ML19269E794
| ML19269E794 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 06/25/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19269E789 | List: |
| References | |
| NUDOCS 7906290645 | |
| Download: ML19269E794 (12) | |
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hg UNITED STATES 8
NUCLEAR REGULATORY COMMISSION g
WASHINGTON, D. C. 20555 3
- j
\\...J l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.16 TO PROVISIONAL OPERATING LICENSE NO. DPR-45 DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACBWR)
DOCKET NO. 50-409 1.0 Introduction By letter dated February 26, 1979, as supplemented May 9 and 23, 1979, Dairyland Power Cooperative (DPC) requested an amendment to Provisional Operating License No. DPR-45. The amendment would modify the Technical Specifications for the La Crosse Boiling Water Reactor (LACBWR) to permit operation in fuel Cycle 6.
2.0 Discussion Dairyland Power Cooperative has proposed changes to the Technical Speci-fications of the Lacrosse Boiling Water Reactor (LACBWR).
(References 1,2,5).
The proposed changes relate to the replacement of 28 fuel assemblies constituting refuel'.4 of the core for sixth cycle operation at power levels up to 165 MWt (100% power).
In support of the reload application. the licensee has provided the documentation listed in the reference section. The fuel loading for cycle 6 will consist of 26 Exxon type III fresh fuel assemblies, two Allis-Chalmers type I fresh fuel assemblies, 32 Exxon type III fuel assemblies irradiated for one cycle, and 12 Allis-Chalmers type II fuel assemblies irradiated for more than one cycle. The 12 type II fuel assemblies will be located in the periphery of the core where they will experience fairly low maximum power and minimal effects of power peaking and control rod movements. The 32 exposed type III assemblies will be intermixed in a checkerboard fashion with the 28 fresh fuel assemblies to control radial power peaking and to limit power density increases.
790629069 6,
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. Because the LACBWR is unique from other BWRs reviewed by the staff, some of the more basic differences are listed below:
- The Lacrosse BWR operates with a core-average void fraction of 21%, which is lower than other BWRs, due to its high operating pressure. This results in a smaller moderator density differencs between in-channel and between-channel, and thus has a marked effect on local peaking within fuel assemblies.
- The fuel cladding is stainless steel, resulting in poorer neutron economy, a higher initial enrichment, and a lower conversion ratio.
- The core is quadrant symmetric, but the symmetry lines are tilted 45* from the core " flats," rather than running parallel to the fl ats.
- The Lacrosse core is small, consisting of 72 fuel assemblies with 29 control blades.
Instead of the more common " checkerboard" patterns, the reactor is operated with all but eight centrally located blades fully withdrawn at beginning-of-cycle (B0C).
LACBWR uses a combination of Zircaloy and stainless steel channel boxes. The stainless steel channel boxes aid in the control of Beginning-of-Cycle (B0C) excess reactivity. Cycle 6 will use eight stainless steel channel boxes for this purpose.
(Reference 1).
- Although the Lacrosse reactor has an incore power distribution mn5itoring system, it is not used for licensing purposes.
- Instead, a design peaking factor is conservatively precalculated for each cycle, and no credit is taken for the lower operating peaking factor.
3.0 Evaluation 3.1 Nuclear Design For Cycle 6 operation, the lead burnup assembly within the core will not exceed a burnup in excess of 15,000 MWD /MTU. With a BOC-6 core average exposure of 4203 MWD /MTU this will result in a cycle length of approximately 3931 MWD /MTU and a estimated core average exposure of 8134 MWD /MTU at E0C-6.
Because of the above burnup limit, the End-of-Cycle (E0C) condition for LACBWR will consist of a number of partially inserted control rods and the reactor will have the potential for approximately 2300 MWD /MTU of remaining full power capability (Reference 2).
Therefore, where the severity of the transient analyses, as described in Section 3.3, are exposure dependent and calculated based on E0C-AR0 (end of cycle - all rods out) condition, the analyses results are conservative.
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. The neutronic input for the transient analysis consists of: moderator void coefficient, moderator temperature coefficient, doppler coefficient, neutron lifetime, and delayed neutron fraction.
A conservative value for each of these parameters was applied to the Cycle 5 analysis to bound all future cycles having the Exxon reload fuel (Reference 2).
The results of these analysis remain acceptable for Cycle 6 operations.
To select the B0C-6 rod pattern and withdrowal sequences the licensee performed a series of calculations to determine the axial power distribu-tion and peaking factors throu]hout the cycle (Reference 1).
