ML20127A889
| ML20127A889 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 06/17/1985 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | Harold Denton, Thompson H Office of Nuclear Reactor Regulation |
| References | |
| CON-NRC-85-2 GL-85-02, GL-85-2, VPNPD-85-3, NUDOCS 8506210285 | |
| Download: ML20127A889 (8) | |
Text
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l%sconson Sectnc rouscouver auna 1,,
19e5 231 W. MICHICAN. P 0. BOX 2046, MitWAUMI, WI $3?01 VPNPD-85-3 NRC-05-2 Mr. II. R. Denton, Director 1
Offico of Nuclear Roactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C. 20555 Attention:
Mr. Ilugh L. Thompnon, Director Divinion of Licon31ng Gontlemons DOCKET NOS. 50-266 AND 50-301 l
RESPONSE TO GENERIC LETTER 85-02 liTEAM GENERATOR TUBE INTEGIIIT I i
POINT __IAEACll NUCLEAR PLANT, UNITS l AND 2
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Conoric Lottor 85-02 untitled " Staff Rocommendod Actionn Stemming From NRC Integrated Program for the renolution of Unrunolved Safety Innuun Hogarding Steam Gonorator Tubo Integrity" roquanted licanucon to furnish the NRC a doucription of their ovorall programo for asuuring utoam gonorator tubu integrity and for steam gonorator tubo rupturo mitigation.
Specifically, wo woro requentud to provido a doncription in sufficient detail to runpond to each of thu NRC ntaff-recommanded actionn no pronentud in Enclouuro 1 to the letter.
Liconnoon woro alno anked to discuou the practicon thoy omploy to onouro adoquato inopoction campion are takon in the ovent that Category C-2 ronultu are obtainud during tho initial technical upocification-roquirod innorvico inspection nampling of utoam gonorator tubou.
Attached to thiu lottor in Wiuconnin Electric Powor Company's rouponno to tho Staff'n rocommandationu for the Point Donch Nucinar Plant, Unita 1 and 2.
Also attached lu a dincunnion of the practicon wo havo employod in the pant for inupoeting additional utoam gonorator tubon when C-2 rouulta woro realized after an initial tubo namplo.
In thin rouponno, and in portionn of other ronponnon, wo havo discunnod pant practicon wo have unod and expoet to uno again in the futuro at the Point Donch Nuclear Plant.
Wo wiuh to clarify that our discunnion of thuno practicou in thin lottor doon not conotituto a regulatory commitmont to f
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Mr. H. R. Denton Juno 17, 1985 i
f nocessarily continuo such practicos in the future, particularly when more offective techniques and methods are developed for onsuring steam generator tubo integrity.
Of courso, in those casus whero our responses discuss elomonts of our program which are a part of the Point Beach Technical Specifications, any changos would l
first be submitted for NRC approval as required by 10 CFR 50.59.
Please contact un if you have any questions concerning this information.
I Vory truly yours, 1
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Vico Prosiden -Nuclear Power i
C. W. Fay Encionuron Copy to NRC Resident Inupactor I
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O RESPONSES TO INFORMATION REQUESTED IN ENCL.05URE T0 NRrlIENERTC LETTER 85-02 la. Prevention and Detection of Loose Parts (Inspection,},
Response
The Point Beach Nuclear Plant Unit I and 2 steam generators have undergone visual inspections of the steam generator secondary sides in the recent past. A 100% annular search of the Unit 2 steam generators secon-dary side was perfonned during the Spring 1982 refueling outage.
The results of this inspection were reported in our Annual Results and Data Report for 1982. The inspection was made using a fiber optic device. One piece of wire was found and recovered from the "A" steam generator and nothing was found in the "B" steam generator.
No signs of damage to the steam generator were apparent.
The Unit I replacement steam generators were visual inspected as part of the preservice inspection of the units in 1984.
The inspection was conducted using a fiber optic device and no loose parts were observed.
Since all steam generators have been visually inspected on the secon-dary side, we have complied with the first of the three visual inspection criteria identified in the staff recommended action. We are in basic agreement that conditions identified in criteria (2) and (3) of the staff's recommendations, specifically secondary side modifications or repairs and eddy current indications in a peripher11 tube free span portion which cennot be attributed to mechanism other than a loose part or foreign object, should result in additional secondary side visual inspections.
It is the policy of Wisconsin Electric to maintain the Point Beach Nuclear Plant steam generators in a wet layup condition whenever possible and thereby minimize the potential for corrosion of the tube bundle when exposed to air. The containment atmosphere at Point Beach has not been subjected to industrial atmosphere pollutants, such as sulfurt therefore, we would not anticipate exposure of the steam generator tube bundles to corrosive chemical species when they are opened to air. We would note that EPRI NP-2656, " Evaluation of Secondary System Layup " indicated that the potential for carbon steel corrosion for periods of exposure to air of less than 10 days is insignificant.
