ML20031F189

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Forwards List of NRC Sponsored Confirmatory Research & Technical Assistance Programs Re PWR Steam Generator Tube Integrity
ML20031F189
Person / Time
Issue date: 12/14/1977
From: Liaw B
Office of Nuclear Reactor Regulation
To:
Office of Nuclear Reactor Regulation
Shared Package
ML13319A640 List: ... further results
References
FOIA-81-313 NUDOCS 8110190323
Download: ML20031F189 (3)
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UNITED ST ATES 8 '-

h[/.I W ASNNG TON. D. C. 20555 NUCLEAR REGULATORY COMMisslON i

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DEC 1419/l HEMORANDUM FOR:

L. C. Shao, Chief Engineering Branch, 00R FROM:

B. D. Liaw Engineering Branch, D0R

SUBJECT:

NRC SPONSORED RESEARCH & TECHNICAL ASSISTANCE PROGRAMS RELATED TO PWR STEAM GENERATOR TUBE INTEGRITY Attached for your inforr.ation is a listing of NRC sponsored con-firmatory research and technical assistance programs related to PWR steam generator tube integrity.

Programs for the determination of the effects of tube ruptures such as off-site release, DNBR and fuel failures are also included.

f B. D. Liaw Engineering Branch Division of Operating Reactors cc:

V. Stello, D0R H. Conrad, DSS R. Mattson, DSS B. Turovlin, DSS

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D. Eisenhut, D0R J. Rajan, DSS l

J. P. Knight, DSS R. Landry, D0R i

J. T. Telford, DDR R. Gamble D0R R. Cudlin, D0R F. Odar, DSS l

J. Guibert, 00R l

R. Stuart, D0R F. Almeter. 00R L. Frank, OSD J. Muscara, RES

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i CONTACT:

B. D. Liaw /

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NRC SPONSORED CONFIRMATORY RESEARCH & TECHNICAL ASSISTANCE PROGRAMS RELATED TO PWR STEAM GENERATORS A.

CONFIRMATORY RESEARCH PROGRAMS i

1.

Steam Generator Tube Integrity Program Contractor:

Pacific Northwest Laboratory (PNL)

Principal Investigator:

M. Vagins NRC Program Manager:

J. Muscara, MMB/RES 2.

Stress Corosion Cracking of Inconel Tubing Test Program Contractor:

Brookhaven National Laboratory (BNL)

Principal Investigator (s):

D. Van Rooyen, J. Weeks NRC Program Manager:

J. Muscara, M48/RES j

3.

Standards for Material Compatibility and Eddy Current Inspection Contractor:

Battelle Columbus laboratory as subcontractor Principal Investigator (s):

J. Weeks (PNL), S. D. Brown (BCL)

NRC Program Manager:

L. Frank, SCSD/',3D B.

TECHNICAL ASSISTANCE PROGR/MS 1.

Corrosion, Coolant Chemistry, Chemical Cleaning and De-contamination in LWR plants Contractor: Brookhaven National Laboratory (BNL)

Principal Investigator (s):

J. Weeks, D. Van Rooyen NRC Technical

Contact:

F. M. Almeter, EB/ DOR /NRR 2.

Effect of Steam Generator Tube Plugging on Peak Clad Temperatures Following LOCA (just completed)

Contractor:

Idaho National Engineering Laboratory (INEL)

Principal Investigator:

E. Gruen NRC Technical

Contact:

R. Landry, RSB, DOR /'i?.R 3.

Effect of Steam Generator Tube Rupture on Peak Clad Temperatures Following Postulated Accidents (Proposed)

Contractor:

Idaho National Engineering' Laboratory (INEL)

Principal Investigator:

E. Gruen NRC Technical

Contact:

K. Landry, RSB/ DOR /NRR 9

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4 Statistical Assessment of Steam Generator Tube Degradation

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i and ISI Sampling Plan

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Contractor: Sandra Laboratories Principal Investigator:

R. Easterling NRC Technical

Contact:

R. Gamble, EB/00R/NRR 5.

Analytical Determination of Steam Generator Tube Integrity Contractor:

Brookhaven National Laboratory (BNL)

Principal Investigator:

M. Reich j

NRC Technical

Contact:

J. Rajan, MEB/OSS/NRR l

6.

