ML20211F124
| ML20211F124 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 02/17/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20211F120 | List: |
| References | |
| NUDOCS 8702240504 | |
| Download: ML20211F124 (8) | |
Text
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o, UNITED STATES NUCLEAR REGULATORY COMMISSION y
g g,
E WASHINGTON. D. C. 20555 y.....J SAFETY EVALUATIOP BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.
3 TO FACILITY OPERATING LICENSE NO. NPF-39 PHILADELPHIA ELECTRIC COMPANY LIHERICK GENERATING STATION. UNIT 1 DOCKET NO. 50-352
1.0 INTRODUCTION
By letter dated November 17, 1986 as amended on December 22, 1986 and as supplemented on January 2 and 29, 1987, Philadelphia Electric Company (the licensee) requested an amendment to Facility Operating License No.
NPF-39 for the Limerick Generating Station (LGS), Unit 1.
The proposed amendment would change the Technical Specifications (TS) to permit operation of the unit with partial feedwater heating (PFH) and increased core flow (ICF) limits and would delete License Condition 2.C(13) which presently prohibits the use of PFH. Specifically, TS 3/4.2.3 " Minimum Critical Power Ratio (MCPR)," TS Table 3.3.6-2 " Control Rod Block Instrumentation Setpoints,"
and TS 4.4.1.1.2 " Reactor Coolant System Surveillance Requirements" would be revised to pemit operation of Unit I with a reduction of incoming feed-water temperature (partial feedwater heating, PFH) of up to 60"F and an increase in reactor core flow rate up to 105% of rated flow. License Con-dition 2.C(13), " Operation With Partial Feedwater Heating at End-of-Cycle" would be satisfied since the basis for the condition, namely that the applicable safety analyses to permit operation with PFH had not been performed, has been satisfied by the submittal of such analyses by the licensee. Near the end of a fuel cycle the depletion of fissionable material from prior power production results in a condition wherein the 100% of rated powercondition(3293MWt)cannolongerbemaintained. From this point, as the licensee states, a "coastdown mode" of operation with all power control rods " full out" may be followed to extend the fuel cycle. This would result in a gradually decreasing rate of power production until an optimum point is reached at which the licensee would remove the unit from service for refueling. The changes approved by this amendment, partial feedwater heating and increased core flow, take advantage of the boiling water reactor operating characteristics to allow an extension of the fuel cycle. The partial feedwater heating provisions also provide increased operational capability by providing for operations with some of the feed-water heaters removed from service.
As support for the proposed modifications, the licensee provided a General Electric Company report, NEDC-31323 " Increased Core Flow and Partial Feedwater Heating Analysis for Limerick Generating Station Unit 1 Cycle 1" dated October 1986. In response to staff requests for infonnation on this report and on the licensee's application, the licensee submitted additional information by letters dated December 22, 1986 January 2, 1987 i
and January 29, 1987.
8702240504 870217 PDR ADOCK 05000352 P
2.0 EVALUATION License condition 2.C(13) for LGS contains, in part, a requirement that analyses of operation in the partial feedwater heating (PFH) mode must be provided to the staff for review and approval prior to operation in that mode. The analyses for PFH and increased core flow (ICF) were provided by the licensee in Reference 1.
These analyses provide the required minimum critical power ratio (MCPR) operating limits for the proposed operation of LGS up to a maximum feedwater temperature reduction of less than or equal to 60' F at rated power and up to a maximum core flow of 105%
of rated flow. Certain abnormal transients and accidents analyzed in the LGS FSAR have been examined for the effect of the proposed operational mode. The staff's evaluation of these considerations is discussed below.
2.1 Anticipated Operational Occurrences The most limiting anticipated operational occurrences are:
l a)
Generator Load Rejection with Bypass Failure (LRNBP) b)
FeedwaterControllerFailure(FWCF)
The evaluations were performed at 104.5% power (consistent with the origiral FSAR analysis input assumptions), 1057 core flow with rated i
feedwater temperature of 360* F at end of cycle (EOC1). Nuclear trans-i ient data consistent with the original FSAR analysis were-developed and the combination of the extended load line limit analysis (ELLLA) and PFH were incorporated in the analyses. Based on the limiting transients identified by previous analyses, the proposed MCPR operating ifmits are developed to include the cases of turbine bypass inoperable and end-of-cycle recirculation pump trip inoperable. The new required operating limit MCPRs shall be 1.24 and 1.18 (based on ODYN Option B results for the feedwater controller failure without bypass transient with reactor pump trip and without reactor pump trip, respectively) for maximum core flow 105% of rated and a maximum feedwater reduction of 60' F.
The new 1
calculated operating limit MCPR values are incorporated into the proposed I
i technical specifications.
