ML20027C543
| ML20027C543 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 10/05/1982 |
| From: | Bowen L COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20027C542 | List: |
| References | |
| NUDOCS 8210180146 | |
| Download: ML20027C543 (34) | |
Text
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- UNITED STATES OF AMERICA
-
- NUCLEAR REGULATORY COMMISSION
~
BEFORE TH ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
COMMONWEALTH EDISON COMPANY ) Docket No. 50-454
) 50-455 (Byron Station, Units 1 and 2) )
AFFIDAVIT OF LESLIE A. BOWEN I, Leslie A. Bowen, being duly sworn do hereby swear and state:
- 1. I prepared an affidavit dated June 6, 1982 which was filed in support of Commonwealth Edison Company's
(" Edison") motion for summary disposition, dated June 7, 1982, with respect to DAARE/ SAFE Contention 9a. This contention raised several issues concerning a phenomenon called " water hammer" and its potential effcct on Byron Station.
- 2. My June 6, 1982 affidavit set forth Edison's position with respect to the installation of temperature sensors on the bypass system piping to the steam generators at Byron Station. Specifically, it was stated that Edista planned to install such sensors (June 6 Affidavit, pp. 3-4).
The sensors would serve to alert operators of the potential existence of steam in the bypass system piping, a precursor of bubble collapse water hammer such as the recent occurrence at )
KRSKO. I I
C210180146 821014 PDR ADOCK 05000454 i G PDR '
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- 3. It is the purpose of this affidavit to set forth Edison's present position with respect to the use of temperature sensors at Byron Station. That position follows.
- 4. During the intervening uonths since filing my June 6 affidavit, I investigated further into the operating guidance provided by Westinghouse for the feedwater bypass I system. That guidance indicates that optimum operation of the feedwater bypass system requires constant feedwater flow through the upper steam generator nozzle. The existence of this flow, a matter not taken into account during the prepa-ration of the June 7 affidavit, would preclude the possibility of a water hammer event such as that which occurred at KRSKO.
Hence, temperature sensors would be unnecessary and therefore, Edison no longer intends to install temperature sensors on the
! feedwater bypass system piping at Byron Station. The NRC Staff has been informed of this development by the attached i
letter.
l / ,' C6 k Leslk6 h. owen State of Illinois (
County of Cook Subscribed and sworn to before I
me this js / day of October 1982,
! /
/*
/
,/'
u .( . . u - ,e 4 c c < t Notary Public
! My commission expires on My Commission Expires June 12.1986
! A 7 Commonwealth Edi:;rn
, 8, .bh One Fust Nahenal P: ara Chicago. Illinois
. . ' p .' AC::fe5s Reply 11 Fost Office Box 767 Chicago, minois 60690 September 9, 1982 l
l Mr. Harold R. Denton, Director t
Of fice of Nuclear Reactor Regulation l U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 Waterhammer Prevention NRC Docke t Nos . 50-45 4, 50-455,
, 50-456 and 50-457 1
Reference (a): July 30, 1982, letter from '
l i
- 8. J. Youngblood to L. O. De lGe o rg e .
(b): August 5, 1982, Memorandum from S. H. Chesnut to B. J. Youngbloo d summarizing July 27, 1982, meeting with Westinghouse.
Dear Mr. Denton:
This is to provide information requested in reference (a) regarding the design features and procedural controls related to prevention of waterhammer o f the type which apparently occurred a t the KRSKO plant in Yugoslavia in July, 1981.
Attachment A to this letter contains our responses to each of the questions contained in reference (a). Several of the questions requested information pertaining directly to the KRSK0 plant. Those responses have been prepared with the assistance of Westinghouse but, as indicated in the text and in the meeting discussed in reference (b), there still remains some uncertainty in the circumstances surrounding the KRSKO event.
It should be noted that in Commonwealth Edison's Motion for Summary Disposition of DAARE/ SAFE Contention 9(a) concerning water hammer events, L. Bowen stated in her affidavit that Edison would be installing temperature sensors per recommendations made by Westinghouse. During the intervening months since filing of that a f fidavit , further research into operating guidance provided for the l feedwater bypass systems has been performed which indicates that i
optimimum operation of this system requires constant feedwater flow through the upper steam generator nozzle. This flow precludes the possibility o f waterhammer such as that which occurred at KRSKO.
