

~ g The referenced le e

r quested the Supply System provide specific fnformatfon regard g

c fons taken to preclude cracking of get pump holddown The f 1

fng responses correspond directly with ~he questions o

d fn yo er.

2g beams have been installed, but will be retensfoned f

a 3 k'reload to a 25 kfp preload before fuel load.

This fs x

cte to increase beam operating time to crack Initiation at 2.5% probability level to a range of 19 to 40 years During operation, periodic inspections will be conducted as part of our overall fnservfce inspection progra~.

Inspection frequen-cfes'ill be developed fn the future based on lead plant inspec-tion results and the results of future GE testfng.

These inspec-tions should provide adequate warning of potential beam failure.

2.

It fs our posft~on that reducing the tension preload a 25 kfps on the beams provides an adequate long term solution.

If a problem is still present, as identified by our fnservfce fnspec-

tfons, improved heat treated beams may ~s purchased frpm GE.

Tests indicate the improved beams may provide c~uble the time :o crack fnftfatfon as compared

o tne current beams.

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QUESTION NO.

48 Provide a commitment to NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking".

RESPONSE

The Supply System's response to NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Line Nozzle Cracking", will be completed by January 8,

1982.

The current status of our position on the feedwater and CRD cracking problems is as follows:

Feedwater Nozzle o

The WNP-2 welded sparger feedwater nozzle is an NRC approved desi.gn (NUREG-0619, Page 15).

o The WNP-2 nozzles are unclad.

o The RWCU System has been rerouted so it discharges to all nozzles.

o The need'or a low flow feedwater controller as described in GE report NEDE 21821-A has not yet been established.

WNP-2 current design employs a low flow feedwater controller.

The Supply System is currently evaluating whether or not the existing controller meets the intent of NUREG-0619 and NRC generic letter 81-11.

o An augmented Inservice Inspection program for the feedwater nozzle has been submitted in response to FSAR Question 121.8.

This response will be considered in our reply to NUREG-0619.

CRD Return Line o

CRD return line has been cut. and capped as allowed by NUREG-0619, Page 31.

o CRD return line has been rerouted through redundant equalizing valves to the exhaust water header.

o The control rod drive preoperational test will demonstrate that the system is fully operational and that all components including the hydraulic drive mechanisms,

pumps, and flow control valves fun'ction properly.

The CRD System will be configured with the modifications noted in the NRC concern.

'f li

o In order to assure satisfactory system operation with the single failure of an equalizing valve, the proposed design modification will include the addition of two equalizing valves installed in a parallel configuration.

The failure of either valve will not impair CRD operation for any foreseen operating or accident condition.

o There will be no increased potential for carbon steel corrosion products to be deposited in the drives.

A11 lines in the WNP-2 Hydraulic System after the drive water filters are made of stainless steel.

o The NRC requested GE by letter of January 28, 1980, to recalculate the makeup flow capacity for the 251-inch BWR-5 without the CRD return line.

This generic information has been provided by letter of May 2, 1980, from Mr. R. L. Gridley, GE to Mr. D.

G. Eisenhut, NRC-,

concurrently with this docketed response for LaSalle.

The results indicate that the'51-inch BWR-5 CRD system without a return line (capped Nozzle 10) can achieve a

vessel makeup flow in excess of its calculated boiloff rate of 180 gpm.

This confirms the same boiloff rate as previously documented in a March 14, 1979, submittal from GE.

Furthermore, since the CRD system is not designed to perform an ECCS function, the additional testing to demonstrate the required return-flow capacity to the vessel is not warranted.

Summation - This item is closed.

QUESTION NO.

49 There are several safety systems connected to the reactor coolant pressure boundary that have design pressure below the rated reactor coolant system (RCS) pressure.

'here are also some systems which are rated at full reactor pressure on the discharge side of pumps but. have pump suction below RCS pressure.

In order to protect these systems from RCS pressure, two or more isolation valves are placed in series to form the-.interface between the high pressure RCS and the low pressure systems.

The leak-ti'ght integrity of these valves must be ensured by periodic leak testing to prevent.

exceeding the design pressure of the low pressure systems thus causing an intersystem LOCA.

Pressure isolation valves are required to be category A or AC per IWV-2000 and to meet the appropriate requirements of IWV-3420 of

~

Section XI of the ASME Code except as discussed below.

Limiting Conditions for Operation (LCO) are required to be added.to the technical specification which will require corrective action; i.e.,

shutdown or system isolation when the final approved leakage limits are not met.

