ML13331B133
| ML13331B133 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 02/17/1989 |
| From: | Baskin K Southern California Edison Co |
| To: | |
| Shared Package | |
| ML13331B132 | List: |
| References | |
| NUDOCS 8902270295 | |
| Download: ML13331B133 (23) | |
Text
Enclosure 1 BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA EDISON
)
COMPANY and SAN DIEGO GAS & ELECTRIC COMPANY )
for a Class 104(b) License to Acquire,
)
DOCKET NO. 50-206 Possess, and Use a Utilization Facility as
)
Part of Unit No. 1 of the San Onofre Nuclear )
Amendment No. 165 Generating Station
)
SOUTHERN CALIFORNIA EDISON COMPANY and SAN DIEGO GAS & ELECTRIC COMPANY, pursuant to 10 CFR 50.90, hereby submit Amendment Application No. 165.
This amendment consists of Proposed Change No. 207 to Provisional Operating License No. DPR-13. Proposed Change No. 207 adds license condition 3.M, "Cycle X Thermal Shield Monitoring Program."
Proposed Change No. 207 is being added to require that a monitoring program be established for the reactor thermal shield during Cycle X of plant operation.
In the event of conflict, the information in Amendment Application No. 165 supersedes the information previously submitted.
Based on the significant hazards analysis provided in the Description of Proposed Change and Significant Hazards Analysis of Proposed Change No. 207, it is concluded that (1) the proposed change does not involve a significant hazards consideration as defined in 10 CFR 50.92, and (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change.
890.(:2270:29:j5 890217
(
PDR DOCK 05000206 P
-2 Subscribed on this /7 day of
, 1989.
Respectfully submitted, SOUTHERN CALIFORNIA EDISON COMPANY By:
Kenneth P. Baskin Vice President Subscribed and sworn to before me this
/7~L day of
/9e9' OFFICIAL SEAL AGNES CRABTREE Notary Public-Califomnia LOS ANGELES COUNTY My Comm. Exp. Sep. 14. 1990 Nota Public in and for the County of Los ngeles, State of California Charles R. Kocher James A. Beoletto Attorneys for Southern California Edison Company By:
J s A. Beoletto
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of SOUTHERN
)
CALIFORNIA EDISON COMPANY
)
and SAN DIEGO GAS & ELECTRIC
)
Docket No. 50-206 COMPANY (San Onofre Nuclear
)
Generating Station Unit No. 1
)
CERTIFICATE OF SERVICE I hereby certify that a copy of Amendment Application No. 165 was served on the following by deposit in the United States Mail, postage prepaid, on the 17th day of February
, 1989.
Benjamin H. Vogler, Esq.
Staff Counsel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 David R. Pigott, Esq.
Samuel B. Casey, Esq.
Orrick, Herrington & Sutcliffe 600 Montgomery Street San Francisco, California 94111 L. G. Hinkleman Bechtel Power Corporation P.O. Box 60860, Terminal Annex Los Angeles, California 90060 Michael L. Mellor, Esq.
Thelen, Marrin, Johnson & Bridges Two Embarcadero Center San Francisco, California 94111 Huey Johnson Secretary for Resources State of California 1416 Ninth Street Sacramento, California 95814 Janice E. Kerr, General Counsel California Public Utilities Commission 5066 State Building San Francisco, California 94102
-2 C. 3. Craig Manager U. S. Nuclear Projects I ESSD Westinghouse Electric Corporation Post Office Box 355 Pittsburgh, Pennsylvania 15230 A. I. Gaede 23222 Cheswald Drive Laguna Niguel, California 92677 Frederick E. John, Executive Director California Public Utilities Commission 5050 State Building San Francisco, California 94102 Docketing and Service Section Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ms A Beol ett6
e DESCRIPTION AND SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS OF PROPOSED CHANGE NO. 207 TO THE PROVISIONAL OPERATING LICENSE DPR-13 This is a request to add license condition 3.M, "Cycle X Thermal Shield Monitoring Program."
