IR 05000244/1994024
| ML17263A854 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 11/15/1994 |
| From: | Lohmeier A, Modes M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17263A853 | List: |
| References | |
| 50-244-94-24, NUDOCS 9411220175 | |
| Download: ML17263A854 (14) | |
Text
'
U. S.
NUCLEAR REGULATORY COHHISSION
REGION I
DOCKET/REPORT NO.
LICENSEE:
FACILITY:
DATES:
50-244/94-24 Rochester Gas and Electric Corporation Rochester, New York 14649 Robert E. Ginna Nuclear Power Plant October ll - 17, 1994 INSPECTORS:
o m ier, r.
eactor ngineer Haterials Section jytvpsion f or S
ty ate APPROVED BY:
ic ae
.
o es, ie Haterials Section Division of Reactor Safety ate Areas Ins ected:
(1) The program for design, fabrication, and installation of two replacement steam generators; (2) the technical issue of monitoring transient operation cycles for those components designed for a limited number of operating cycles over the 40 year licensed plant lifetime; and (3) actions taken by Rochester Gas and E'lectric Company (RGKE) for resolution of NRC unresolved items.
Jtesu Its: '(I} The replacement steam generator design is simi'lar in thermal performance to.the original steam generator and reflects attention to correcting the original steam generator problems.
The steam generator replacement process is clearly defined, and the installation and testing plan carefully prepared; (2) the retention of cyclic operating data is consistent with Tech Spec (UFSAR) Section 6. 10.2.g and evaluation of the cyclic operation component usage factors is appropriate.
Procedures to affect data retention and evaluation have not been identified in your procedural manuals; (3)
Unresolved issues, URI 91-201-07 and URI 93-020-01, remain open; URI 94-015-01 and URI 50-244/93-021-01 were closed.
94ii220i75 94iii6 PDR ADOCK 05000244
DETAILS 1.0'COPE OF INSPECTION (Inspection Procedure 92903)
The scope of this inspection was to review actions taken by Rochester Gas and Electric Company (RGSE) for resolution of unresolved items.
Special attention was given to your program for the design, fabrication, and installation of replacement steam generators.
The technical issue of monitoring transient operation cycles, for those components designed for limited numbers of oper ating cycles over the 40 year licensed plant lifetime, was also reviewed.
2.0 STEAN GENERATOR REPLACEMENT 2.1 Background As a consequence of the degrading operational reliability of Ginna steam generators, t)at required plugging of tubes with reduced efficiency of the steam generator, Ginna engineering has, for several years, been giving consideration to replacing the origina')
steam generators (OSGs) with replacement steam generators (RSGs).
The utility decided to replace the original steam generators with those of a similar design, but included many design changes which eliminate some of the factors causing deterioration of the OSGs.
The design chosen was that submitted by Babcock and Wilcox Canada, a manufacturer that has produced similar steam generators for other nuclear power generation plants in the United States of America.
2.2 Replacement Steam Generator Design The RSG design, although different in many structural details, is very similar in performance to the OSG.
The tubes have been changed to 4765 3/4 inch diameter, 0.0431 wall thickness, Inconel 690 in a triangular array.
The primary side flow has changed from 34.6 to 31.1 million lb mass/hr with a pressure drop change from 33.5 to 31. 1 psi with OX plugging of tubes.
The primary side volume has changed from 133 cu ft to 132.5 cu ft for both inlet and outlet plenums, and from 654.5 cu ft to 710 cu ft in the tubes.
Secondary side volume has changed from 4580 to 4513 cu ft.
Secondary side water mass at 100X power has changed from 84,500 lb mass to 86,200 lb mass, with the secondary mass flow unchanged at 3.3 million lb mass per hour.
Nore significant changes in RSG design, noted by the inspector, include a
steam line orifice change from 4.37 sq ft to 1.4 sq ft and the initial steam pressure rising from 800 psia to 875 psia.
The heat transfer area in the absence of plugged tubes changed from 44,430 sq ft to 54,000 sq ft.
With 20X plugged tubes, the available heat transfer area oF the RSG design is 43,200 square feet as compared to 37,765 sq ft with 15X tubes plugged of the OSG.
In design of the RSG, Ginna provided for changes in design details which addressed steam generator problems that have given rise to considerable numbers of tubes to be plugged.
These problems included tube and tubesheet corrosion, high cycle fatigue, water hammer, moisture carry-over, pressure boundary weld failures, post-weld stress corrosion cracking of tube U-bends, secondary side loose parts, secondary side access, and primary nozzle weldin The RSG design provides for resolution of these problems through the following features:
The tubesheet design provides for a closed crevice through full depth hydraulic expansion of the tubes into the tubesheet.
