ML20196H337

Responds to GL-96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions, W/List of Attendees & Handouts Encl

ML20196H337
Person / Time
Site:	Wolf Creek
Issue date:	07/22/1997
From:	Stone J NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
GL-96-06, GL-96-6, TAC-M96887, NUDOCS 9707250083
Download: ML20196H337 (36)

Text

lE ]$ ,

y i UNITED STATES j g j NUCLEAR REEULATCRY COMMISSION WASHINGTON, D.C. namaa aani fMSE i

) p $ l

%,***-**p July 22,1997 l l

LICENSEE: Wolf Creek Nuclear Operating Corporation FACILITY: Wolf Creek Generating Station ,

i

SUBJECT:

RESPONSE TO GENERIC LETTER 96-06. " ASSURANCE OF EQUIPMENT '

OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS" (TAC NO. M96887) l Backaround On June 25, 1997, members of the NRC staff met with representatives from Wolf Creek Nuclear Operating Corporation (WCNOC) to discuss their planned actions in response to Generic Letter 96-06. The licensee's response is contained in a letter to the NRC dated January 29, 1997. The meeting was held in the NRC offices in Rockville, Maryland.

Summary Generic Letter 96-06 identified three conditions that could occur within systems inside containment:

1. Waterhammer in containment cooler cooling water systems.

2. Two phase flow in containment cooler cooling water systems following a LOCA or MSLB. and  :

3. Thermally induced overpressurization of isolated pipe inside containment.

For the first condition. Wolf Creek has experienced waterhammer in the  ;

containment cooler cooling water system, ihis condition would exist upon a  ;

loss of offsite power (LOOP) with or without a LOCA or MSLB. However, the i licensee has determined that the containment water system is designed to '

withstand the effects of this waterhammer and no )lant modifications are ,

planned to comply with the requested actions of G. 96-06. The NRC staff will  !

review the licensee's assessment and ' actions to address the waterhammer issue during GL 96-06 followup evaluations.

For the two phase flow issue. Wolf Creek has orifices installed in the containment cooler cooling water system discharge line which, at steady state i conditions following LOCA or MSLB will maintain system pressure above the i saturation pressure of the water in the system. However, it was determined l

that refill of the cooling lines takes longer than the original assumptions (greater than 65 seconds vs. 60 seconds). The licensee revised the containment pressure / temperature response based on a refill time of 70 seconds. The results show an insignificant effect of the containment I

pressure / temperature response. Because Wolf Creek's current design precludes l two phase flow in the containment cooler cooling water system, no plant t l c u eul-eO C)F~O\,t 9707250083 970722 PDR P

ADOCK 05000482 PM hhh -

1

• . e ..

\

l

, l modifications are planned to comaly with the requested actions of GL 96-06 for the two phase flow issue. The NRC staff will review the licensee's assessment i

and actions to address two phase flow during GL 96-06 followup evaluations. '

For the thermally induced overpressurization of isolated pipe, a preliminary scoping study has identified 32 potential locations. A worst case calculation i indicates that at least one section of piping could exceed current licensing basis stress allowables. An operability evaluation, considering the worst i case temperature increase, indicates that all )iping sections in question are operable. Detailed evaluations are currently ]eing performed and the licensee is developing contingency modifications. The results to date has eliminated l three locations and has identified five others for potential elimination from the )otential for thermally induced overpressurization. The licensee has esta)lished a target for implementing any modification required to mitigate any identified overpressurization condition by the end of the ninth refueling outage, scheduled to begin October 4. 1997. However, the commitment is to complete the modifications by the end of the tenth refueling outage, scheduled for the spring of 1999. Following the meeting, the licensee formally committed, by letter dated July 1. 1997, to complete any necessary plant modifications prior to startup from the tenth refueling outage, currently '

scheduled for the spring of 1999.

In trying to resolve the overpressurization issue, the acceptability of the use of ASME Code. Appendix F criteria was discussed. The staff stated that this issue is currently being reviewed in the context of industry concerns pertaining to Sotential unreviewed safety questions (US0s). The licensee was considering macing a request to use Appendix F criteria. If modifications were found to be required, the modifications would meet the current licensing basis code requirements. l The licensee stated that if all modifications were not com)leted at the end of the ninth refueling outage. they would complete the ones tlat could be done  ;

with the unit on line, but would have all completed by the end of the tenth  !

refueling outage. For those modifications that can't be completed by the end ,

of the ninth refueling outage. the licensee proposed to restart the unit as '

long as a valid operability assessment was in place. The staff stated that these matters would have to be handled on a case-by-case basis. It would be important to be able to shcw substantial progress toward resolving the issue. ,

an acceptable operability determination, the schedule for completion of the remaining work and why completion of the work was not possible.

l l

l l

l l

l

. .d . ..] ,

Enclosed is the handout provided at the meeting by the licensee and the list l -of participants in the meeting.

l {

James C. Stone. Senior Project Manager

- Project Directorate IV-2 Division of. Reactor Projects .