Reason for Proposed Change During core reload for Cycle X operation, the thermal shield was inspected to determine the condition of the flexures, keys and supports.
The inspection was conducted after degraded conditions were discovered at another plant of similar design.
The results of the inspections, analyses of the results and details of a proposed monitoring program are included in the Attachment 2 Report, "Engineering Evaluation of the SONGS 1 Thermal Shield Supports," WCAP 12148, dated February, 1989 by Westinghouse. This proposed change will add a license condition which in turn will require a program to monitor the condition of the thermal shield to confirm acceptable operation during the next cycle of operation.
Existing Condition None Proposed Condition See Attachment 1 Significant Hazards Consideration Analysis As required by 10 CFR 50.91(a)(1), this analysis is provided to demonstrate that a proposed license amendment represents a no significant hazards consideration. In accordance with the three factor test of 10 CFR 50.92(c), implementation of the proposed amendment was analyzed using the following standards and found not to:
- 1) involve a significant increase in the probability or consequences for an accident previously evaluated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety. Following is a summary of the details provided in Attachment 2 and a significant hazards analysis based on the evaluation.
-2 INTRODUCTION The thermal shield at San Onofre Unit 1 is supported at the bottom by six support blocks, which are attached to the core barrel by a system of bolts and dowel pins, and at the top by one flexure and four displacement limiters. During the November, 1988 refueling outage at San Onofre Unit 1, degradation of the thermal shield support block fasteners was discovered.
The six thermal shield support blocks are located at 0, 60, 120, 180, 240, and 300 degrees. Figure 1 shows the thermal shield support design.
Figure 2 is a drawing of the thermal shield support block details and the attachment of the 5 bolts and 4 dowels in each of the support blocks.
The dimension of the center bolt is 3/4" X 2 3/4", the top bolts are 7/8" X 5" and the lower bolts are 1/2" X 2 3/4". In between the upper and lower bolts, there are two 3/4" dowel pins and at the lower bolts elevation, two additional 3/4" dowel pins which all act to limit the relative vertical motion between the thermal shield and the support block. The dowels limit the bending and shear loads experienced by the bolts.
The flexure between the top of the thermal shield and the core barrel is designed to provide tangential and radial restraint to the top of the thermal shield. There were originally six flextures at locations 210, 850, 124', 2050, 244, and 325*.
By 1978, it was discovered that all flexures except 1 were cracked.
There are four upper displacement limiters which are located roughly 9.5 inches below the top of the shield. These displacement limiters are located at 0, 90, 180 and 270 degrees and initially limited the relative radial motion between the barrel and thermal shield at these locations.
The purpose of this evaluation is to evaluate the safety implication of San Onofre Unit 1 start up and continued safe plant operation for the next fuel cycle, cycle 10 (approximately 18 months), with known degradation of the fasteners at the 0* and 2400 blocks, and assumed fastener degradation at the 300* thermal shield support block.
BACKGROUND The visual inspection at San Onofre Unit 1 consisted of core barrel exterior and interior exams.
The inspections were performed utilizing remote camera equipment with the core barrel installed in the vessel.
The scope of the inspection and the results are summarized below.
The core barrel exterior examinations were performed through the four core barrel lift rig holes which allowed viewing of the following between the thermal shield OD and vessel ID:
Five of six flexures at the following locations:
21,
- 1240, 2050, 2440 and 3250.
The 124* flexure was intact while the other flexures were known to be cracked or broken from previous inspections (1978 and earlier).
Four of six support block exteriors at 60, 1200, 240* and 3000
-3 Five of eight specimen holder exteriors at locations 29*, 33.5*,
130.50, 1500 and 258.30 Three of eight specimen holder tube interiors at -100, 1500 and 258.30 Flexure bolting to the core barrel at 1240 The inspection results of the core barrel exterior were:
Cracked tack welds on bolt locking bars at the 2400 block (Bolts 1 and 2, Figure 2); Cracked lock weld on one dowel pin weld at 2400 block 1240 flexure remained intact and functional - in addition, no signs of damage to bolts on the core barrel Five other flexures still in place even though cracked No protruding bolts or dowel pins on the four support blocks visually inspected No evidence of rubbing or binding on the five specimen holders (no excessive movement) either exterior or interior The core barrel interior exams involved temporarily removing fuel assemblies to create inspection openings through the lower core plate.