LWR Water Chemistry / Corrosion t

Contractor: Brookhaven National Laboratory (BUL) j Principal Investigator:

J. Weeks NRC Technical

Contact:

H. Conrad/B. Turovlin, HTEB/ DSS /NRR 7.

Transient Analysis (Effects of Steam Generator Tube Ruptures)

Contractor: Brookhaven National Laboratory (BNL)

Principal Investigator:

M. Levine NRC Techaical

Contact:

F,. Odar, AB/ DSS /NRR l

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NOTE TO:

B. C. Buckley, Section Leader, Plant Systems Branch, 00R FROM:

S. D. MacKay, Plant Systems Branch, DOR

SUBJECT:

OCONEE POWER OSCILLATIONS During a visit to the Oconee power station on December 27, 1977, additional information was obtained regarding the nature and cause of reactor power oscillations on Unit No. 3 and the action taken by the licensee to reduce the magnitude of these oscillations.

The licensee had observed reactor power oscillations in Units 1, 2 and 3 at power levels between 50% and 75% of full power that had a frequency of 0.25 Hertz and a peak-to-peak amplitude of approximately 1.5% of full power. These oscillations were considered normal.

However, following the first refueling outage of Unit No. 3 in the Fall of 1976, these 0.25 Hertz oscillations gradually increased in amplitude from 1.5% in November 1976 to 7% in June 1977 (see Reference 1).

O In an effort to determine the cause of the oscillations and to

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eliminate them, the licensee gathered data on all parameters that were oscillating and checked all settings in the integrated control system.

Parameters that oscillate at 0.25 Hertz include: reactor power (out-of-core detectors), reactor coolant inlet temperature, reactor coolant outlet temperature, steam generator pressures, steam generator water levels, feedwater flows, feedwater pump pressures and reheater drain level. The only abnonnality noted in checking the process system parameters was that steam generator A provided 18"F less super-h0 heat than steam generator B and the least amnnnt of men, haat in 2nv kC a

1 e ca-th mugn steam generator.

It has beergminted out that this may be related to the fact that steam generatg= mkg the effects 0f~TheA ha IVE ubeswide rather than one tube wide in nues ted ccr.trclwem on r.ne oscillations, the power from the steam j

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i (i,. t from steam generator 8 to be 2*F hotter than the reactor coolant

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returning from steam generator A.

This unbalance stopped the oscillations but the settings wc e restored for normal operation.

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operation, the oscillations are not present. This and other information

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together with previous knowledge that the steam generator water level

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had a natural frequency of 0.25 Hertz led to the conclusion that the f

oscillations might be reduced by increasing the hydraulic resistance

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of the downcomer region of the steam generator. There is an adjustable

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,f orifice in the steam generator for this purpose.

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p-a n ma It was planned that during the October 1977 shutdown, the orifice plate i

position would be checked and if it were 2 1/2 inches open, the came as Units 1 and 2, it wc.uld be completely closed..During the shutdown bh't it was found that the orifice plate was in the expected position.

2 1/2 inches open. However, attempts 'to close the orifice failed as

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the plate could be moved only 1/4 to 1/2 inches toward the closed i

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It was secured in that position, between 2 and 21/2 inches %

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After refueling, the plant was started up on December 3rd and the amplitude of the power oscillations was less than one percent. However,,

after operation for three days at 75% power it was apparent that the L

  • oscillations were increasing in amplitude. Figure 1 snows that on i

December 18 while operating at approxic stely 75% power, the peak-to- /"

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peak amplitude of the power oscillations was 2.2 percent of full power.

QcmMnhof the phenomenon leads to the following

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qualitative understanding. The oscillatory tendency originates in the steam generator at the natural frequency of oscillation of the water levels in the downcomer and tube bundle regions. This effect may or rray not be enhenced by the process system or the control system.

In either case, a change in the water level in the tube bundle will; cause a change in the effective heat transfer area coupling the reactor coolant system to the steam system resulting in a change in the rate of heat removal from the reactor coolant and thereoy causing a change in the temperature of the coolant returning to,the reactor

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core. The oscillating water level therefore causes the ' reactor water-temperature to oscillate and results in reactivity c5anges that cause the reactor power level to oscillate. These temperature and power level changes also result in the oscillation of the reactor outlet temperature but it is not clear whether this temperature oscillation in '

turn significantly influences the water level in the tube bundle region.