Hance, Commonwealth Edison no longer intends to install temperature sensors on the bypass feedwater line.
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l
. . H. R. Denton September 9, 1982 l I
Please direct further questions regarding this matter to this of fice.
One signed original and fifteen copies of this letter are provided for your review.
Very truly yours, I
Yk T. R. Tramm -
Nuclear Licensing Administrator l
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Question 1 o -
For the KRSK0 reported event, provide KRSK0 plant and operational information which describes:
a) Plant operating state . activities, (i.e. testing) underway at the time of the reported waterhammer event and the operating conditions f or the steam generator (s) f eedwater, auxiliary f eedwater, and tne f eedwater bypass systems.
Response
As a general comment, the exact time at which the incident occurred is not known. The damage was discovered several weeks af ter the Hot Functional Tests during which it is believed to have occurred. The conditions at I tne time of the incident cannot, therefore, be established with confidence. Based on the evidence, it can only be assumed that at some time, the necessary condition for bubb'e collapse waterhammer did exist.
Subsecuent to the incident, the KRSK0 plant feedwater system was modified to address the issue of steam generator (SG) preheater tube vibration.
The material wnicn follows is based on the system as it existed at the time of tne water hammer incident.
, Tne incident .i t believed to nave occurred during the plant Hot Functional i
Tests, more specifically, during tne tests of the Auxiliary Feedwater System (AFS) pumps conducted during July 1981. At tnis time, the steam generator pressure could have been as hign as 900 psia and the water temperature as nign as 532 0F. It was during the Hot Functional Tests tnat a nammer or bang was neard.
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. i i
During the course of the AFS tests, ambient temperature water was i introduced through the auxiliary nozzle. The flow path included a section of 6 inch bypass piping between the junction of the 4 inch AFS pipe and tne auxiliary nozzle.
l l
l Tne main f eedwater system was probably not in service when the incident is believed to have occurred.
(
Question 1(b)
I
( Details of damage discovered in August, 1981, how and when damage was detected, and wnat evidence (if any).of a water hammer occurring in July, 1981 at the time of stated occurrence.
Response
A flow diagram showing tne main and auxiliary f eedwater systems is p're s e n t e d in Figure 1. Isometric sketches of the loop 2 bypass piping inside containment and outside containment, up to the junction of tne Auxiliary Feedwater System, are shown in Figure 2 and 3, respectively.
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-- e o- _ ., - - ._. , _ - .- -
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l . . -3 Tne extent of the damage to the loop 2 bypass piping inside and outside containment, is presented in Table 1. In Figure 2, the numbers in ovals
( identify tne pipe hangers which were affected. On botn Figures 2 and 3, tne length in meters of the pipe sections are given. A bulge or blister was observed in the bypass piping, in the horizontal section, downstream of toe secondary shield wall, approximately midway' between the shield wall and elbow in Figure 2.
The bulged region was approximately six to eight inches long, witn the bulge located on tne top side of the pipe.
At the bulge, the pipe diameter was increased approximately one cuarter inen' .
Tne only indication of any change to loop I was the observation that tne bypass line had moved down seven to nine mm at the secondary shield wall
- penetration.
t Tne loop 2 bypass line and hanger damage was discovered during a final l
hanger inspection on August 7, 1981. Also at about tnis time, it was observed tnat the AFS pipe paint back to the AFS pumps was blistered or discolored indicating that hot water and/or steam had at some time been p re sent in the normally ambient temperature system.
Tne oasis for believing tnat a water hammer event did occur is the nature of tne damage and because a logical secuence of events can be postulated, Table 2, wnicn could lead to a waterhammer.
One point not included in Table 2 is that the control and isolation valves in tne discharge lines of the AFS pumps are normally open and,-
tnerefore, would not have prevented backleakage.
Question 1(c)
Plant corrective action (s) taken in terms of redesign, re p a i r., operator instruction or procedures, etc. for avoidance in the f uture. -
Response
Witn respect to~KRSK0 plant repair, the section of bypass piping, containing the bulge, between and including tne second and third upstream
- elbows was replaced. Also tne hanger damage was repaired.