The inspection covered the following:
All six support blocks, both upper and lower fasteners The bottom of the vessel Flexure bolting The inspection results were:
Thermal Shield Support Blocks Block Component Condition 00 Center Bolt Inboard appx. 15 threads (1 1/2")
Top Long Bolts OK Lower Bolts OK Lower Dowels OK 600 Center Bolt OK Top Long Bolts OK Lower Bolts OK Lower Dowels OK
-4 120*
Center Bolt OK Top Long Bolts OK Lower Bolts OK Lower Dowels OK 1800 Center Bolt OK Top Long Bolts OK Lower Bolts OK Lower Dowels OK 240*
Center Bolt Inboard appx. 1/2-3/4" Top Long Bolts Right inboard 10-12 threads (3/4-1")
Left -
OK Lower Bolts OK Lower Dowels OK 3000 Center Bolt OK Top Long Bolts OK Lower Bolts OK Lower Dowels OK No support block hardware in the bottom of the vessel Flexure bolts intact A visual inspection of the bottom of the reactor vessel, with the camera inserted at 60*, 120, 2400 and at the center through the lower core plate revealed no support block hardware. Based on the inspection results, engineering analyses were done to demonstrate that operation of the unit does not pose a safety problem for the next cycle.
Some potential safety consequences of thermal shield support failure are:
- 1. The generation of loose parts in the Reactor Coolant System (RCS)
- 2. Effects of the thermal shield on the core barrel
- 3. Reduction of core flow ENGINEERING ANALYSES This analysis is based on the following analyses and resulting conclusions that San Onofre Unit 1 can operate safety with degraded thermal shield support block fasteners for the next fuel cycle.
All engineering analyses are contained in WCAP-12148 (Westinghouse Proprietary Class 2) and WCAP-12149 (Westinghouse Proprietary Class 3),
"Engineering Evaluation of SONGS 1 Thermal Shield Supports", February, 1989.
- 1. Analysis Demonstrating Acceptability for 18 Months a) Prior to the outage, bounding analyses were performed to assist in establishing inspection criteria and to identify areas of possible degradation.
-5 Analyses were performed to determine what further degradation might occur and its acceptability. The analyses were performed to model the current observed condition of the thermal shield lower support blocks. Analyses were also performed to estimate the expected progression of the thermal shield degradation over the next fuel cycle.
The analyses concluded that the current estimated worst condition of all bolts and dowel pins failed at the 0', 240* and 3000 blocks is not expected to degrade the remaining blocks at the 60*, 1200 and 180* locations over the next 18 months of operation. This is provided that the remaining flexure does not fail.
The likelihood of failing this flexure in the next 18 months is very remote considering that it has remained operable for approximately 12 EFPY and that the vibration stress induced from the relative vibrations between the core barrel and thermal shield is lower now (with three degraded blocks) than before when no blocks were degraded.
b) Evaluations were done for the effect of normal, vibratory, thermal and seismic loads on the structural integrity of the thermal shield. As a result of these analyses, the integrity of the support blocks, the 124* flexure, and the core barrel is maintained over the next 18 months of operation.
c) Substantial further degradation of the thermal shield support system is not expected to occur for the next 18 month fuel cycle since the flexure is expected to remain intact and remaining loads on the support block fasteners are equal to or less than they have previously experienced. Nevertheless, the absolute worst case is assumed that the 1240 flexure and all fasteners for all six blocks are broken. In this unlikely case, the thermal shield remains stable without encountering conditions of undamped instability. Moreover, with the fasteners broken and the blocks intact, the thermal shield can withstand the 0.67g DBE seismic event without dropping. As a result, this worst case will not impact the safe operation of the plant.
d) Thermal shield tilting or falling are judged to be not credible since it has been demonstrated by analysis that the shield will always be adequately supported even in the case of fastener failure in all of the support blocks.