Since the Inconel tube bundles at Point Beach are much more corrosion resistant than carbon steel, we would expect no corrosion problems for the brief periods when wet layup cannot be main-tained.
O 2
Ib. Prevention and Detection of Loose Parts (Quality Assurance)
Response
1 Procedure controls at the Point Beach Nuclear Plant include a Routine Maintenance Procedure (RMP) for removal of steam generator manways or hand-holes. These procedures are invoked whenever the steam generator primary l
or secondary sides are opened for access. Detailed accountability for all tools, equipment, and materials is maintained by means of material log-in and log-out requirements. Other administrative controls include lanyards i
for small equipment and personal items and design features such as lock wires on equipment to prevent loss in the steam generator.
The primary side of the steam generators is always subject to a detailed closeout visual inspection prior to replacement of the manways. Accountability for components or parts removed from the internals during major steam generator repairs is not specifically addressed by the RMP but would be a part of the specific repair or modification procedures written to cover such an activity.
The Point Beach Nuclear Plant steam generators are also fitted with loose part monitoring devices. These monitors permit prompt detection of metallic loose parts in both the primary or secondary sides of the steam generator in the vicinity of the tubesheet by means of acoustical monitors which readout and alarm in the plant control room.
2a.
Inservice inspection Program (Full Length Tube inspection)
Response
It has been the recent policy of Wisconsin Electric to conduct full length inspection of the initial sample of steam generator tubes from the hot leg around the U-bend to the end of the cold log. We have been con-ducting such inspections for the last two outages on both unitst however, we have several concerns with making this a requirement.
Because of the reduction in eddy current probe size necessitated by such full length tube inspections, there is some degradation in the sensitivity of the eddy current detection capabilities.
If cold leg inspections from the cold leg side are required, significant increases in personnel exposure will result.
The requirement to inspect all initial sample tubes full length is there-fore questionable for enhanced inspection results.
Supplemental sampling procedures at Point Beach have, generally, been limited to a partial length inspection over those portions of the tube length where degradation was found during initial sampling. This is con-sistent with the recommended staff policy.
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'. o 2b.
Inservice Inspection Program (Inspection Interval)
Response
The Point Beach Nuclear Plant Technical Specifications presently require steam generator inservice inspections to be not more than 24 months apart. The inservice inspection may be limited to one steam generator on an alternating sequence basis.
If two consecutive inservice inspections, covering a time span of 12 months or more, yield inspection results that fall in the C-1 category, the inspection interval may be extended to 40 months. Thus, in the extreme, the interval between inspections of an individual steam generator at Point Beach could be as long as 80 months.
Although this is greater than the 72 months in the staff's recommendation, we believe this inspection interval is adequate.
This belief is based on our experience that steam generator degradation problems in a particular unit, separate from foreign object damage, effect each steam generators in that unit at essentially the same rate.
Therefore, inservice inspection of the conditions in one steam generator provides adequate' prediction of the condition in the other steam generators.
3a.
Secondary Water Chemistry Program
Response
The Point Beach Nuclear Plant has been operated since 1979 with a license condition which required that we implement a secondary water chemistry monitoring program to inhibit steam generator tube degradation.
This program includes:
1.
Identification of a sampling schedule for the critical parameters and control points for these parameters.
2.
Identification of the procedures used to quantify parameters that are critical to control points.
3.
Identification of process sampling points.
4.
Procedures for the recording and management of data.
5.
Procedures defining corrective action for all control point chemistry conditions; and 6.
A procedure for identifying the authority responsible for the inter-pretation of data, and the sequence and timing of administrative events required to initiate corrective action.
The plant procedures which implement this program include three action levels for progressively more stringent corrective action.
The parameter values for normal operation and initiation of these action levels are based on the Steam Generators Owners Group (SG0G) guidelines in EPRI-NP-2704-SR Rev.1, June,1984.
Corrective actions at Action Levels 2 and 3 include power reductions and unit shutdown when continued full power operation may result in excessive steam generator corrosion.
The procedure 1
also identifies the plant personnel responsible for monitoring and report-ing the secondary chemistry condition, and provides for prompt notification of plant management, including the plant manager, when various action levels are reached.
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9 3b.
Condenser Inservice Inspection Program
Response
Wisconsin Electric personnel have been very sensitive to the effects of condenser inleakage on steam generator water chemistry. We have an aggressive program to monitor and control both condenser air and water inleakage. - Condenser air inleakage is generally maintained below five cfm.