I Lower initial operating pressure and stean flow rate (due to lower feed-I water temperature) provide more overpressure margin for the limiting MSIV closure flux scram event. Hence, it is concluded that the pressure barrier integrity is maintained under PFH conditions. The licensee has analyzed the overpressurization limiting transient (MSIV closure) for increased core flow (ICF) without PFH. The analysis of this bounding transient predicted a peak vessel pressure of 1273 psig which is below the ASME code limit of 1375 psig and the analysis results are therefore acceptable.
The fuel loading error accident, rod drop accident, and rod withdrawal error have been evaluated by the licensee for ICF and/or PFH operation.
The rod withdrawal error transient is limited by a rod block system. The addition of a "high flow clamped" trip setpoint limit of 106 percent and
- allowable value of 109 percent of rated flow for the rod block monitor upscale alann in TS Table 3.3.6-2 ensures that the rod blocks currently in the TS cennot be exceeded. This is the same requirement that has been in effect since initial plant operation. The reactor coolant system recirculation flow upscale trip setpoints and allowable values are increased and the values for the recirculation pump motor-generator (MG) set scoop tube mechanical and electrical stops are increased. These changes are necessary to accommodate the increased core flow operation and are acceptable.
The licensee has stated that the fuel loading error and rod drop accident are not adversely effected by the proposed changes. For the fuel loading error, the licensee has reported in Reference 3 a meximum increase in CPR of 0.04 from the value of 0.11 stated in the FSAR for this event at rated conditions. Thus the fuel loading error remains a non-limiting event. With regard to the rod drop accident, the LGS utilizes a banked position with-drawal sequence (BPWS) for control rod movement. Based on prior staff review of BPWS as presented by General Electric (Ref. 5. Section 5.2.5.1.3),
the staff agrees that this event is not adversely affected by the proposed changes.
2.2 Loss of Coolant Accident Analysis A loss of coolant accident (LOCA) with ICF and PFH was addressed by the licensee in Reference 2.
The LOCA analyses with ICF alone bound operation with ICF and PFH. Since the peak clad temperature for ICF increases by less than 10' F for the limiting break compared to the rated core flow condition, the calculated peak clad temperature (PCT) of approximately 2100* F remains below the 10 CFR 50.46 cladding temperature limit. No chanoes to the current maximum average planar linear heat generation rates (MAPLHGR) are required. In Reference 2, GE stated that PCT changes through-out the remainder of the large break spectrum will be of a similar magnitude (less than 10' F). At the request of the staff, the licensee provided additional information (References 4 and 8) on the effect of increased core flow (ICF) and reduced feedwater temperature on the LGS LOCA analysis.
Consideration was given to the break spectrum range of 60 to 100 percent of the design basis accident (DBA) for the separate effect of ICF for several classes of BWR plants with the resulting conclusion that increased core flow results in a peak clad temperature increase of less than 10 degrees F throughout the large break spectrum. The separate effect of reduced feed-water temperature is to reduce the calculated peak clad temperature. A discussion was presented for both reduced feedwater temperature and increased core flow conditions which bounds the conditions described in the proposed amendment. Based on the staff's review of the additional information pro-vided by the licensee in References 4 and 8, which discusses the LGS specific LOCA analyses, the staff agrees with the conclusion in NEDC-31323 that the effect of ICF will not alter the limiting break size. The calculated peak clad temperature remains below the 10 CFR 50.46 cladding temperature limit and is acceptable.
The impact of the proposed operatino mode on containment LOCA response was considered by the licensee. A conservative analysis resulted in a peak dry-well deck downward differential pressure 2.6 psi higher than the value of
. 26.0 psid in the LGS FSAR. However, this is still below the design limit of 30.0 psid reported in the FSAR. The licensee stated that the drywell and suppression chamber temperatures, external pressures and maximum allowable leakage rates are bounded by the results reported in the FSAR.
It was also stated that the chugging loads, condensation oscillations and pool swell loads were found to be bounded by the appropriate design loads.
We find this acceptable.
2.3 Thermal-Hydraulic Stability Reference 2 included a discussion of themal-hydraulic stability (THS) for the LGS. The current LGS technical specifications implement a generic set of operating recommendations (Ref. 6) to assure acceptable plant performance in the least stable portion of the power / flow map and to provide operator instructions for the detect-and-suppress mode of operation. The THS compliance for all licensed GE BWR core fuel is demonstrated on a generic basis by Reference 7 and has been approved by the staff (NRC Safety Eval-uation Report Approving Amendment B to NEDE-24011-P contained in Appendix US-C to Reference 5). The staff concludes that acceptable THS provisions have been made to cover the proposed modifications.