This is due to the design of the block ledges on the core barrel and the blocks fixed in position between the core barrel and the thermal shield.
- 2. Consequences of Postulated Failure of the Thermal Shield Support System a) Evaluations were performed to evaluate structural consequences of tilting or dropping of the shield. The worst possible scenario that could be postulated is a complete failure of the support system such that the shield either tilts because of a loss of some of the supports or the shield drops because of a loss of all the lower supports. The analysis of the above case has shown that this is not a credible scenario. Nevertheless, an evaluation at normal operating conditions of the structural
-6 effects of this worst conceivable case has been performed and an estimate has been made of the resulting effects on flow and pressure drop as well as core inlet flow distribution. The primary motivation for performing this evaluation was to ensure that even in this incredible case, the dropped shield would still be restrained by the lower radial supports and that normal core flow would not be significantly reduced and the analysis supported this conclusion. The design of the supports ensures that the blocks will remain in place and will continue to support the thermal shield.
b) A dropped thermal shield has been evaluated for flow distribution effects and has been determined to result in a minor reduction of flow in peripheral assemblies and an increase in center assemblies. This redistribution does not affect core limits. In the case of a titled thermal shield, there is an insignificant reduction in core flow. This is further addressed in Section 9.1 of Attachment 2.
- 3. Implementation of a Monitoring Program The unit can operate safely with the postulated credible worst case degradation of the thermal shield supports. Confirmation that the thermal shield does not undergo further degradation will be established by the implementation of internals vibration monitoring. This monitoring program is discussed in more detail in Section 10 of Attachment 2.
Further supporting documentation for the concept of neutron noise monitoring is included as Attachment 3 The noise signals from the ex-core, power range flux detectors will be monitored to detect changes in internals vibration. Accelerometers will be mounted on the exterior surfaces of the reactor vessel to detect loose parts that could be generated due to further degradation or further disengagement of presently degraded fasteners.
a) Internals vibrations will be monitored using the signals from the ex-core, power range detectors. Information will be used to detect significant thermal shield degradation.
b) The thermal shield Loose Parts Monitoring Program is intended to monitor the potential degradation of thermal shield fasteners or fragments in the reactor vessel. It is anticipated that sequential fastener failure (if it occurs) has the potential to result in an increasing number of loose parts collected in the reactor vessel.
- 4. Inspection of Bottom of Vessel/Loose Parts Analysis A visual inspection of the bottom of the reactor vessel, with the camera inserted at 600, 1200, 240* and at the center through the lower core plate revealed no support block hardware. Since there are at least three broken support block bolts which are in the process of rotating out of the core barrel, loose pieces of bolt originating form the thermal shield support will tend to fall and be carried by the downward inlet flow into the lower plenum. It is expected that these fragments would collect in the bottom of the vessel in the low flow areas. In addition, the
- e
-7 possibility that the thermal shield vibration may result in the generation of thermal shield loose parts in the Reactor Coolant System (RCS) has been evaluated based on another Westinghouse plant. Based on thermal shield loose part analyses completed for objects of similar geometry/size, etc. for a similar Westinghouse plant, operation with these parts present in the RCS is judged not to significantly affect plant safety. It is also possible that loose parts could fall out and remain trapped in the region between the lower support plate and lower core plate. Analyses have shown there would be no adverse effect on the rest of the plant due to loose parts if this were to happen.
ANALYSIS Conformance of this proposed change to the standards for a determination of no significant hazard as defined in 10 CFR 50.92 (three factor test) is shown in the following:
- 1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
RESPONSE
Operation of the facility in accordance with this proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated. This change requires the implementation of a monitoring program to confirm continued operability of the reactor thermal shield. Considering the results of the analysis contained in the attached report, the probability of those events previously evaluated will not increase.