The power level oscillations were minimal %en the reactor was first placed into operation after refueling because at that point in core life the absolute magnitude of the temperature coefficient of reactivity was relatively small, i.e., about 1/5 of that at the end of the previous cycle. However, as the fuel burned up and baron was removed from the moderator, the temperature coefficient of reactivity increased and thereby caused the reactor power oscillations to increase in amplitude. Thus, if no corrective action is taken',- the amplitude would be expected to increase throughout ' core life and become approximately the same as that at the end of the previous cycle.

It was also noted that the power level foscillations are measured on l

the out-of-core neutron detectors whereas the in-core detectors e

indicate no oscillation. There is no in-core indicatio_n of power s/

oscillation because the in-core pows _r_ level,_sjgnals are generated

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, by the activation and decay of rhodium powered detectors tnat have

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an effective response time of several minutes and therefore du.not respond to a 0.25 Hertz oscillation. The out-of-core detectors I

respond quickly to neutron flux level changes and therefore follow the 0.25 Hertz oscillations that could arise from core power level

~ changes or changes in the density of the water in the reactor vessel downcomer between the core and the detectors. A calculation of the

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changes in neutron attenuation due to the oscillating temp vature of the incoming water compared to an estimate of the 4 activity effects of that temperature oscillation shows that the reactivity effects are predominant and therefore the out-of-core detectors indicate an actual power level oscillation.

3 A good deal of ~ additional data and some testing are necessary to confirm N

- our present understanding of the phenomenon and allow us to quantify

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some of the effects that have been observed.

The licensee is continuing its evaluation of this matter, devising special tools to permit movement of the orifice plates in the steam gentvators and developing further tests of the instrumentation and control s.

Additional data will be obtained within the first 28 effective full power days when the reactor is at 757. power for the correlation test of the in-core and out-of-core neutron detectors.

t is expected that the amplitude of the oscillations, will not exceed the value previously experienced, and there is apparently no significant safety consideration at that value. However, it is recomended that the licensee be encouraged to obtain additional data and a better quantitative evaluation of this phenomenon so that the oscillations may be eliminated or reduced to an insignificant level.

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S. D. MacKay, Plant Syst s Branch

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Division of Operating Re ctors

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, References

[1'. <" Trip Report - Oconee Power Oscillations" 10/31/77 R. Woodruff, IE.

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., Report," August 1969.

'3.f BAW-10002, Supplement 1, June 1970.

'z-Enclosure Figure 1 --Oconee Unit 3 Power Level Oscillations j

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V. Stello D. Eisenhut W. Butler F. Jape IE F. Long IE R. Woodruff, IE J. Sniezek IE K. Seifert IE N. Moseley, IE R. Baer M. Mendonca J. Buzy, OLB i

P. Collins, OBL A. Schwencer I'

D. Neighbors S. MacKay L. Beltracchi

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Response of Resistance Temperature Detector (RTD) to 0.2:i Hertz

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temperature oscillations.

On January 18, 1977, I learned from a B&W employee that the response time of the RTDs was three to four seconds but might be as high as five seconds. Using four seconds as the time constant, we obtain (0.'25) 4 wT=2 w

= 6.28 radians A calculation of the response of this sicole lag system shows an attenuation of 16 dB and a phase lag of 81'.

TnisshowsthattheRTDelement.will experience a temperature oscilla' tion that is 160 as great as the oscillation in reactor coolant temperature at the 0.25 Hert:

frecuency. Thus the indicated temperature oscillation of 0.18'F (peak to peak) observed on December 18,1977 at 10 am.

reflected an actual temperature oscillatien of 0.18/0.16 or 1.13*F.

2.

Reactor Response to Oscillation of Inlet Water Temperature The transfer function of a PWR is fairly flat in the region er 0.25 Hert: and the reactor should respond with nearly full amplitude to reactivity oscillations at this frequency. Since the fuel temperature oscillations will be very small, reactor power will respond to T inlet directly. Thus the power level response is estimated to be approximately nK T

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JAN 2 7 573 Thus a 1*F rise in downcomer water temperature will cause the neutron detectors to indicate a 0.3% power level increase.