The initially selected rod patterns and fuel loading described in Reference 1 were rotated 90 for purposes of equalizing exposure of the control rods.
Since the fuel loading pattern and control rod patterns are both rotated together, (Reference 2), the analysis results are not affected by this The restrictions on power escalation rate and on control rod change.
movement which were used, and approved in Cycle 5 will also be used in Cycle 6 (Reference 1).
The " worst case" cold shutdown margin at BOC-6, with the most reactive control rod out, was calculated to be greater than the 0.5% Ak/k Technical Specification limit.
In addition, the " worst case" ejected rod at B0C, for full power and full flow conditions, was determined to te less than the 2.5% tK/k Technical Specification limit.
Based on the above, the staff finds the nuclear design of the Cycle 6 core acceptable.
3.2 Mechanical Design The staff's review and approval for the use of the Exxon, type III, fuel in the LACBWR was first issued for Cycle 5 operation (Reference 4).
In Reference 4 we also addressed the mechanical designs of the Allis-Chalmers fuel and the Exxon fuel. Based on our earlier review and approval, (Reference 4) and Cycle 5 operating experience, (Reference 2) we find the mechanical design of the reload fuel proposed for Cycle 6 acceptable.
3.3 Thermal-Hydraulics The liccasee has stated that all plant parameters that are relevant to the thermal-hydraulic analysis are either unchanged from Cycle 5 to Cycle 6, or were calculated in the Cycle 5 analysis such that they bound all future cycles at LACBWR (Reference 2).
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. The licensee's calculated results are based on the XN-2 correlation with critical power ratio (CPR) for the thermal-hydraulic analyses. We con-cluded in Reference 6 that the XN-2 correlation could be used for the mass flux of 0.08 to 2.4E6 lb/hr-ftgPR of 1.32 within the following ranges:
Lacrosse core with a safety limit M
, pressure of 615 to 1500 psia, and inlet subcooling of -45 to 450 Btu /lb. The safety limit of 1.32 was im-posed to assure that during transients there is a 99/99 reliability / con-fidence level that the fuel rods will not experience transition boiling.
3.3.1 Operating Limit MCPR at Rated Power and Flow Various transient events can result in a reduction in the CPR.
To ensure that the fuel cladding integrity safety limit, as determined by the minimum critical power ratio (MCPR=1.32), is not violated during anticipated abnormal operational transients, the most limiting transients.. ave been analyzed to determine which transient results in the largest reduction in critical power ratio ( ACPR).
The licensee submitted the results of analyses for those transients which produce a significant decrease in MCPR in References 7 and 8.
The types of transients evaluated were overpressure, moderator tempera-ture decrease, coolant flow increase and decrease, coolant inventory decrease and reactivity insertion.
The most limiting transient for Cycle 5 was the Rod Withdrawal Error (RWE). The operating limit MCPR for the type II and type III fuel from the Cycle 5 analysis was 1.59.
Fcr the Cycle 6 RWE analysis, the operating limit MCPRs were determined to be 1.73,1.49, and 1.50 for the type I, type II, and type III fuel respectively. Therefore, the licensee has proposed to retain the operating limit MCPR of 1.59 for the type II and type III fuel based on the more conservative Cycle 5 RWE analysis (Reference 1).
While reviewing LACBWRs transient analysis, (Reference 7) it was noted that the turbine trip and generator load rejection pressurization transients with postulated failures of the bypass valves were not performed.
- Instead, the licensee addressed these events as being similar to the MSIV closure overpressure analysis which was provided. Even though we agree that phenomenalogically these events may be similar, the effects on the calculated A CPR will differ significantly (see Section 3.4).
- However, since these pressurization transients are only limiting at, or near, the (E0C-AR0) conditions, the LACBWR E0C condition based on the 15,000 MWD /MTU burnup limit (see Section 3.0) will preclude the two pressurizations transients as limiting events.
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. Based on the above, the staff finds the proposed operating limits of 1.73,1.59 and 1.59 acceptable for the type I, type II, and type III fuel respectively during Cycle 6 at LACBWR when the plant is operating within the specified 15,000 MWD /MTU burnup limit discussed in Section 3.0.
3.3.2 Operating Limit MCPR for Less than Rated Power and Flow For core flows somewhat less than rated recirculation flow, the operating limit MCPR must be increased. The most limiting transient at lower than rated power and flow conditions is the speedup of both recirculation The licensee pumps caused by a failure of the motor control system.
has analyzed this transient and provided the resulting curves of operating limit MCPR versus core flow in the proposed Technical Specifications.