The thermal shield will stay in place and will not cause damage to the internal structure that will initiate a LOCA. Also, the broken bolts or backed-out dowel pins that drop into the bottom of the vessel will stay in the bottom of the vessel and not cause any significant fuel damage. Note that the material for the bolts and dowel pins is ductile, not brittle, so that the bolts and dowel pins will not break into small pieces that are small enough to lodge between the fuel pins. Most likely, the broken bolts and dowel pins will stay in the low velocity zone of the pressure vessel (i.e., the center bottom of the pressure vessel head) until they can be recovered in the next refueling cycle.
The consequences of accidents previously evaluated will not increase because the thermal shield is designed to protect the pressure vessel from neutron induced embrittlement and the engineering analysis has shown that the thermal shield will remain functional for the next 18 months.
The consequences of any pressurization event analyzed in the UFSAR (such as loss of feedwater flow, feedwater line break, loss of reactor coolant flow) will not increase. Also, the thermal shield will stay in place and not cause any obstruction of flow. Therefore, the consequences of those events previously analyzed in the UFSAR that are affected by core flow and its distribution will not increase.
-8
- 2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
RESPONSE
A new or different kind of accident from any previously evaluated will not be created since engineering analysis shows that the "worst expected case" is that the fasteners for the blocks (60*, 120*, 180*)
and one flexure are still intact. Furthermore, with the existing design characteristics of the blocks (ledging on the core barrel and fixed in position between the core barrel and thermal shield), the "worst postulated case" is that all fasteners on six blocks and all flexures are broken, whereas all blocks are locked between the core barrel and the thermal shield. In both cases, the thermal shield still fulfills its original function. Therefore, no new event will be created.
- 3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
RESPONSE
The margin of safety will not be significantly reduced since no safety limit will be changed or affected by the thermal shield.
Further, since the thermal shield will continue to perform its designed function, no safety margin is degraded.
CONCLUSION Based on the above analysis of the degraded thermal shield support system at San Onofre Unit 1, it has been determined that no significant hazards consideration exists per 10 CFR 50.92. It is expected that San Onofre Unit 1 can start up and continue safe plant operation for the next 18 month fuel cycle, cycle 10.
Safety and Significant Hazards Determination Based on the safety evaluation it is concluded that: (1) the proposed change does not involve a significant hazards consideration as defined by 10 CFR 50.92; and (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Environmental Statement. - Proposed License Condition 3.M - WCAP 12148, "Engineering Evaluation of the SONGS 1 Thermal Shield Supports," February, 1989 - "Supporting Document to Section 10 of WCAP 12148, Internals Vibration Monitoring Using Neutron Noise."
SPECIMEN *L*E SPECIMEN TUBE SPECIMEN BASKET EXPANSION JOINT FLEXURE FIXTURE CTYP 6 PLACES)
LIMITER KEY cTYP 4 PLACES)
THERMAL SHIELD SUPPORT 3LDCK CTYP 6 PLACES)
FIGURE 1
THERMAL SHIELD Section B-B Section A-A Thermal Shield Core Barrel upport Block FIGURE 2 SONGS I Thermal Shield Support Block Design
ATTACHMENT 1
3.M Cycle X Thermal Shield Monitoring Program The neutron noise/loose parts detection system shall be used to monitor the condition of the reactor vessel thermal shield during Cycle X. Periodic monitoring of both neutron noise and loose-parts vibrations confirms that no long term unacceptable trend of degradation is occurring. The details of this program are described below.
(1) Acceptance criteria for neutron noise/loose parts monitoring shall be established by performing baseline evaluations within 90 EFPD of full power operation following return to service for Cycle X operation.
(2) The neutron noise/loose parts monitoring system shall be OPERABLE in MODE 1 with:
a) At least two horizontal loose-parts detectors monitored for at least five (5) minutes each day; and, b) at least three (3) neutron noise inputs monitored for at least five (5) minutes once a week.