Also, surveillance requirements, which will state the acceptable leak rate testing frequency, shall be provided in the technical specifications.

4 Periodic leak testing of each pressure isolation valve is required.

to be performed at least once per each refueling outage, after valve maintenance prior to return to service, and for systems rated at less than 50% of RCS design pressure each time the valve has moved from its fully closed position unless justification is given.

The testing interval should average approximately one year.

Leak testing should also be performed after all disturbances to the valves are complete, prior to reaching power operation following a refueling outage, maintenance, etc.

The staff's present position on leak rate limiting conditions for operation must be equal to or less than 1 gallon per minute for each valve (GPM) to ensure the integrity of the valve, demonstrate the adequacy of the redundant pressure isolation function and give an indication of valve degradation over a finite period of time.

Significant increases over this limiting value would be an indication of valve degradation from one test to another.

Leak rates higher than 1

GPM will changes are below 1 GPM above the system design precludes measuring These items will be reviewed on a be considered if the leak rate previous test leak rate or, 1

GPM with sufficient accuracy.

case by case basis.

0

RESPONSE

The valves which separate the Reactor Coolant System (RCS) from interfacing low pressure systems are listed in Table I.

These valves are included in the WNP-2 Pump and Valve Inservice Testing Program which was developed in accordance with the ASME Boiler and Pressure Vessel

Code,Section XI, Subsection IWV.

The Supply System's position is that the requirements of the Code provide adequate assurance of valve integrity.

Specifically:

A)

The Supply System will leak rate test the valves listed in Table I at least every two years (IWV-3422).

This position is justified by the following:

l.

All the valves listed in Table I have direct monitoring position indication which verifies valve position, in the Control Room.

2.

The low pressure portions of these interfacing systems are protected against an intersystem LOCA by the following:

a)

The normal functional differential pressure forces the check valves on their seats.

The air operator of these testable check valves cannot open the valves at normal differential pressure.

(HPCS-V-5, LPCS V 6g RHR V 41Ag Bg Cg RHR V 50A~

Bg RCIC V 66)

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b)

Electrical interlocks prevent the motor-operated valves from opening when the differential pressure across the valve exceeds specified limits (LPCS-V-5, RHR-V-42A, B, C) or when the RCS pressure exceeds specific values (RHR-V-53A, B, RHR V 8g RHR V 9g RHR V 23'HR V 123A~

B)

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c)

Whenever excessive leakage is present at a pressure boundary isolation valve, this leakage will increase pressure in the downstream side of these systems which will annunciate a high pressure alarm.

d)

Excessive leakage will be channeled into the suppression pool where an increase in suppression pool level will be indicated.

e)

The high pressure core spray pump suction piping is protected by an additional check valve on the pump discharge.

B)

The Supply System will specify the leak test medium and the test acceptance'riteria as permitted by the ASME Code (IWV-3425 a 3426).

C)

The periodic leak test, will be done prior to entering Operational Condition 2.

D)

After maintenance which is deemed by the Owner to affect leak tightness of the valve, leak testing will be performed in accordance with ASME Section XI prior to the valve's returning to service.

Summation The NRC needs to have their IST people meet with the Supply System.

NRC will set up a meeting an'd get back to us.

TABLE I REA".TOR COOLANT SYSTEM PRESSURE ISOLATION VALVES Valve Number.

HPCS-V-4 HPCS-Y-5 LPCS-V-5 LPCS-Y-6 RCIC-V-66 L RCIC-V-13 RHR-V-8 l-RHR-Y-9 RHR-V-23 R-V-41A, B, C

RHR-V-42A, B, C

RHR-V-50A, B

Valve Tvae TC TC TC Globe TC TC Size Function 12" (0)

HPCS injection, line containment isolation valve 12" (i) 12" (o)

LPCS injection line containment isolation valve 12" (i) 6 II 20" (i)

RCIC Rx head spray containment isolation valve (o)

(i)

RHR shutdown'I'i cooling supply to RHR pumps containment isolation valves 20" (o) 6" (o)

..SR to Rx head spray containment isolation valve 14" (i)

RHR LPCI containment isolation val ve 4

14-(~)

18" (i)

RHR shutdown cooling return line to Rx vesse'ontainment isolation valve RHR-Y-53A, B RHR-V-.123A, B

Gl obe

]8II (o)

(i)

By-pass valve for Y-50A, B containment isolation valve o - outside i - inside G'

gate TC testable check

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iIn I +un 3.-"..3.2 Reac.or c"olant system leakage shaIl be Iimted to:.