Inconel 690 is used as tube material, providing greater corrosion resistance than Inconel 600.
The tubes are thermally treated.
Tubes are flush-welded to the primary face of the tubesheet after the lower shell and primary head are welded and have received post-weld heat treatment.
This-process precludes tube sensitization.
Sludge accumulation on the tube sheet is reduced by means of a higher recirculation ratio than that of the OSG.
Inspection and maintenance ports are provided 'for accessible sludge lancing to remove remaining sludge deposits.
A lattice grid of tube support plates provides an improved tube supporting arrangement with an optimal Inconel 690 tube to Inconel 690 support material combination to minimize fretting.
High cycle fatigue failures in the U-bend region are prevented by the fan bar support system in that region.
Mater hammer is prevented through use of a goose neck pipe at the feedwater ring inlet and the feedwater ring is constructed with welded Inconel 690 J-tubes which have maximum erosion resistance.
2.3 Hoisture Carryover is reduced with a major change in steam separator design to give a guaranteed maximum 0.10X carryover.
Potential pressure boundary weld failures are prevented through the use of forged and plate components, the absence of corner welds, strict adherence to pre-and post-heat treatment of welds.
The possibility of secondary side loose parts is reduced through the elimination of fasteners and completely welding the structure.
Primary side access is provided with 18 inch manways, and secondary side access is provided through 6-8 inch handholes, 14-2 inch inspection ports, and l-l8 inch manway.
Optimal primary nozzle welding is achieved with 316 LN safe ends, using narrow gap welding.
Fabrication, Installation, and Testing The inspector reviewed the fabrication, installation, and testing procedures used for the RSGs.
The replacement process was clearly defined, and the installation and testing plan was carefully pr epare The RSG fabrication procedures, drawing outlines and details, and materials lists were clearly presented.
The manufacturing and test plan produced by Babcock and Milcox Canada, reviewed by the inspector, is a step-by-step description of the manufacturing sequence, with verification and gA hold points.
This docum'ent provides a clear description of the process sequences from which audits can be made.
The licensee discussed the OSG removal and RSG replacement sequences with the inspector.
The structural adequacy of containment penetrations was reviewed, and the route of removal and replacement of the steam generators appeared to be planned and in accordance with practices in keeping with nuclear power plant safety procedures.
2.4 Conclusions The RSG design is similar in thermal performance to that of the OSG.
The RSG design reflects attention to correction of the details with which the OSG had problems.
The replacement process is clearly defined, and the installation and testing plan carefully prepared.
3.0 TRANSIENT OPERATING CYCLE NONITORING 3.1 Background and Scope p
(W The primary system components are designed to meet the requirements of Section III of the American Society of Mechanical Engineers (ASHE) Boiler and Pressure Vessel Code for nuclear vessels.
The Code requires a design-by-analysis approach to evaluate whether the components can sustain the prescribed steady state pressure and thermal loadings along with the cyclic application of these loadin s.
g The utility (owner of the components)
specifies the types and.numbers of loadings which are anticipated during the plant lifetime.
Components are designed in accordance with these specifications.
In the case of cyclic loading the specification will state the numbers and types of transient operation anticipated during the plant life.
These transients are described in the Updated Final Safety Analysis Report (UFSAR) for,the nuclear power plant.
Because primary system components are designed to sustain a limited number of transients, the plant technical specification (TS) stipulates the requirement to record cyclic operation of the plant.
This data identifies critical areas of the components subjected to the operating transients, and can be used to determine whether the design fatigue life of the component is approaching exhaustion.
The criteria for exhaustion of fatigue life is reflected in a cumulative usage factor (CUF), which is a summation of the ratio of expected numbers of cycles, at the applied strain range, to the cycles at that strain rang An appropriate factor of safety, in terms of strain level or cycles, is utilized in the same manner as a factor of safety for str ess level in relation to fracture stress.
The Code limits the calculated CUF to less than 1.0.
The intent of the inspection was to ascertain that the licensee is retaining the records of cyclic operation for those components designed for limited numbers of cycles specified in Technical Specification Section 6. 10.2.g 3.2 Operating Cycle Nonitoring In order to demonstrate that the cyclic operation of the plant primary system remains within the range of transient operating cycles assumed in the original design analyses of the plant, the licensee provided, for inspector review, a
report dated August, 1989, entitled,
"R.E. Ginna Nuclear Power Plant RCS Transient Validation Report."