Also with respect to plant operation, the customer was instructed to maintain tne steam generator water level above the top of the auxiliary feedwater discnarge pipe inside the steam generator as much as possible.
With tne discharge pipe covered, only hot water and not steam could leak back into the bypass and AFS piping, tnus greatly reducing the potential f or waternammer.
To reduce tne likelihood of backleakage, the Auxiliary Feedwater System l check valves which were known to leak, valves 11005, 11007, 11077 and 11079, on Figure I were refurbished. The check valves associated with the turoine driven AFS pump discharge were not inspected as a pyrometer cneck under not condition indicated no leakage.
In tne eventuality that the presence of steam is suspected in the bypass lines of all loops, based on temperature data and water level status and h i s t o ry , tnere would not be any operable loops with which to shut down tne plant. In this situation, the recommended course of action is to slowly refill one loop at a time with the AFS. An analytical study by the Westinghouse R&D Center shows th:t for the KRSK0 plant, the safe '
refilling flow rate is in the range of 15 to 123 gpm per steam generator. To be conservative, we recommend the 15 gpm valve or as close to this as can be provided.
Under normal conditions, between 0 and 100% power some flow is provided continuously tnrougn the auxiliary nozzle, thus effectively preventing the backflow of hot water or steam f rom the steam generator. When the feedwater flow is through the main nozzle, a tempering flow of one to two p e rc e nt is maintained through the auxiliary nozzle. The purpose is to maintain the auxiliary nozzle at feedwater temperature thus reducing the induced tnermal stresses when the feedwater is transferred f rom the main to auxiliary nozzles, during plant unloading, for example. However, the tempering flow also ef f ectively prevents backleakage.
At ve ry low load or not standby conditions, when the feedwater flow to l eacn steam generaor is minimal, the operator is instructed to supply the f eedwater continuously rather than intermittently. The reason is to minimize tne probability of feedline cracking, but it also eliminates the possibility of backleakage.
l .
c The principal modification was to provide a series of temperature measurements on the bypass piping of each loop in the first vertical leg i upstream of the auziliary nozzle. Two strap on RDT's have been installed in each loop, one three meters below the first elbow and a second four meters below the elbow. The RTD's are connected to the plant's DATA-SCAN Temperature Monitoring System which allows for printing out the temperature values in the control room on reouest. The' system activates an alarm if the temperature values exceed predetermined setpoints.
Recommended operating guidelines were provided to the customer for utilizing tne temperature data during the operations of Plant Heatup and Cooldown and f or Power Operation.
i Question 2 l
1 With respect to plant (KRSKO) operational states information (See Item 1), provide the following information:
1 1
(a) Water level in the steam generator relative to the auxiliary feedwater nozzle elevation.
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7 i
Response
During normal operation, the indicated water level is constant with power i level at 488 inches above the tube sheet. At full power, the actual water level is approximately four inches higher because of velocity head and circulation ratio effects. The di ferential increases and then decreases with power level; it is 0 inches at 0 power, approximately six inches at 50 percent power and approximately four inches at 100 percent power. Tne top of the auxiliary nozzle discharge pipe is 473 inches above the tube sheet. The water level is, therefore, nominally 15 to 21 inches above the top of the auxiliary nozzle discharge pipe, depending on power level. Accounting for normal channel accuracy of + five percent of sp an or + 12 inches, which is believed to be conservative, the water level could be as close as three inches above the top of the discharge pipe.
l Question 2(b)
Steam generator p ressure, temperatures, and flow rates.
Response
At the time when the waterhammer event igbelieved to have occurred, tne steam generator pressure could have oeen as high as 900 psia with a saturation temperature of 532 F.
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Question 2(c)
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Flow rates, temperatures and pressures in the f eedwater and auxiliary f eedwater sources supplying flow to the steam generator.
Response
t It is possible that the waterhammer incident occurred during testing of the AFS pumps. The pumps provide feedwater to the auxiliary nozzles l
t h ro ug h the four inch diameter AFS piping and a section of six inch diameter bypass piping. The capabilities of the motor-driven and turbine-driven AFS pumps are 350 gpm and 700 gpm, respectively, against a back pressure in the steam generator eouivalent to tne set pressure of the lowest set safety valves. The water would be at ambient temperature since the pumps take suction from the condensate storage tanks.