(3) The data provided by the loose part/neutron noise monitor shall be analyzed once per week and compared with the established criteria. If the data exceeds the acceptance criteria:
a) Within 1 day the NRC will be informed of the exceedance.
b) Within 30 days the conditions will be evaluated and a report provided to the NRC documenting future plans and actions.
(4) Each channel of the loose-part detection system shall be demonstrated OPERABLE in MODE 1 by performance of a:
a) CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> b) CHANNEL FUNCTION test at least once per 31 days The surveillance requirements for neutron noise monitor are covered by the Appendix A Technical Specification 4.1.1 for the Power Range Neutron Flux.
(5) With the neutron noise/loose-parts detection instrumentation inoperable for more than 30 days, prepare and submit a Special Report to the commission pursuant to Appendix A Technical Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the system to operable status.
(6) In the case of a seismic event of 0.25g or greater as indicated on site sensors, loose parts/neutron noise monitoring will be conducted immediately and a controlled shut down shall be initiated. Before operations are resumed, its will be demonstrated that no thermal shield damage has occurred due to the seismic event.
(7) The provisions of Appendix A Technical Specification 3.0.4 are not applicable to this license condition.
ATTACHMENT 3
Supporting Document to Section 10 of WCAP 12148 Internals Vibration Monitoring Using Neutron Noise
- 1) Neutron Noise Technique The core barrel and thermal shield vibrate in beam and shell modes.
Core barrel beam and shell mode shapes are illustrated in Figure 1.
The shell modes can have various shapes. The dominant shell mode, n=2, has an oval shape as shown in Figure 1. Beam and shell modes have been detected by treatment of the ex-core power range flux detector signals.
The neutron flux signal is composed of a DC component and a fluctuating or noise component as illustrated in Figure 2. The noise component includes content that reflects internals vibrations. As shown in Figure 1, motion of the core barrel changes the downcomer water gap.
This results in changes in the signal level of the ex-core detectors. Frequency spectra of the signals show peaks associated with the vibration modes of the internals. Beam modes can be distinguished from shell modes by the phase difference between cross-core detector pairs; the phase difference for beam modes is 180 degrees, the phase difference for shell modes is 90 degrees.
Small motions of thermal shields have been detected using the neutron noise technique. Figure 3 shows in the ex-core neutron noise data from a Westinghouse plant with a thermal shield supported at both ends. The peak centered on approximately 12 Hz has been identified as a shell mode (1).
N = 2 shell modes of the thermal shield have been identified at 11.2 and 11.7 Hz in a similar plant by accelerometer measurements made during the preoperational vibration measurement program. The 11.7 Hz mode indicated an rms response on the order of 2 mils (2).
The 11.1 Hz mode which was inferred to have antinodes near the ex-core detector azimuthal locations is also expected to have a similar amplitude, indicating that thermal shield motions on the order of 2 mils rms can be detected. This amplitude is a fraction of the calculated average rms shell mode amplitudes of the SONGS 1 thermal shield shell modes at the approximate locations of the detectors for both expected present conditions and postulated further degradation. The results from French plants (3) of Westinghouse design include detection of thermal shield n=2 and n=3 shell modes.
On the basis that thermal shield beam modes produce changes in relative annular water gaps similar to those of thermal shield shell modes, the thermal shield beam mode responses calculated with the flexure broken are also expected to result in neutron noise signal levels higher than those previously detected as discussed above. A discussion of the likelihood of detection of changes due to the failure of the remaining unbroken flexure is included in Section 10 of WCAP 12148.
-2
- 2) Past and Present Application of Neutron Noise Monitoring The neutron noise technique was recognized more than a decade ago as a technique for monitoring reactor internals vibrations. Correlation between core barrel beam motion and neutron noise sensed by a strain gage was shown by Gopal and Ciaramitaro (4), and a correlation between accelerometers mounted on reactor internals and neutron noise was shown by Thompson, et al (5).
Inadequate core barrel clamping force led to motions detectable by neutron noise in the Palisades plant (6).