The report provides a list of equipment specification transients together with the corresponding numbers of cycles expended over 19 years operation (as of 1989) for the reactor vessel, pressurizer, and steam generators.
Also shown for each transient is a
conservative estimate of the numbers of cycles estimated for 60 years operation.
For all transients, the numbers of operating cycles after 60 years operation remained less than the assumed number of cycles for which the components were designed.
Because 60 years is 50X higher than the 40 year licensed cyclic life, the evaluation is conservative.
Because there has been no significant change in the rate of operating cycle application, the inspector finds that the numbers of cycles expended to date (1994) remain less than the assumed design cycles.
The inspector also found from this study that the licensee has demonstrated compliance with Technical Specification Section 6. 10.2.g by having the ability to obtain transient data from its operating records.
3.3 Fatigue Life Usage Because the equipment specifications define the types and numbers of transients to be considered in fatigue evaluation of the primary reactor coolant system components, the analyses of fatigue life usage over 40 years operation under the action of these transients is computed by the equipment vendors.
If the analyses show the equipment components CUF to be less than the limit of 1.0, then operation under the action of the numbers and magnitudes of transient cycles within the levels for which the equipment has been designed will not result in the component reaching the end of its fatigue life.
The inspector noted that the transient operation of the reactor vessel, steam generator, and pressurizer is within th'e transient range assumed in the design of the equipment.
It can therefore be concluded that CUFs of the RCS equipment components are below Ginna has reported the lifetime (40 year)
CUFs computed by the equipment vendors to be less than 1.0.
The inspector noted the component usage factors of the reactor vessel stud bolts (.511), primary nozzles (. 155),
and safety injection nozzle (.47).
The highest pressurizer component CUFs are at the surge nozzle (.264)
and spray nozzle (.096).
Each of these CUFs are substantially below the limit of l.
3.4 Procedures Although there was evidence that Ginna had retrieved operating data from its files that provided the input for evaluation of the fatigue life usage of the RCS equipment components, the inspector found that there was no written procedure to obtain the data, evaluate the data, and provide for management review.
The licensee stated that they intend to develop a procedure for this purpose to be included in the site procedures.
3.5 Conclusions
'
The retention of cyclic operating data is consistent with Section 6. 10.2.g.
Continuing evaluation of the cyclic operation component usage factors is appropriate for maintenance of nuclear components.
Procedures to effect data retention and evaluation have not been identified in the procedural manuals.
4.0 REVIEW OF UNRESOLVED ITENS 4. 1 URI 93-020-01 Containment 3.6.3 Technical Specification Requirements (Open)
The inspector reviewed an issue related to the restoration of an inoperable boundary to operable status within allowed time limits in accordance with Technical Specification (TS) Section 3.6.3.
RG&E indicated to the inspector that the Safety Evaluation by the Office of Nuclear Reactor Regulation related to Amendment No.
54 "Requirements for Containment Isolation Boundaries for Facility Operating License No. DPR-18" did not recognize the fact that RG&E rescinded the allowable outage time (AOT) proposal prior to the safety evaluation.
Since the safety evaluation was based on consideration of the AOT proposal, it was not responsive to the rescinded proposal.
Pending resolution of the AOT issue between NRR and RG&E, the inspector considers URI 50-244/93-020-01 to remain open.
4.2 URI 50-244/93-021-01 Emergency Bus Load Shedding (Closed)
During inspection 93-021, a violation of NRC requirements was identified related to the failure to perform testing to verify load shedding from the emergency buses during the 18 month surveillance tes't period.
RG&E, in a reply to the Notice of Violation, accepted and identified the basis for the violation and recognized the need to verify load shedding capabilities of safeguard loads with undervoltage in a safety injection mode.
RG&E identified
'a corrective actions taken through testing conducted on October 10-11, 1993, in combination with other surveillance tests during the 1993 outage.
The tests met the requirements of Technical Specification (TS) 4.6. l.e.3.(a)
and, verified that the safeguards would have performed as required.
RG&E identified corrective action to be taken to avoid further violations of TS Section 4.
These included review of Section 4 to ensure that there are, implementing procedures for every surveillance required, therein, a review of the implementing procedures to ensure compliance with the TS.
and upgrading procedures that verify load shedding capability will include safeguard loads for conditions of undervoltage and safety injection.
The inspector verified that full compliance with TS 4.6. l.e.3.(a)
was achieved on October 12, 1993, at the completion of surveillance testing.
On this basis, URI 93-021-01 is closed.