Office of Nuclear Reactor Regulation

-Docket No. 50-482

Enclosures:

1. Handout.

2. Attendees cc w/encls: See next page 3

l 4 i

L I

.. ~ g. -

July 22, 1997

. Enclosed is the handout.provided at'the meeting by the licensee and the list of participants in the meeting.

Orig, signed by James C. Stone. Senior Project Manager.

Project Directorate IV-2 Division of Reactor Projects-Office of Nuclear Reactor Regulation Docket No.'50-482 DISTRIBUTION: -(w/encls. 1 and 2)*

Docket Filec PDIV-2 Reading  :

Enclosures:

1. Handout PUBLIC JStone

2. Attendees PGwynn. RIV WJohnson, RIV  :

cc w/encls: .See next page DISTRIBUTION: (w/ encl. 2) >

SCollins (SJC1)

FMiraglia-(FJM)

AThadani (ACT)

RZimmerman (RPZ)

JRoe (JWR)

EAdensam (EGA1)

WBateman (WHB)

EPeyton'(ESP)

TMarsh':(LBM) '!

DWatsman (RHW)

GHubbard (GTH)

KManoly (KAM)

BWetzel (BAW)

'*Hard Copies E

s t

DOCUMENT NAME: GL9606MT WC 0FC PDIV-2/PMn PDIV-2/LA NRR:SPLB _,, NRR:EMEB,.

NAME JStone Y EPeYfob TMarsh Y DWds'ma[

! DATE 7/ I//97 7/ t3/97 7/d/97 7/ TV97 l 0FFICIAL RECORD COPY V

t-t

.., - . _ , - . . . - . .w,--.. - - --

.l ,

cc w/encls:

Jay Silberg. Esq. Chief Operating Officer Shaw. Pittman. Potts & Trowbridge Wolf Creek Nuclear Operating Corporation 2300 N Street. NW P. O. Box 411 Washington D.C. 20037 Burlington, Kansas 66839 Regional Administrator. Region IV Supervisor Licensing U.S. Nuclear Regulatory Comission Wolf Creek Nuclear Operating Corporation 611 Ryan Plaza Drive. Suite 1000 P.O. Box 411 Arlington. Texas 76011 Burlington. Kansas 66839 Senior Resident Inspector U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Resident Inspectors Office P. O. Box 311 8201 NRC Road Burlington Kansas 66839 Steedman. Missouri 65077-1032 Chief Engineer Mr. Otto L. MaynarJ Utilities Division President and Chi:.! Executive Officer Kansas Corporation Commission Wolf Creek Nuclear Operating Corporation 1500 SW Arrowhead Road Post Office Box 411 Topeka, Kansas 66604-4027 Burlington. Kansas 66839 Office of the Governor State of Kansas Topeka. Kansas 66612 1

Attorney General '

Judicial Center 1 301 S.W. 10th  !

2nd Floor Topeka, Kansas 66612 l County Clerk ,

Coffey County Courthouse Burlington Kansas 66839 j

Vick L. Cooper. Chief l Radiation Control Program l Kansas Department of Health -

l and Environment Bureau of Air and Radiation Forbes Field Building 283 Topeka, Kansas 66620 I

l l

l l

l l

t t

1 .

Wolf Creek Nuclear Operating Corporation b e

' Generic Letter 96-06 Meeting

, 2 June 25,1997 l i

T

=

Generic Letter 96-06 >

Agenda Summary

System Description

• Waterhammer in Containment Cooler Cooling Water System

• Two Phase Flow in Containment Cooler l Cooling Water System Overpressurization ofIsolated, Water-filled Sections of Piping

• Discussion ,

1 s

L. i Summary -

1 I Waterhammer i

• Waterhammer can occur in Wolf Creek L containment cooler cooling water system.

• Wolf Creek is designed to w 1 stand effects of ,

waterhammer.

i

• No plant modifications necessary as a result of l review of Generic Letter 96-06.

i 2

=.

Summary -

Two Phase Flow

>

• Wolf Creek design precludes two phase flow at

steady state conditions following LOCA or MSLB.

• No plant modifications necessary as a result of review of Generic Letter 96-06.

i 1

l

Summary Thermally-Induced Overpressurization

• Wolf Creek has identified 23 penetrations and 9

other locations that need to be evaluated.

• Conservative, worst case analysis shows at least  ;

one location which could exceed present stress allowables.

Operability evaluation determined all locations can perform design function.

i i

I

i.