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This accounts for part of the amplitude of the oscillations.

4.

Phase Relationship Between Power Level and Downcomer Density The effective transport time from the downcomer to the core is approximately one and a half seconds. The power level will lag the reactivity change slightly and the total lag time between a downcomer density increase and a power level rise will approach two seconds. With a two-second lag, the power level will be 180* out of phase with the downcomer density, i.e., when the power level is maximum, the density will be minimum and there-fore the two effects will be additive to produce the maximum neutron level reading.

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F s u 2 7 1978 The moderator temperature coefficient -ff-for cycle 3 has been reporteo to be -0.53 x 10-4 Beginning of cycle and

-2.55 x 10-4 at end of cycle. Thus if the temperature coefficient werei.2x 10-4 on December 18 when the temperature oscillation was 1.13*F, the resultant power level oscillation should have been about 1.4 percent power. The data shows however, that the power level oscillations were approximately 2.2% power. A temperature power transfer coefficient of 1.8% power per 'F would be needed to explain this behavior.

Although this calculation is neither sophisticated nor very accurate, it does indicate that the power level oscillations observed could result directly from the inlet temperature

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oscillations.

3.

Neutron Detector Response to Downcomer Water Temperature Oscillations The downcomer annulus contains a layer of water 81/4 inches (21 c=)

thick between the reactor core and the neutron detector. For this thickness of water, the effective attenuation length will be approximately 10 cm (3 meV neutrons) and the change in attenuation per l'F will be dN 1

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p dx =

p7 f = - 0.0014 per 'F at 550*F at 2000 psi wnere' ff = 2 x 0.0014 = 0.0030 per 'F

= 0.3 percent per 'F w_

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WASHINGTON. D. C. 20555 FEB 01 '978-MEMORANDUM FOR:

D. Eisenhut,-Assistant Of rector for Operational Technology, 00R FROM:

L. C. Shao, Chief. Engincering Branch, 00R

SUBJECT:

OCONEE UNIT 2 STEAM GENERATOR LEAK On October 7,1977 Oconee Unit 2 was shut down to locate a 0.2 GPM tube leak in steam generator B.

The subsequent investigation failed to locate the leak and the unit was returned to 70% power on October 27, 1977. After returning to power operation, the leak developed again at a stable rate of 0.08 GPM until November 3,1977 when the leak rate increased to 0.65 GPit during an attempt to increase power.

The unit was again shut down and a second unsuccessful inspection for the leak was made. On November ll,1977 the unit returned to power and operated at reduced power until December 28, 1977 when it was shut down to repair a failed stator in a control rod drive.

On January 10, 1978 it was reported that the source of the leak was discovered to be a crack in lane tube C77-T25. The crack was

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inspected using fiber optics and was described by the licensee as a 90' circumferential crack located at the upper tube sheet.. The tube was plugged ano 2he unit returned to power. There is currently no detectable leakage in the unit.

A similar occurrence in July 1977 involved tube C77-T2 in steam generator 3-B.

This tube was also a lane tube which showed a 60*

to 90' circumferential crack. Although this crack had not propa-gated fully around the tube in a short time as was the case in other lane tubes it was felt that this was a phenomenon affecting only peripheral tubes where lower cross flow velocities supplied le:s j

kinetic energy for crack propagation.

i The recently plugged tube, C77-T25;however, is not a peripheral tube but is located near the middle of the open tube lane in an area of higher cross flow velocity. However, since the unit was operating at reduced power levels (less than 70%) the flow velocity and in turn the energy available to drive the crack was substantially i

decreased. Therefore the crack would tend to stabilize at reduced power levels. We believe that this recent event is consistent with past experience for lane tubes and is not a new phenomenon.

Further-more, close monitoring of the leakage rate in conformance with the s

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FEB 011978 technical specifications will continue to ensure safe operation of the unit.

I L. C. Shao, Chief.

Engineering Branch Division of Operating Reactors

Contact:

J. Strosnider 49-28060 t

cc:

V. Stello J. Guibert R. Cud'in R. Ma

.on J. Knight S. Pawlicki R. Bosnak B. Turov11n R. Stuart B. D. Liaw e

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