At flows less than about 80% of rated core flow, different operating limits are specified for the type I, type II, and type III fuels.
We conclude that operating with MCPRs equal to, or greater than, those proposed will asure that the MCPR will not exceed the safety limit for the most severe operational transients.
3.4 Overpressure Analysis In Reference 9, the licensee presented the results of an overpressure analysis to demonstrate that an adequate margin exists below the ASMEThe code allowable vessel pressure of 110% of vessel design presssure.
transient' analyzed was the MSIV closure, and the following assumptions were made.
(1)
Initial power was 102%
(2) No credit for the MSIV-closure scram (3) No credit for the power-flow scram (4) No credit for the overpower scram (5) No credit for the shutdown condenser (6) High pressure scram at 1340 psia (7) First safety valve opens at 1419 psia and discharges 294,000 lb/hr.
The maximum pressure attained is 1419 psia, which is significantly below Addf-the ASME code allowable vessel pressure of 1555 psia for LACBWR.
tional overpressure protection is provided by two more safety valves set at 1441 psia, which did not open in the MSIV-closure analysis.
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. The above discussion of the MSIV analysis was provided in our review and approval of the Cycle 5 reload submittal (Reference 4).
As discussed in Section 3.3, these analyses are also applicable to Cycle 6 and therefore remain acceptable. The purpose of repeating our Cycle 5 eva%ation is to call attention to the sequence of events assumed in the V,IV analysis for comparison with the licensee's turbine trip or generator load rejection transient analysis without failure of the bypass valves (Reference 7).
From the above MSIV sequence, the reactor scram is initiated by the high pressure signal and the resultant ACPR was calculated as 0.09.
In the licensee's analysis of the turbine trip transient without bypass failure, the bypass valves were assumed to begin opening in less than one second resulting in resumption of steam flow. The resu'. tant ACPR for the turbine trip with bypass was calculated to be 0.16.
Examination of oCPRs for these two transients show that the ACPR for MSIV event is lower than the ACPR for the turbine trip with bypass.
For a turbine trip with failure of the bypass valves, the void collapse and pressure rise in the vessel can be expected to occur more rapidly than the MSIV event because of the 6.5 second closure time of the MSIVs.
The resultant neutron flux and subsequent clad surface heat flux which result from the turbine trip without bypass can be expected to exceed the values obtained from the turbine trip with bypass. This is because the bypass opening results in steam flow which reduces the pressure, increases the voiding, decreases the neutron flux, and lowers the fusi clad surface heat flux. Therefore the ACPR for the turbine trip without bypass can be expected to exceed the ACPRs for either the MSIV event, or the turbine trip with bypass. As stated in Section 3.3, the E0C-AR0 (end of cycle - all rods out) condition is more severe for the pressurization transients because the time required to provide negative reactivity to the high flux region by insertion of the control rods is longer when the rods are fully withdrawn.
Therefore, the MSIV analysis, as provided by the licensees (Reference 9),
is acceptable for determining the maximum overpressure event, but it does not bound the ACPR for the turbine trip or generator load rejection with failure of the bypass valves at E0C - ARO conditions. However, because these pressurization transients are only limiting in ACPR at, or near, the (E0C - AR0) conditions, we find that the ACPR established by the RWE (see Section 3.4) bsunds the ACPR for these pressurization transients during the proposed cycle length at LACBWR.
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. 3.5 Thennal Hydraulic Stability Hydraulic instability is most likely to occur at natural circulation.
A low flow trip in LACBWR at 30% flow provides protection from operating at natural circulation (Reference 1). Thus, there is sufficient -
margin between the minimum recirculation flow allowed and natural circulation. We find this acceptable.
3.6 Accident Analysis 3.6.1 ECCS Analysis Our review and approval of the LACBWR ECCS analysis was provided in Reference 4.
Curves of limiting Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) versus exposure are incorporated in the Technical Specifications (Reference 1). Operation at full rated power with fuel assembly average planar heat generation rates below these MAPLHGR limits will ensure that the peak cladding temperatures will not exceed the Interim Acceptance Criteria limit of 23000F and that less than 17% of the cladding will react with steam at any location.
We conclude that the ECCS analysis and results for Cycle 5 are applicable for Cycle 6 and therefore acceptable.
3.6.2 Fuel Loading Error The licensee has examined the effects on ACPR and the linear heat generation rate (LHGR) for the " worst case" fuel loading error (Reference 1). The methods of analysis used are the same as the methods used and approved in Cycle 5.