Containment Cooler (High Point)

Containment isolation Valv -

X Orifices

[

Essential Service Water To lXI lXI  !!Xl ll><H- Ultimate -

O Heat Sink Other Service Water N lXl lX: lXl l l><H-e Components LOCA and LOOP Service Water pumps stop Essential Service Water pumps start sequence:

- 'A' - 32 seconds

- 'B' - 37 seconds Valves complete swapover in 42 seconds Containment coolers and headers draindown and refill 5 B

_ _ _ _ . _ _ _ _ _ _ . _ _ _ _ __ ______.____m __ m__ ___

. .-li Waterhammer in Containment ll Cooler Cooling Water System  !

j .

i

• Waterhammer occured during the performance

ofintegrated emergency diesel generator and 4 i safeguards actuation tests performed in the Spring of 1993, during Refuel VI. .

Subsequent inspections revealed no damage to the containment cooler, cooling water piping and nozzles. However, one pipe strut was bent.

6

b.

Waterhammer in Containment Cooler Cooling Water System Change Package 05818 evaluated forces and redesigned four pipe supports.

• Piping and nozzles meet ASME Code stress allowablesi.

• Pipe supports modified in the Spring of 1996, during Refuel V111.

• Tubes and headers in the coolers meet ASME Code stress allowables.

i

Waterhammer in Containment .

Cooler Cooling Water System

• Two types of waterhammer considered in evaluation of GL 96-06

- Column Rejoining (occurs on LOOP with and without LOCA or MSLB)

• At Wolf Creek this occurs during the refilling stage. .

• Worst case is Train B - longest drain down period.

- Condensation Induced (occurs only on LOOP with LOCA or MSLB)

• At Wolf Creek this occurs only during the draindown stage.

• Worst case is Train B.

i 8  ;

i

Waterhammer in Containment Cooler Cooling Water System

Analysis for Column Rejoining Waterhammer

• LOOP without LOCA - controls design (Previous Analysis)

- System analyzed for waterhammer magnitude of 350 psig.

• LOOP with LOCA (analysis for GL 96-06)

- Magnitude of waterhammer will not be more severe than waterhammer caused by LOOP without LOCA.

9 j i

~=.

Waterhammer in Containment l'

Cooler Cooling Water System j

• Thermal hydraulic analysis with "RELAP" computer code.

Stress analysis with "ADLPIPE" using time history input.

l

• Modification to four supports.

• Piping, component nozzles, and supports meet ASME Code requirements.

10

^

Waterhammer in Containment 4

Cooler Cooling Water System Analysis for Condensation Induced Waterhammer (GL 96-06 Evaluation)

• Loading condition considered is LOOP

w/LOCA.

i

• Pressure pulse calculated for horizontal pipe run,

partially drained, prior to pump restart.

• Calculated pressure pulse is 326 psig.

i i

11

Waterhammer in Containment -

"! i Cooler Cooling Water System 1

i

• "ADLPIPE" stress analysis with pressure pulse .

applied as a time history loading.

• Pipe stress and nozzle loads were within ASME Code stress allowables.  :

• Pipe support loads in almost all cases were lower than previously analyzed numbers, and in all cases meet ASME Code stress allowables.

l 12

6. ,

Waterhammer Summary -

i

• Waterhammer can and has occurred at Wolf Creek.

• Piping, component nozzle, and support stresses ,

meet AS ME Code stress allowables.  !,

i

• Cooler tubes, headers and supports meet ASME Code stress allowables.

13  !

2..

Two Phase Flow in Containment .

Cooler Cooling System l

• Wolf Creek is not susceptible a1: steady state.

• During refill, some temporary flashing is predicted .to occur in cooler.

, 14

l .>

Two Phase Flow in Containment .\

Cooler Cooling System I

For two phase flow to occur, pressure must be equal to or less than saturation pressure.

Lowest System Pressure Highest Saturation Pressure 14" Orifice 40.3 PSIA 27 PSIA i

30" Orig

e 17.7 PSIA 7.7 PSIA i Coolers 92 PSIA 78 PSIA 15 i

i

______-.-..I . .' l

Two Phase Flow in Containment -

Cooler Cooling System

'During Refill of Cooler, Leading Edge of Water

Column is predicted to Flash

- To reach full system flow and pressure takes slightly longer than original assumptions (~65.2 seconds vs.

60 seconds).

- Insignificant effect on containment P/T response.

- Containment P/T calculation has been revised to account for slightly longer refill time.

t 16

s.

50 45 -

+- - - .

t liii 40 - . .

i

.g a i f

- 35 -..

.+ + .-

5 e

30 - .

c 1 .

a 25 -

+. . .

+ ..

c r e 20 - , . - .

E e

g- 15 =..

e '

oo 10

'60.0: seconds ' ' ~ 70.0 seconds-time delay time delay 5 -.