2.4 Flow Effects NEDC-31323 (Reference 2) presents the results of a safety and impact evaluation of the limiting normal operational transients, loss-of-coolant accidents, fuel loading error accidents, rod drop accidents, and rod with-drawal error events. In addition, the effect of increased pressure dif-ferences on the reactor internals components, fuel channels and fuel bundles was also analyzed to show that the design limits will not be exceeded. The effect of ICF on the flow-induced vibration response of the reactor internals was also evaluated to ensure that the response is within acceptance limits.
The themal-hydraulic stability was evaluated for ICF/PFH operation, and the increase in the feedwater nozzle and feedwater sparger usage factors due to the feedwater temperature reduction was determined. The impact of ICF/PFH operation on the containment LOCA response was also analyzed.
j This evaluation in section 2.4 of this safety evaluation addresses only l
those portions of Section 3.1, 3.2.1, 4 and 5 of Reference 2 which are pertinent to load impact of reactor internals flow-induced vibration and feedwater nozzle and feedwater sparger fatigue usage. Subsequent to its review of Reference 2, the staff requested clarification from the licensee regarding those three areas. In response to the staff's request, the licensee submitted a letter from J. W. Gallagher (PECo) to W. R. Butler (NRC)datedJanuary 29,1987(Reference 8).
2.4.1 Reactor internals load Impact All the reactor internals (e.g., core plate, shroud support, shroud, top guide, shroud head, steam dryer, control rod guide tube, control rod drive housing and jet pump) were evaluated under the consideration of additional loads imposed by the ICF and PFH operations. The conclusion,
. as stated in NEDC-31323, is that the stresses produced in those components are within the ASME Code,Section III, Subsection NG allowables, the criteria referenced in the FSAR. Based on the reported results, the staff finds this acceptable since the FSAR design limits were satisfied.
2.4.2 Flow-Induced Vibration To ensure that the flow-induced vibration response of the reactor internals for plant operation with ICF up to 105% rated flow is acceptable, the prototype (Browns Ferry 1) plant test data were used as the bases for this vibration assessment. The Browns Ferry 1 test results are described in NEDE-24057-P-A which was accepted by the staff in a letter from R. Tedesco (NRC) to G. Shemood (GE) dated October 28, 1960. Results from the Browns Ferry 1 test program include data up to 113% core flow.
Test measurements of all sensors in all instrumented reactor internal components were reviewed, evaluated and compared with the acceptance criteria. The absolute sum of the peak alternating stresses of all the vibration modes was obtained. Using this method the maximum vibration level of 61% of the acceptance criteria was determined from a jet pump strain cauoe at 113% of rated flow. Hence, the data showed that reactor J
internals response to flow induced vibration is within acceptable limits l
up to 113% core flow. As shown in NEDE-24057-P-A, Fitzpatrick is the only plant with instrumented fuel channels. Since the Limerick rated flow per fuel bundle is less than the Fitzpatrick rated flow per fuel bundle, the Fitzpatrick fuel channel test results can be applied to Limerick. An assessment based on Fitzpatrick test data described in NEDE-24057-P-A shows that the maximum recorded vibration of the fuel channels was less than 2% of the allowable for conditions corresponding l
to at least 128% of rated flow. Therefore, the operation of Limerick Unit I at 105% of rated core flow will not result in unacceptable fuel l
rod or fuel channel vibration.
2.4.3 Feedwater Nozzle and Feedwater Sparcer Fatique Usage At the end of the 1970's, inspections at 22 of 23 boiling water reactor plants identified cracking in the feedwater nozzle and sparger at 18 reactor vessels. The NRC staff studied the issue and recommended hard-ware modifications, analysis methods, and inspection schedules for nozzles and spargers in NUREG-0619. Partial feedwater heating will affect the fatigue usage of the feedwater nozzle and sparger. The staff reviewed this request using the guidelines described in NUREG-0619 and the associated Generic letter 81-11.
The licensee uses a GE designed triple themal sleeve sparger to prevent the themal cycling phenomena, thus reducing the likelihood of crack initiation at the feedwater nozzle. One end of the sparger consists of the I
three concentric sleeves with two piston ring seals that are fitted to the l
nozzle safe end, and the other end of the sparger consists of the ams that run alono the vessel wall. The first seal has a clearance fit with the nozzle safe end and foms the primary seal between the innermost sleeve and i
the nozzle bore. The innermost sleeve conducts feedwater from the nozzle i
l l
l r
i
. to the sparger a ms. The primary seal prevents the mixing of relatively cooler feedwater with the hotter reactor coolant. Attached to the middle sleeve is an outer sleeve which is fitted tightly in the nozzle bore. The secondary seal at that tight interference joint reduces potential bypass flow. The staff's original concern about this GE design, as expressed in NUREG-0619, is that wear or corrosion would eventually reduce the sealing ability. GE also mentioned that corrosion of carbon steel safe-ends under the piston-ring seal posed a potential problem.