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URE GUHOARY L AK"GE.

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gpm UH~DEHTiF~EO LAKAGE.

25 gp I

v a I 1 aakage averagec over any 24 h"J per 'co.

1 gpm leakaoe psig rcm any fied 'n Table at a reactor coolant system pressure at f:OCO =':0 reactor coolant system pressure isol'etio% valve speCi-

3. 4.3. Z-l.

At'." SCAB '

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GP "ETIO'iNL CGNOIT'OHS 1, 2 and 3.

ACiIQH:

Mith any PRESSURE BOUHOARY L AK'6=, be in at leas-HOT SHU:VGA wi hin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLO SrUTI",G'n'H: i hin s e naxv 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Mi h. any reac cr ccolant sys am leakace crea ar -.han -he limits in o and/or c, abcve,

".educe c:.e leakace ra.a to. w'.thin the 1 mits wi:hin 4 hcurs or be in at least HGT 5:-iUTDCn'H within the, next 1"',.ours and in CQLQ SIiUTOOMH within he follcwin"I 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

P /~g C

c Pith any reac.or coolant system pressure isolation valve leakage greater han the above limi, isola a lhe hign pressure pcr:icn cf the affec.ad svstem from the 1Iow pressure portico wi hin 4.".cur by

'use of at least two closed manual or;deactivated automatic valves,.

or "e in at lees-HOT SHUTGOMH wi,"in the nex-12 hcurs and in COLO SHUTGQMH within +he following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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1 The reactor ccolant sys.am laakace shaIl be "e:-.,Cnstraiad to be wi:hin

.he above limits by.

a.

":cnitoring the pr'.mary containmen gaseous rad oac ivlty at Ieas cnc i.o pnel

~ C pc,a v a per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, iculata and 0

1 monitoring the primary c.n-ai".-,en sump>>ow

. a a at Ieast once par 12 hcurs, and c ~

Yl n I or> ro the 'prima rv conte i a

1 eas once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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SURYEILLAHCE REOUIREHEHTS (Continued)

-"..4.3.2.2 Each reactor coolant system pre sure isolat',on valve speci ied T-'

3 2.-.

'"-ll b ee except'that in lieu of any leakage testing required b>>'pecification 4.0.5, each valve shall be demonstrated OPERABL" by verifying leakage to be within its 1 emit:

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At least once par 18 mont.".s.pm;ere 3 ~sve y spa>>:l p

~(Wr.w

~'k/

Pc Prior to entering STARTUP whenever

-'he plan-'as been in COLD SHUTDOWH for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more and is leakage

.as

'ing has not been perfor'ed in the previous 9 months.

c.

Prior to returning the valve, to service following maintenance, repair or replacement work on the valve.

d.

Within 2~ hours following valve ac uation Cu~ tc, automatic or menue.l ac ion or flow through the valve.

The provisions o

Specisicatson 4e0.4 are not a"plicable for entry into OPERA" T:OHAL CONDITIONS 2 or 3.

1 A SAtl L UHIT 3/4 4-8

3. lc ""1

WNP-2 DSER QUESTION NO.

50 Does the design criteria for component supports in WNP-2 systems categorize the stresses produced by seismic anchor point motion.

of piping and the thermal expansion of piping as primary or secondary?

NRC position For the design o~upports, these stresses should be considered primary.

Expansion stresses in the support themselves may be categorized as secondary.

F

RESPONSE

I When a component is covered by the ASME Boiler and Pressure Vessel

Code, the stresses due to relative displacement (thermal

& seismic anchor point) are treated as primary or secondary

stresses, per ASME Section lIZ, for piping.

For loads due to relative displacement, supports are analyzed as per ASME Code Section NF.

The applicants are committed to assess the design with respect to the NRC position.

10/16/81 Item 1

WHP-2 will provide information to indicate those areas on the RPV that can't be fully mechanically inspected by inservice examinations.

10/16/81 Item 2 WHP-2 will provide an update of the list of non-examinable welds (provided in response to 121.19) in January 1982.

10/16/81 Item 3 (g 121.20) llNP-2 will provide an expanded response to discuss the design vs. operating temperature and pressure on those lines exempted from Preservice volumetric and/or surface examination based on paragraph IliC-1220(a)'f section XI for those lines to which this exemption criteria was applie) that have a design temperature of 212 (versus'00

).