An ASME/ANSI standard (7) has been issued to monitor core barrel beam mode vibrations to detect loss of core barrel clamping force using the neutron noise technique. The SONGS 1 programs will follow the basic phases of this program, baseline surveillance and diagnostic.
Extensive neutron noise monitoring has been performed on French plants (3) including 24 reactors with circular thermal shields and numerous plants without circular thermal shields. In addition, the neutron noise method has been used extensively on German reactors.
Examples of past uses of neutron noise are discussed in the following.
Westinghouse has supported the use of neutron noise to monitor for wedging of a loose part in the bottom of the reactor vessel. In this case, the natural frequency of the core barrel beam mode was monitored for changes. A change in natural frequency was detected following reinstallation of the core barrel after an in-service inspection. Analysis and additional data were acquired to diagnose the source of the change. Based on these results and the subsequent return of the natural frequency to prior values, it was concluded that the cause of the change was contact of the core barrel with a lower radial support due to the reinstalled position of the core barrel.
Westinghouse has also used the neutron noise technique to investigate the effects of a flow anomaly on internals vibration. This study included comparison of the natural frequencies of the internals of the plant studied to expected natural frequencies, and comparison of the responses during the occurrences of the anomaly to responses without the anomaly to show that the anomaly did not result in significant changes in core barrel vibration amplitude.
The above indicate that neutron noise is a recognized technique that has been useful for the study of reactor internals vibration.
Neutron noise monitoring for thermal shield support degradation is also being performed in a Westinghouse plant. Following the repair of the thermal shield at the Haddam Neck Plant, an internals vibration monitoring program was implemented by NUSCO (8).
Ex-core neutron noise data were acquired during the startup following the repair and during subsequent operation. Results from these data and further data to be obtained during the remainder of the cycle are used to monitor the thermal sheild for changes in motion and to establish a baseline for long term monitoring.
-3 References
- 1) D. N. Fry, 3. March-Lauba and F. J. Sweeney, NUREG/CR-3303 ORNL/TM-8774,
- 2) N. R. Singleton, et al, "Four Loop Internals Assurance and Test Program," WCAP 9352, June, 1978.
- 3) C. Puyal, et al, "Use of Low Frequency Fluctuations for the Surveillance of Structures, Sensors and Thermohydraulic Phenomena on the 900 MW and 1300 MW Reactors,"
presented at SMORN V, October 1987, Munich, Germany.
- 4) R. Gopal and W. Ciaramitaro, "Experiences with Diagnostic Instrumentation in Nuclear Power Plants," Progress in Nuclear Energy, Vol. 1, pp 759-779.
- 5) 3. P. Thompson, et al, "Experimental Value of Ex-core Detector Neutron Noise to Core Barrel Amplitude Scale Factor," Transactions of the American Nuclear Society, 32 (1979), pp. 797-798.
- 6) 3. A. Thie, "Core Motion Monitoring," Nuclear Technology, Vol. 45, Mid August 1979, pp. 5-45.
- 7) ASME/ANSI OM-1987,Part 5, "In-service Monitoring of Core Support Barrel Axial Preload in Pressurized Water Reactors."
- 8) Haddam Neck Safety Evaluation. Evaluation of Thermal Shield Support Failure and Repair, SECL-87-568.
03110
RESPONSE MODES 43 41D 42 44 BEAM OR PEULUM MODE S= IMDE FIGURE 1 BEAM AND SHELL MODE SHAPES
Steady Noz-.._ ed F.lu:x Noise F IGURE 2
raiakwK' ah"M
- t. o
".6 W019E PWA 1.11 to 9.0
__0-MW I A
/ 1"I 15. to 10
..0 WIW CL 19-9 flz 0ooa "A
- 5.
to 14.
9 929 39 49 5
FREQUENCY (Hz)
Fig. 34.
Long-tern variation In the amplitude of resonances in the Sequoyah I ex-core neutron noimexpectrum.
(From D. N. Fry, J. March-Leuba and F. J. Sweeney, NUREG/CR-3303, ORNLITM-8774 F IGURE 3