During the testing, the pumps were sta ted and stopped. The intervals between pump stop and restart were up to 30 minutes long. During those intervals, there was no f orward flow through the auxiliary nozzles.
The steam generator water level was not intentionally lowered below the auziliary nozzle discharge pipe prior to each test to accept the AFS water.
I l
Two isometric pipe sketches of the bypass piping are presented in Figures l
l 2 and 3. Figure 2 shows the bypass piping inside containment from the six inen auxiliary nozzle to the containment vessel penetration. Figure l
, 3 shows the section of bypass piping from the containment penetration l
back to the point where the f our inch Sch 80 AFS pipe connects.
Wit'h respect to the postulated backleakage path of hot water and/or l
steam, the pertinent valves are all.in the AFS. The AFS motor-driven pump discharge lines are each provided with two check valves and with a pneumatically operated, normally open, control valve which is operable locally and from the control room.
Tne AFS turbine driven pump, by means of a cross connect line, connects with the two motor-driven pump discharge lines. The flow path from the turbine driven pump to either SG auxiliary f eedline line includes two check valves and a pneumatically operated control valve as is the case f or the motor-driven pumps. In addition, each leg of the cross connect line is provided with a pneumatically operated isolation valve. The control and isolation valves are operable locally and f rom the control room.
- 2. Byron and Braidwood (B/B)
Figures 6 and 7 are simplified diagrams of the B/B AF and main feedwater systems. Detailed P&ID's accompany this response. The l following table lists the valves that are controlled from the l
l Control room.
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. -11 AF System
- 1. Suction Valves from the Essential AF00 7A- l &2 Auto open Service Water (ESW) System AF0078-l&2
- 2. AF Steam Generator Flow Control AF005A thru H Auto capability Valves
- 3. AF Steam Generator Isolation AF013A thru H Valves Feedwater System -
- 4. Feedwater Reg Vaive FWOO6A(B,C&D)
Isolation Valve
- 5. Feedwater Reg Valve FW510(520,530, Auto capability 540) i
- 6. Feedwater Reg Valve FW510A(520A, Auto capability 1 Bypass Valve 530A, 540A)
- 7. Feedwater Isolation Valve FWOO9A (B,C&O) Auto capability
Tne remainder of the valves shown in Figures 6 and 7 are either manual valves, automatically operated valves or locally operated valves.
Accompaning this response are one line diagrams that show the elevation of the upper f eedwater nozzle at the steam generator and its relation to the otner piping.
Question 4 .
Provide steam generator " internals" design details for botn the Byron and KRSK0 plants. Tnis information should be in the form of drawing which clearly define Feedwater and Auxiliary Feedwater nozzles, penetrations, flow distributors, etc. This information will be used to determine similarities and differences between the proposed Byron Steam Generators and KRSK0 Steam Generators.
Response
- 1. KRSK0 The internals of the KRSK0 Model 04 steam generator are shown in Figures 4 and 5. Figure 4 for the upper shell shows the auxiliary nozzle and internal extension. Also indicated is the actual water level for 100 percent power.
r In Table 3, the KRSK0 SG internal elevations and setpoints are tabulated.
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l 2. B/B Tne internals of the B/B models D-4 and D-5 are shown in figures 8 and 9 respectively. Figures 8 and 9 show tne auxiliary nozzle and interal i
extension and actual water level f or 100 percent power.
l
[
i l In Taoles 4 and 5 the B/B SG internal elevations and setpoints are l
tabulated.
I Question 5 Given the inf ormation reouested in Items (3) and (4), summarize similarities and dif f erences between Byron and KRSX0 f eedwater, auxillary feedwater, feedwater bypass systems, and steam generators. Relate to the Byron plant operational procedures and temperature sensor installation.
l
Response
Tne bypass line and AF system pipe routings are similiar f or the KRSX0 and B/B plants. Tne number of check valves in the bypass and AF lines are essentially the same with the exception of an additional check valve in the B/B piping at the upper steam generator nozzle. Valves in the AF .
system are normally open in botn plant designs.