'0/16/81 Item 4 (g 121.20) lfNP-2 will revise the response to.indicate that a

volumetric examination in lieu of 'a surface examination

~

for HPCS piping ~ 1/2" thick will be conducted on lOX of those welds exempted by chemistry; otherwise the commit-ment of 105 sample by a surface method was acceptable.

Additionally a table itemizing those welds exempted by chemistry will be attached to the response of g 121.20, showing sizes, and operating temperatures and pressures.

0

10/16/81 Item 5

(g 121.21)

The response to this question will be revised to add the following paragraph:

"During the conduct of inservice examina-tions the criteria for evaluating a crack-like indication will not be limited to signal ampli-fication alone.

Appropriate consideration will be given to other factors such as the l.ocation of. the indication.

Those indications determined to be crack-like will be evaluated."

I I'

10/16/81 Item 6

(g 121.22) l)NP-,2 will revise the response to this question to add the following statements:

"Any additional welds which require PSI will be included in that program in Amendment No.

3 scheduled for submittal 12/81.

Figures 3.6-147a through d in section 3.6 of the FSAR define the break exclusion areas.

Sec-tion 3.6.1 commits to an augmented inservice inspec-tion program on these lines.

Section 5.2.4 of the FSAR will be'mended to reflect these augmented requirements.

(Section 5.2.4 draft attached.)

For those lines beyond the outboard containment isolation valves which are not normally pressurized, the inspection boundary will stop at the outboard isolation valve.

This approach is consistent with that taken by the piping designer.

No pipe whip restraints are installed beyond the containment iso-lation valve by the designer because these lines are not pressurized and therefore are not subject to pipe whips."

AUGt'lENTED INSERVICE INSPECTION TO PROTECT AGAINST POSTULATED PIPING FAILURES An augmented Inservice Inspection Program will be implemented for NNP-2, on highenergy* Class 1 piping systems which penetrate containment for which the effects of postulated pipe breaks would be unacceptable.

This program will entail a volumetric examination of all circumferential butt welds (surface examination for socket welds) between the first pipe whip restraint beyond the inside containment isolation valve, and first pipe whip restraint beyond the outside containment isolation valve on high-energy Class 1 lines greater than one (1) inch which penetrate the containment.

In those cases where the piping beyond'he containment isolation valve is not pressurized (i.e. low energy),

the augmented Inservice Inspection boundary will stop at the containment isolation valve.

This program will include branch. lines which fall within the augmented Insepvice Inspection boundary to the first pipe whip restraint beyond the branch line isolation valve on the first normally closed valve, whichever comes first.

• High-energy lines include those systems that, during normal plant co'nditions, are either in operation or maintained pressurized and where either the maximum operating pressure exceeds 275 psig or maximum operating temperature'xceeds 200 F. If, for a particular line, the above pressure and temperature limits are not exceeded more than 2/ of the time that the system is jn operation, then that line is considered moderate energy and is exempt from the requirement for augmented

'nserYice inspection'.

K

10/16/81 Item 7

WNP-2 will submit a relief request for the requirement to perform a sur'face vs.

a volumetric examination on the RHR pump casing welds.

(guestion 3)

10/16/81 Item 8 Include a statement in the executive summary to the PSI program that 1/2 inch wall thickness, which is part of the Summer 1978 Gode, has been used in this program.

(question 4)

0,

10/16/81 Item 9 klNP-2 will include a statement in the executive summary of the PSI program that the four inch nominal pipe size is taken from the 1978 Summer Addenda of ASME section XI.

10/16/81 Item 10 lNP-2 will provide a statement in the executive summary of the PSI program plan stating that bolting examinations in categories B-G-l, B-G-2 and C-D to ASME section XI 1977 edition to Summer 1978 Addenda will be accomplished.

10/16/81 Item ll WNP-2 will add a statement to the executive summary of the PSI program plan to indicate that the Summer 1978 Code is used in performing branch connection examinations.

10/16/81 Item 12 WNP-2 will provide a statement to the executive summary of the PSI program to indicate that RPY nuts and studs were examined per the Summer 1978 Addenda to section XI.

10/16/Sj Item 13 lJNP-2 will submit a relief request for those RPV and

'iping welds where full code compliance cannot be accomplished.

(Submit by 12/14/81.)

10/16/81 Item 14 HHP-2 will provide a description of the Reactor vessel examination made in the PSI program executive summary.

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