+ - . - -

0 ' " " "i """i"" '""' - ' " "I ' " " "i '""i ' ' " " "

1 E-2 1 E-1 1E+0 1E+1 1E+2 1E+3 1E+4 1E+5 1E+6 Time (second) .

Figure 1. Containment Pressure Response Following a LOCA ,

17

1 350 -

_ 325 - - . - - - - .

u.

?300 - - - --.

+- -

Q e 275 - - -

-+- .

+

u lii u

250 -. . .

8 225 E

.+.. -

7 4

,200 -.

. :+ -

. .+- -

e e

E 175 -- - -

.+ -

c j150 - =

so.o secones- 70.osecones- + -

O -

time delay time delay O 125 -.- , . .

100 ' > ' " "i '"""i""' ' ' " " " ' ' ' " " ' ' - " "i ' '"'"i ' ' ' " -

1 E-2 1 E-1 1E+0 1 E+1 1E+2 1E+3 1E+4 1E+5 1E+6 l Time (second)

Figure 2. Containment Temperature Response Following a LOCA-18 -

i

1

[. l Two Phase Flow Summary -

i l l No two phase flow at Wolf Creek for steady state conditions.  :

• Some delay in refilling system due to flashing of incoming water.

Insignificant effect on containment P/T response.

i i

I I9

[

[ Isolated Piping .(

i Overpressurization  ;

i .

• Preliminary scoping studyidentified 32 l potential locations.

• Worst case calculation indicates that at least one section of piping could exceed current stress

, allowables.

Operability evaluation prepared.

1 20

s- .

Isolated Piping .

Overpressurization Operability Evaluation

• Considered design basis functions of all locations

• Considered different size piping Considered different wall thicknesses Considered different geometries

• Determined that temperature rise is the controlling parameter

• Evaluated worst case temperature rise 21

i- !

Isolated Piping .

Overpressurization Operability Evaluation Details

• In all cases only design function is pressure boundary integrity. In no case must a section be ,

placed m service post LOCA.

• Calculation performed to determine if worst case causes pressure boundary breach.

22  ;

I

\ :i Isolated Piping Il Overpressurization Operability Evaluation Details (Continued)

Simple Volume expansion calculation

- How much must pipe volume increase to accommodate the expansion of water?

- For that volume increase, does material strain exceed material failure value?

- Material failure value taken from ASME Section II minimum elongation values.

23

l. I Isolated Piping Overpressurization Operability Evaluation Details (Continued)

• Assumes 35 F water initially ,

- Conservative since section is downstream of heat load.

• Assumes uniformly heats up to 290 F

- Conservative since maximum containment penetration surface temperature rises to 289 F at 145 seconds following an accident, then surface temperature begins to drop.

24 i

Isolated Piping [j Overpressurization 1

Results of Operability Evaluation:

• Maximum volume expansion of pipe would be 3%.

• Per ASME Section II material specifications, minimum elongation value for Carbon Steel (SA 106, Gr. B) is 14.5% (Wolf Creek worst case).

• Permanent deformation may occur but no pressure boundary loss.

• All sections determined to be operable.

25

Isolated Piping .

Overpressurization Evaluations Performed To Date

!

• Evaluating case specific susceptibility

. - initial temperature of water pressure reliefpaths

- administrative controls

• Three locations eliminated

• Five other potential eliminations Stress analysis in progress on three worst case locations 26 i

Isolated Piping .

Overpressurization Modifications i

i i'

• No modifications identified yet ,

Contingency modifications being prepared

• Target for implementing any necessary modifications Refuel IX, (Fall 1997) i 27 f

Overpressurization Summary

. Preliminary study identified 32 locations i

Case specific analysis has eliminated 3 @ossibly 83#

Stress analyses in progress Contingency modifications being developed Commitment to complete by Spring 1999 Target to complete by Fall 1997 0

28

! I'l Generic Letter 96-06 -

Issues

• All analyses are being performed to existing l criteria (current licensing basis).

• Possible request to utilize Appendix F criteria (eliminate unnecessary modifications).

• Possible that all-required modifications will not be completed by the time we have completed all other RefuelIX work.

29

1

• [ ,

l Enclosure 2 ATTENDANCE LIST l l

HEETING WITH WOLF CREEK NUCLEAR OPERATING CORPORATION' i GENERIC LETTER 96 06 ,

JUNE 25, 1997 A

Wolf Creek- Nuclear Ooeratina Corooration Jin-Shou Hseu R. Muench T. Garrett S. Wideman -

B. Selbe P. Parthasarathy

[ Altran Corooration T. Esselman M. Zwelgle Nuclear Reaulatory Commission J. Stone J. Fair

. G. Hubbard C. Saadu J. Tatum B. Wetzel

B. Westreich K.~Manoly 4

-C. Lauron

~

J l

. . . . . - . - - , . . . -