4
B/B Operation In contrast to the KRSK0 start up operations tne B/B plants utilize a start up f eedwater pump f or heat up and start up rather than the AF l
system. Flow is through the feedwater reg. valve bypass valve, the l t
preneater bypass valve and the upper steam generator. nozzle. The l feedwater reg. valve bypass valves are automatically controlled to maintain steam generator level. Leakage through the feedwater reg. l valves is eliminated by closure of the upstream isolation valve. Hence, {
1 sufficient flow control is present to ensure flow at all times through tne upper nozzle. During hot standby conditions, SG level may be controlled by blowdown. Slug feeding of the steam generators is not recommended.
1 t
During power operation below a nominal 20% power, flow is directed to the upper nozzle by the preheater bypass valve. There is an automatic switchover to the lower nozzle at higher power levels. However, flow is maintained tnrough the upper nozzle via the tempering flow line.
Therefore, during plant operation flow is present at all times to the upper nozzle. With constant flow the conditions for backleakage, eitner of steam or hot water, into the AF lines are eliminated. Furthermore since the steam generator level is automatically controlled at all times, tne possibility of uncovering the nozzle in the steam generator is minimized which limits f urther the potential for backleakage of steam l
evon in aosence of flow.
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- I 1 . .
During operation, on-line testing of valves in the tempering flow and bypass feedwater lines is recommended by Westinghouse. If the testing reouires valve stroking, Westinghouse recommends that the valve not be closed longer than 30 seconds. This recommendation will be followed at By ron/B raidwood.
Duiing plant hot functional tests the same precautions apply. There should be flow through the upper nozzle at all times. If flow is interrupted for any reason it will be re-initiated slowly to ensure that, if steam back leakage into the piping has occurred, water hammer will not occur.
If flow is stopped and re-initiated for.any reason during testing, a physical inspection of piping and hangers will be undertaken to ensure that no damage has occurred due to water hammer.
Because of these operating procedures, the temperature sensors recommended by Westinghouse will not be installed on the B/B feedwater piping. These sensors and associated operating procedures will not add any significant margin of protection f rom water hammer due to the procedural reouirements to maintain flow through the upper nozzle at all times.
Westinghouse has f urther recommended that testing be performed to ensure there is no backleakage through AF system check valves. Procedures will be written to accommodate this recommendation.
'~
Byron plant operating and startup procedures have not been written as yet. Tne recommendations discussed above will be incorporated when these i
tests are written.
Question 6 Provide the test plan which would be used by Byron plant for verification I
that steam generator waterhammer will not occur (as noted in the staff's ,
! Byron SER Evaluation, Section 10.4.7). Describe how this test plan will l
j demonstrate that the Byron plant will not experience a waterhammer such as reported at KRSKO.
l
Response
{
First of a kind steam generator waterhammer tests have been conducted for the Model D4 Steam Generator at the KRSK0 plant. The tests verified that if tne f eedwater bypass system is operated in accordance with the Westinghouse recommended guidelines, SG waterhammer will not occur.
Startup test procedures for Byron /Braidwood will be written in accordance with tne guidance provided in NUREG/CR 1606. These tests, when written, will oe available for NRC revleuL.
Tne KRSK0 test plan did not include testing related to steam backleakage into tne bypass line and AFS. Any test f or this purpose would be directed toward verifying that necessary steps have been taken to prevent l
steam oackleaxage.
1.
As stated above, preoperational tests will be written to ensure that there is no backleakage through AF system check valves as r.ecommended by Westinghouse.
.R=kews
- 1. Docket Numbers STN 50-454, 50-455, 50-456, and 50-457, subject:
i Additional Inf ormation on Byron /Braidwood Waterhammer Prevention, i
B.J. Youngblood, Chief. Licensing Branch No. 1 Division of
. Licensing, NRC, July 30, 1982.
- 2. " Operating Procedures for Counter-Flow Preheat Steam Generator with Main Feedwater Bypass System", May, 1981, Westinghouse Electric Co rp o ra t i o n .