Should air inleakage exceed eight cfm, we initiate a comprehensive inspec-tion program to identify and correct the leakage path.
During refueling outages, the condensers and hot wells are flooded to further aid in identi-fying potential air and water leakage pathways.
During plant operations, sensitive hotwell cation conductivity and sodium monitoring instruments are utilized to identify condenser tube leak-age. Steam generator blowdown cation conductivity is also monitored as an early indication of condenser water inleakage.
Experience has shown these monitors to be sensitive to hotwell inleakage of as little as 0.02 gpm.
When significant condenser water inleakage is identified, prompt action is taken to reduce power, remove the effected tube bundle from service, and identify and plug the leaking condenser tube or tubes. A variety of detec-tion methods are used (including helium or freon detection, foam, and acoustical sensing) to identify leaking tubes. Repair _ techniques have generally consisted of plugging the leaking. tube using rubber plugs.
Inservice inspection of the condenser tubes has included the use of eddy current inspection methods. We do not presently, nor do we plan to, incorporate such inservice inspections into a safety related plant procedure.
The impact of condenser leakage is an economic and not a safety concern.
Unless the steam generator chemistry conditions cannot be maintained, inservice inspection of condenser tubes results in no safety benefit.
We are concerned with the staff's recommendation that condenser inser-vice inspection programs be initiated whenever condenser leakage is such that a power reduction corrective action is required more than once per three month period. 'This recomnendation is difficult to interpret as to the level of power reduction and degree of condenser inspection. Addi-tional guidance as to the staff's intentions regarding this recommendation is necessary.
For your information, because of our concerns with long term condenser tube integrity and our sensitivity to the importance of condensate purity to steam generator water chemistry, we are now planning and engineering the replacement of the condenser tubes at Point Beach.
The existing Admirality tubes will be replaced with Type 304 Stainless Steel Tubes. We anticipate retubing of the Unit 2 during the Fall 1985 refueling outage and Unit 1 during the Spring 1986 outage.
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Primary ~to Secondary Leakage Limit
Response
The Technical Specification limits-for primary to secondary leakage rates for the Point Beach Nuclear Plant are more restrictive than the Standard Technical Specification.
Leakage in excess of 500 gpd in either steam generator requires the plant to be shutdown and placed in cold shutdown.
5.
Coolant Iodine Activity Limits
Response
The Technical Specification limits and surveillance for reactor coolant iodine activity for the Point Beach Nuclear Plant are essentially the same as the Standard Technical Specification and were issued by the NRC effective on April 4,1983. We still disagree, however, with the reporting requirements for iodine spiking required by these specifications. The iodine spiking phenomenon is well understood and the need for such report-ing, which was initiated for data collection in order to assess whether a safety concern existed, is no longer necessary.
Revised Emergency Operating Procedures based on the recommendations of the Westinghouse Owners Group guidelines are presently undergoing final review and approval for implementation on July 1, 1985. The reactor cool-ant pump trip criteria for the steam generator tube rupture E0P are based on subcooling margin. These criteria are such that forced reactor coolant flow should be maintained, assuming offsite power is retained, for steam generator tube rupture events up to and including the design basis break of a single steam generator tube. We therefore believe that we will satisfy this staff recommendation upon implementation of the E0Ps.
6.
Safety Injection Signal Reset
Response
The control logic associated with the safety injection pump suction switchover from the boric acid storage tanks to the refueling water storage tank is dependant only upon low level in the boric acid storage tank. This switchover will occur automatically upon low level regardless of the condi-tion or state of the safety injection signal.
.' 7.
Information Concerning Category C-2 Tube Inspections
Response
It has been the policy of Wisconsin Electric to select a Category C-2 steam generator tube inspection sample size greater than the sample size required by the Technical Specification. The actual number of tubes selected and the extent of the inspection in each tube has been based primarily on consideration of the degradation mechanism believed to be involved.
In the past, inspection of the Unit 1 steam generators for IGA indications in the tube sheet area often resulted in Category C-2 results leading to essen-tially a full inspection of all the tubes in both steam generators. There have also been instances of tube degradations associated with the high heat flux or " kidney" zone of the steam generator tubes in which a C-2 sample included all tubes in the vicinity of the " kidney" zone.
Generally minimum reinspection intervals have been determined from the requirements imposed by the Technical Specification based on the inspection results. However, in the past we can also document situations in which additional steam generator inspection outages were scheduled based on specified days of full power operation. These operating intervals were negotiated with the NRC for the specific situation and based on an under-standing of the rate of degradation such that tube rupture due to the degradation would be unlikely to occur during the operating period.
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