The use of PFH changes the number of cycles of reactor themal transients, specifically rapid cycling. The rapid cycling is caused by small high frequency temperature changes by mixing of reactor coolant with colder incoming feedwater at the nozzle annulus. Because of PFH the feedwater temperature will be lower than the original design basis and this tempera-ture reduction will increase fatigue usage due to an increase in thermal stresses. The licensee studied two cases of PFH that affect the fatigue usage of the feedwater sparger and nozzle: the final feedwater temperature reduction (FFWTR)andfeedwaterheatersout-of-service (FWHOS). Thermal
(
stresses due to temperature differentials are calculated using the con-duction and convention heat transfer method and stress analysis. Fatigue usage is calculated by dividing the total number of cycles corresponding to thermal stresses by the number of ASME Code allowable cycles. The total fatigue factor is the sum of all of the fatigue usage factors for all transients. This analysis method was described in GE report NEDE 21821-02 and was approved by the staff in NUREG-0619. The licensee has shown that the total fatigue usage factor for the feedwater nozzle and sparger for the FFWTp Case and FWHOS Case can be kept below the required value of 1.0 by seal refurbishment after a 28 year period, based on a postulated number of thermal cycles. Although the refurbishment interval would be reduced, only one refurbishment would be required, as was the case for operation without PFH and ICF. We find this to be acceptable.
2.5 Technical Specification Chances The proposed technical specification changes deal with the MCPR operating limits and certain trip setpoints which are identified below:
(a) Minimum Critical Power Ratio AsdiscussedinSection2.1ofthisSafetyEvaluation(SE),changesto the limiting conditions of operation (LCO) are identified for the proposed operational mode. Based on the staff's review, the operating limit MCPRs of 1.24 and 1.28 are found acceptable. The changes are contained on technical specification (TS) pages 3/4 2-8, 3/4 2-9, and 3/4 2-10; TS pages 2/4 2-8a and 3/4 2-10a are added to reflect the modified operation mode.
(b) Instrumentation Setpoints As discussed in Section 2.1 of this SE, changes to the values for the recirculation pump MG set scoop tube mechanical and electrical stops and
4
. the control rod block recirculation flow upscale trip setpoints and allow-able values are made to accorsodate ICF operation. Also, the rod block monitor upscale trip setpoint and allowable values are changed to accom-modate the addition of a high flow-clamped rod block. Based on the results of our review the staff finds the proposed changes acceptable. The changes are contained on TS pages 3/4 4-2, 3/4 3-60a and 3/4 3-60, respectively.
(c) Administrative Changes
. The index was updated to reflect the additional pages and the General Electric analyses document (Ref. 2) was added as a reference. A reference to Figure 3.2.3-1 on TS Basis page B 3/4 2-4 has been chanced to reflect the division of Figure 3.2.3-1 into Figures 3.2.3-la and 3.2.3-1b.
2.6 Conclusion The NRC staff has reviewed the information provided by the Philadelphia Electric Company relative to the proposed license amendment to allow operation of the Limerick Generating Station Unit I with partial feedwater heatino and increased core flow. Based on the results of the evaluation contained in this section the staff concludes that the proposed technical specification changes are acceptable.
3.0 ENVIp0NMENTAL CONSIDERATION This amendment involves a change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance require-ments. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public connent on such finding. Accordingly, this amendment meets the eligibility criteriaforcategoricalexclusionsetforthin10CFR51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement ncr environmental assessment need be prepared in connection with the issuance of this amendment.
4.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Consission's regulatiers and the issuance of this amendment will not be inimical to the conson defense and the security nor to the health and safety of the public.
Principal Contributors:
M. McCoy, R. Li, J. Tsao Dated:
February 17, 1987
REFERENCES Letter, E. J. Bradley (PECo) to H. R. Denton (NRC) dated November 17, 1.
1986 transmitting application for Amendment to Facility Operating License NPF-39, NEDC-31343, " Increased Core Flow and Partial Feedwater Heating Analysis 2.
for Limerick Generating Station Unit 1 Cycle 1", dated October 1986.
3.
Letter, J. W. Gallagher(PEco) to W. R. Butler (NRC) dated December 22, 1986.
Letter, M. J. Cooney(PEco) to W. R. Butler (NRC) dated January 2, 1987.
4.
" General Electric Standard Application for Reactor Fuel (Supplement for 5.
US)", May 1986 (NEDE-24011-P-A-8-U5, as amended).
General Electric Service Information Letter No. 380, Revision 1 6.
February 10, 1984.
G. A. Watford, " Compliance of the General Electric Boiling Water Reactor 7.
Fuel Designs to Stability Licensing Criteria," October,1984 (NEDE-22277-P-1).
Letter, J. W. Gallagher (PECo) to W. R. Butler (NRC) dated January 29, 8.
1987.
I