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RESULTs of EXAMINATION o LOOP 7,' INS {DECONTA!NHENT
- o EMBEDMENT PLATES 60VED, EpLT4 l.00SENED, PIPE CLAMPS l.00SENED ANDH0VED, (AFFECTED llANGERS ARE.jpENTIFIED BY NUNDERS IN OVALS), -
i o PIPE CilANGED E0 CATION O PIPE DULGED 1/31 INCll NEAR SECONDARY Siligtp WALL.
p LOOP 2, INTERHEDIATE EUILplNG ,
i o PIPE MOVEMENT WAS NEGLIGlBLE o
AUXILIARY PIPE PAINT WAS BLISTERED BACK TO Tile MDAFS PUNPS o HDAFS PUMP CilECK VALVES WERE FOUNn TO HAVE SOME NICKS, SCRATCliES, AND UNEVENNESS. .
l l .
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7bWe3 SEQUENCE OF EVENTS
[
t o
TilEiTEAMGENERATORWATERLEVELWASDEl0WTilE EXTENSIONDISCllARGEANDTilEREWASN0FEEDWATERFLOWTilR00611 (DURlWG Tile il0T puncTIONAL TESTING pf Tile AFS, Tile PUNPS WERE S RESTARTED),. -
t o
STEAM LEAKED HACK TilROUGil A SECTION dF Tile EY TilEAFSPIPINGTOTilEHDAFSPCMPSc..(TilEEXTENTOFTil INDICATED EY ELISTERED PAINT ON Tile Af4 PIPING),
i t
o FOR Tile LEAKAGE T0 llAVE OCCURRED, Tile TWO CilECK VALVES IN EA PutlP DISCllARGE EINES HUST llAVE DEEN EEAKING, '
g g $
r k g 4 _. g ._ ( C - ( c ? w d )
SEQUENCEOFEVENTS(CONTINUED) -
o WITil STEAM PRESENT IN Tile SYPASS WNE PIPING, THE AFS PUMPS ifERE STA BRINGlHG IN 001.D IfATER.
Tile 11ATER FAPIDI-Y CONDENSED Tile STEAM RESU IN WATERilAMMER, ',- _,
o AS AN0 tiler POSSIBILITY, STEAM LEAKED PACK TilR00Gli A HORIZONTAL SECTio EYPASS lINE OVER C0lJ) WATER AlREAI)Y PRESENT. AT SOME POINT A STEAM BUBBLE WASFORF1EDDUEToASURFACEDISTWRPANCEANDTilEBUBBLECONDENSEDBY WATERCARRIEDUPFROHTHEE0TT0HOFTilEPIPE, 9
A BAN 4 OR IW91ER WAS llEARD DURING THE fl0T FUNCTIONAL TESTING O
e
. . . f
TABLE 3
,. KRSXO 04 STEAM GENERATOR ELEVATION AND LEVEL DATA ELEVATION FROM TOP PERCENT OF ,
OF TUBE SHEET NARROW RANGE b
Upper NR Ts:;n 566" 100%
Main Deck Plata Top of Swiri Vine. .
. 542" Hi-Hi LeveT -
530*
Actual Water LeveT(/co *4 Bower ) 492"~ 68%
Indicated Water Level 488* . 66%'
Mid. Deck Plata:
Top of Auxiliary Nozzle Discharge 473" 60%
La Level 464."
Lo-Lo Laval ,
420" .
Lower Deckplata-Top of Tisa BundTe 336" Lower NR sep 333" 0%
Main Feedwater Nozzle d .
. e i ; I '
. . TABLE 4 BYRON 1 04. STEAM GENERATOR ELEVATION AND LEVEL DATA ELEVATION FROM TOP PERCENT OF 0F TUBE SHEET NARROW RANGE t
NAL. .
Upper NR Top 566" 100%
Main Deck Mate -
Top of Swirl Vane 542" Ht-Hi Lever ., ,
530" 85%
Actual Water Levar 6do 2. %f_M) ,
492" 687.
Indicated. Water Level. 488" 66%
Mid Deck Plate.
Top of Auxiliary Nozzle. Discharge 473" 60%
Lo. Level 440" 46%
Lo-La Level 420" 37%
i Lower Deckplate l
Top of Tube BundTe 336" .
2Le Lower NR liep 333" 05 Main Feedwater Nozz.le 4 l
- ---- -