Based on the results of the licensee's analysis, the " worst case" fuel loading error will not result in a penetration.of the 1.32 safety limit. We find this acceptable.
In addition to the effect of a " worst case" misloaded fuel assembly on the minimum critical power ratio (MCPR) safety limit, the licensee examined the effect of potential misloadings on the cold shutdown margin assuming no xenon, and zero power. The results of this analysis indicated that if one of the two fresh Type I fuel assemblies was misplaced in one of the four corner positions, the shutdown margin would be reduced below the Technical Specification requirement.
Based on discussion with the licensee on their sequence of fuel loading and core verification, we have concluded that the potential for this particular misloading is highly unlikely because these specific locations were already occupied with irradiated fuel before any fresh fuel was loaded. Therefore we find this condition, and the precautions taken by the licensee to eliminate this particular misloading to be acceotable for Cycle 6 ooerations.
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3.6.3 Control Rod Drop Accident The neutronic input for the control rod drop accident analysis is consistent with a core full of Type III fuel.
The only shutdown phenomenon allowed was the osppler coefficient.
(References 1,3).
The use of the type III fuel input parameters conservatively bound the analysis for all cycles having the Exxon fuel (Reference 2).
The licensee has also submitted the "LACBWR Rod Drop Probability Study" (Reference 3), for staff review. This study analyses the events and probabilities of a rod drop accident necessary to exceed the design basis fuel enthalpy deposition of 280 calories per gram.
Our review and approval of the probability study is not yet complete.
However, because of the conservatisms used in the control rc.i drop accident analysis, and the magnitude of the results (peak fuel enthalpy <_280 cal /gm), we find that the calculational met. hods used by the licensee to calculate physics parameters in cycles 4 and 5 remain acceptable for cycle 6.
4.0 Physics Startup Test Program The safety analysis for the upcoming cycle is based upon a specifically designed core configuration. We also have assumed that, aftv reloading, the actual core configuration will conform to the designed cot. figuration.
A startup test program can provide the assurance that the core confonns to the design.
The proposed physics startup test program for Cycle 6 includes control rod drive functional and scram tests, verification of predicted critical rod pattern, a shutdown margin verification and comparison of incore data with appropriate predictions.
In the future, we anticipate requiring a description of each test and the procedures used to verify the core configuration. This information must be sufficient to provide assurance that the core conforms to the design.
The description is anticipated to include both the acceptance criteria and the actions to be taken in case the acceptance criteria are not obtained. The description of the procedure to verify core configuration usad for Cycle 6 is acceptable for the reasons given in Section 3.6.2.
In addition to the requirements, above, we have requested and the licensee has agreed to provide a written report of the startup test to the NRC within 45 days of the completion of the tests.
Although we cannot at this time approve the licensee's program for all future cycles, we have examined it sufficiently to conclude that the physics startup test program is acceptable for cycle 6.
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. 5.0 Cycle 5 Fuel Performance The licensee identified a total of 17 defective fuel assemblies at the end of cycle 5 (reference 2). The number and extent of failed assemblies is within that expected and predicted by(the NRC staff in our Safety Evaluation supporting Cycle 5 operation reference 4).
Based on the infonnation supplied by the licensee, we have determined that the observed failures are bounded by the existing analysis for Cycle 5.
For Cycle 6, additional measures have been taken by DPC to minimize fuel failures, as discussed in Section 2.0.
These measures include loading the remaining 12 irradiated Type II Allis-Chalmers fuel in For reasons discussed in Section 2.0, DPC the periphery of the core.
has concluded, and the staff agrees, that these measures will tend to minimize fuel degradation of the type II fuel during Cycle 6 operations.
It was noted that at the beginning of cycle 5, some fissile material was present in the core as a result of the fuel failures from Cycle 4.
This fissile source produced high alpha activity in the coolant and relatively The gradual removal of this fissile high off-gas activity at the 80C-5.
source by the cleanup system was evident in Alpha activity measurements taken during cycle 5 operations (reference 2). The high initial off-gas tended to mask the development of fuel defects during cycle 5.
Therefore, at our request, the licensee proposed to modify the off-gas limits to account for this removal.
The licensee also noted that none of the fuel in LACBWR at E0C-5 was even approaching the condition of " grossly" failed fuel observed at the E0C-4 (Reference 2). Therefore, with a much cleaner syste.n expected at the B0C-6, the off-gas activity and primary coolant activity at B0C-6 should be considerably lower.