4.3 URI 91-201-07 Review/Approve Altran SWS Check Valve Passive Failure Report
{Open)
In response to an unresolved item 50-244/91-201-07, a design analysis,
"SWS Check Valve Passive Failure,"
NSL-OOOO-DA043, was completed and approved on September 14, 1992.
This analysis confirmed preliminary results that flow delivered to the service water. system (SWS),
when postulating that an open SWS pump discharge check valve would not close, would exceed the flow delivered from one pump operation.
Therefore, the postulated event is bounded by the failure of one emergency diesel generator (EDG), the limiting failure assumed for the accident analysis.
Therefore, this postulated event is bounded by the single failure of one EDG, the limiting failure assumed for the accident analysis.
RG&E contracted for re-review of Altran Report 90121.6,
"Single Failure of the R.
E. Ginna Service Water System," Revision 1,
November 18, 1991.
This re-review was completed in December 1992.
RG&E's review of the re-review of Altran was completed and will be included in a final report contracted to Bell Engineering.
The final report is presently under review by RG&E.
The due date for completion of the RG&E final review was October 1,
1994.
During the inspection, the inspector observed the final report, with comments.
Pending completion of the review of the Bell Engineering final report by RG&E, the unresolved item 50-244/91-201-07 remains open.
4.4 URI 94-015-01 Hain Feedwater Regulating Valve (NFRV) Trim Change (Closed)
Inspection Report 94-15, in review of the main feedwater regulating valve
{HFRV) trim change
{Section 2.2), considered the Technical Staff Request/Technical Evaluation (TSR/TE) to be an unresolved item, pending resolution of further actions by RG&E.-
These actions include satisfactory
completion of the main feedwater regulating valve stress analysis, long term corrective action for the MFRV positioner failures, and review of the TSR/TE process for assurance of adequate management oversight.
The inspector reviewed the RG&E responses to components of the unresolved item.
RGSE provided for the inspector's review "Piping Analysis for the Feedwater System Line FW-301," DA-ME-94-092, Revision 0, that provided for stress justification of the changed system.
Also provided for inspector review was the document M-94-007, "Root Cause Analysis Summary Report," which identified the cause of the valve malfunction and provided for the necessary corrective action.
In an Inter-Office Correspondence, REP-94-09-30A, RGEE described the management oversight and corrective action initiated, that demonstrated tliat the TSR/TE process was implemented correctly.
On the basis of the inspector's review of the foregoing responses of RGSE, URI 94-015-01 is considered closed.
5.0 CONCLUSIONS
~
The RSG design is similar in thermal performance to that of the OSG.
The RSAG design reflects attention to correction of the details with which the OSG had problems.
The replacement process was clearly defined, and the installation and testing plan carefully prepared.
The retention of cyclic operating data is consistent with Technical Specification Section 6. 10.2.g.
Continuing evaluation of the cyclic operation component usage factors is appropriate for maintenance of nuclear components.
Procedures to affect data retention and evaluation have not been identified in the procedural manuals.
URI 91-201-07, Review/Approve Altran SWS Check Valve Passive Failure Report, and URI 93-020-01, Containment 3.6.3 Technical Specification Requirements remain open.
URI 94-015-01, Main Feedwater Regulating Valve (MFRV) Trim Change, and URI 50-244/93-021-01, Emergency Bus Load Shedding are closed.
6.0 MANAGEMENT MEETINGS The scope and purpose of the inspection were discussed at an entrance meeting conducted on October ll, 1994.
During the course of the inspection, the inspector's findings were discussed with the licensee representatives listed in Attachment A.
An exit meeting was held on October 17, 1994, at which time the findings were discussed with members of the engineering and compliance staff.
The licensee acknowledged the findings and conclusions, with no exceptions taken.
Attachment A lists the members of the engineering and licensing staff in attendance at the entrance and exit meetings and those contacted during the course of the inspectio '
ATTACHMENT A Persons Contacted Rochester Gas and Electric Com an R.
- Q C.
- G
O
W.
- p
- Q Eliasz Flynn Forkell Hermes Lil1ey Harkowski Rochino Tono Voci Wilkens Wrobel Lead Engineer, Nuclear Safety and Licensing Nuclear Engineer, S/G Replacement Project Manager, Electrical Engineering Lead Engineer, Safety and Licensing Manager, equality Assur ance Lead Mechanical Engineer Lead Mechanical Engineer Engineer, S/G Project Hanager, Mechanical Engineering Department Manager, Nuclear Engineering Services Manager, Nuclear Safety and Licensing
- 'ndicates those in attendance at the exit meetin I
~