To better monitor the condition of the fuel during cycle 6 operation, such that the B0C activity levels will not mask the development of fuel failure indication (off-gas activity) the licensee has proposed changes to the LACBWR Tbchnical Specifications on the primary coolant and off-gas activities (Reference 5).
We have reviewed the changes proposed by the licensee to provide a more effective means of monitoring potential fuel failures at LACBWR.
Based on the results of cur analysis of the fuel failures at E0C-4 (Reference 4) which bound the fuel failures observed in Cycle 5 (Reference 2), we have detennined that the Technical Specification changes proposed by the licensee (Reference 5) are acceptable.
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. 6.0 Additional Technical Specification Changes In our review of the May 2,1979 transient event at the Oyster Creek BWR (Reference 10), we have identified two additional Technical These are Specification requirements that are applicable to LACBWR.
(1) a Technical Specification on the minimum number of non-isolated forced circulation loops is required, and (2) a reactor water level safety limit Technical Specification must be established.
The event at Oyster Creek resulted in a substantial decrease in core water level because of partial isolation of all recirculation lines during the transient. The major concern which arose from this event is that the reactor core could be effectively isolated from its source of reactor coolant, the annulus region. Therefore, we have concluded that a Technical Specification on the minimum number of non-isolated forced circulation loops is required at LACBWR to assure adequate comunication between the core and annulus We have reviewed analyses of natural circulation through regions.
loops similar to those of LACBWR and have also evaluated the specific LACBWR forced circulation loop and reactor vessel internals configurations.
Based on these evaluations, we have concluded that one open loop will provide adequate comunication between core and annulus regions at LACBWR. The requirement for the non-isolation of one forced circulation loop, under all conditions where possible depletion of core inventory may occur, i.e., shutdown condenser operation and MSIV non-isolation, has been incorporated into the Technical This Specificatiots as Specification 4.2.2.6 on the above bases.
chnge has been discussed with and agreed to by the licensee.
A second requirement for LACBWR that the staff has determined to be necessary is to extend the existing reactor vessel water level The safety limit to apply to all modes of reactor operation.
Technical Specification 4.0.2.1.4 applies existing safety limitq only in the shutdown andg efueling modes. We have concluded that extending the applicability of this safety limit will assure that steps are taken by the licensee to maintain water level above a height equivalent to that of the top of the active fuel during all modes of operation. This change has been discussed with and agreed to by the licensee.
2143 247 7.0 Environmental Considerations We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
. 8.0 Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date: May 25, 1979 2143 248
References 1.
Letter from F. Linder (DPC) to D. L. Ziemann (NRC), LAC-6130 transmitting LAC-TR-067, Refueling Plan for Cycle 6 of LACBWR and Proposed Tachnical Specifications, February 26, 1979.
2.
Letter from F. Linder (DPC) to D. L. Ziemann (NRC), LAC-6274 transmitting additional information and LAC-TR-068, LACBWR Cycle 5 Fuel Performance and Finalized Refueling Plan for Cycle 6, dated May 9,1979.
3.
Letter from F. Linder (DPC), to D. L. Ziemann (NRC), LAC-6184, transmitting "LACBWR Rod Drop Probability Study", dated March 29, 1979.
4.
Letter from R. Reid (NRC) to J. Madgett (DPC) transmitting Amendment No.11, and Safety Evaluation Report for Cycle 5 Operation of LACBWR, dated March 3,1978.
5.
Letter from F. Linder (DPC) to D. L. Ziemann (NRC) transmitting proposed Technical Specification Changes, dated May 23, 1979.
6.
Memorandum from D. F. Ross to K. R. Goller, " Thermal-flydraulic Review of LACBWR Reload," May 12, 1977.
7.
" Supplement No. 3 - Part II, LACBWR" and " Response to Question 4, Transient Analysis for LACBWR Reload Fuel," letter from John P. Madgett (DPC) to Robert W. Reid (NRC), LAC-4523, dated February 25, 1977.
8.
Letter from J. P. Madgett (DPC) to Robert W. Reid (NRC), LAC-4935, dated October 5, 1977 with attached LAC-TR-057 requesting approval for use of 32, Type III, Exxon assemblies in Cycle 5 reload configuration.
9.
Letter from John P. Madgett (DPC) to Robert W. Reid (NRC), LAC-4654, dated April 27, 1977.
10.
Letter from I. R. Finfrock, Jr., Jersey Central Power and Light Company, to Director of Nuclear Reactor Regulation, USNRC, May 12, 1979, "0yster Creek Nuclear Generating Station Docket No. 50-219 Transient on May 2, 1979."
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