ML19270G422
| ML19270G422 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 04/24/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19270G421 | List: |
| References | |
| NUDOCS 7906070013 | |
| Download: ML19270G422 (13) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 43 TO OPERAllNG LICENSE NO. UPR-19 COMMONWEALTH EDISON ORESDEN NUCLEAR POWER STATION UNIT 2 DOCKET NO. 50-237 1.0 Introduction Commonwealth Edison (the licensee) has proposed changes to the Technical Specifications of Operating License DPR-19 for Dresden Nuclear Power Station, Unit 2.
The changes are proposed in con-junction with the replacement of 160 fuel assemblies constituting refueling of the core for seventh cycle operation at power levels up to 2527 MWt (100% power).
In support of the application, the licensee has provided the General Electric (GE) BWR Reload 4 licensing submittal for Dresden 2 (Refer-ence 1), proposed Technical Specification changes (Reference 2),
information on the Dresden 2 Loss of Coolant Accident (LOCA) analysis (Reference 3), and additional information related to the refteling (References 4, 5, 19 and 20).
This reload is the first for Dresden 2 which involves loading of GE 8x8 retrofit (8x8R) fuel. The description of the nuclear and mechan-ical design of the Bx8R fuel and the older design 8x8 fuel is contained in GE's licensing topical report for BWR reloads (Reference 6).
Reference 6 also contains a complete set of references to topical reports which describe GE's analytical methods for nuclear, thermal-hydraulic, transient and accident calculations, and information regarding the applicability of these methods to cores contcining 7x7, 8x8, and 8x8R fuel.
Portions of the plant-specific data, such as operating conditions and design parameters which are used in transient and accident calculations, have also been included in Reference 6.
The staff's safety evaluation (Reference 8) of the GE generic reload licensing topical report has concluded that the nuclear and mechanical design of the 8xBR fuel, and GE's analytical methods for nuclear, thermal-hydraulic, and transient and accident calculations as applied to mixed cores containing 7x7, 8x8, and 8x8R fuel are acceptable.
Approval of the nuclear and mechanical cesign of 8x8 fuel was origi-nally cased en information in Reference 7 and expressed in the staff's evaluation (Ref erence 9) of that document.
7906070 h, e
. Based on the staff's review, the plant-specific input data for tran-sient and accident analyses presen*ed in Reference 6 are acceptable Additional plant and cycle-dependent data and informa-(Reference 8).
tion are provided in Reference 1, which closely follows the outline of Appendix A of Reference 6.
Because of the staff's review of a large number of generic considera-tions related to use of 8x8R fuel in mixed loadings with 8x8 and 7x7 fuel, and on the basis of the evaluations which have been presented in Reference 8, only a limited number of additional areas of review These include have been included in this safety evaluation report.
the plant and cycle-specific input data and results presented in Reference 1, the physics startup test program described in Reference 5, the proposed modifications to the Technical Specifications, and those items requiring special attention during reload reviews.
For evaluations of issues not specifically addressed in this safety evaluation report, the reader is referred to Reference 8.
2.0 Evaluation 2.1 tiuclear Characteristics For Cycle 7 operation of Dresden 2,160 fresh 8x8R bundles of type The remainder 265L will be loaded into the core (References 1 and 4).
of the 724 fuel bundles in the core will be 7x7 and 8x8 fuel loaded for earlier cycles.
Approval of the nuclear characteristics of these fuel types has been given in References 8 and 9.
For the core design presented in Reference 4, the licensee has eval-uated the shutdown capabilities of their control rod system and stanoby liquid control system during Cycle 7.
Based upon our review of these evaluations, we conclude that these shutaown capabilities are acceptable throughout Cycle 7.
2.2 Thermal-Hydraulics 2.2.1 Fuel Cladding Integrity Safety Limit NCPR As stated in Reference 8, for BWR cores which reload with GE's retro-fit 8x8R fuel, the allowable minimum critical power ratio (MCPR) resulting from either core-wide or localized abnormal operational W
. transients, is equal to 1.07.
With this MCPR safety limit, at least 99.9% of the fuel rods in the core are expected to aveid boiling transition during these transients.
The 1.07 safety limit minimum critical power ratio (SLMCPR) proposed by the licensee represents a.01 increase from the previous 1.06 SLMCPR. The basis for the revised safety limit is addressed in Reference 6.
This change continues to meet the recommendations of Standard Review Plan 4.4 and on that basis has been found acceptable in Reference 8.
Modifications to the Technical Specification have been incorporated per this finding.
2.2.2 Operating Limit MCPR Various transient events could reduce the MCPR from its normal opercting value. To assure that the fuel cladding integrity safety limit MCPR will not be violated during any abnormal operational transient, the most limiting transients have been reanalyzed by the licensee to determine which event results in the largest reduction in critical power ratio.
Each of the events has been conservatively analyzed for each of the fuel types and for the full range of exposure through the cycle.
The transient events analyzed were load rejection without bypass, feedwater controller failure, loss of 145*F feedwater heating and control rod withdrawal error.
The calculational methods, which include cycle-independent initial conditions and transient input parameters, are described in Refer-ence 6.
Our acceptance of the values used and related transient analysis methods appear in Reference 8.
Supplemental cycle-dependent and transient input parameters used in the analysis appear in the tables in Section 6 and 7 of Reference 1.
Our evaluation of the methods used to develop these supplementary transient input values have already been addressed and appear in Reference 8.
The overall transient methodology, including cycle-independent transient analysis inputs, provides an adequately conservative basis for the determina-tion of transient reductions in CPR.
Calculated system responses and reductions in CPR during each of the operational transients have been provided in Reference 1.
The fol-lowing table gives the limiting CPR reduction, the event for which the limiting reduction in CPR occurs, and the required operating limit MCPR for each fuel type:
a
_4 Operating Fuel Type Most Severe CPR Reduction Limit MCPR 7x7 0.17 (Load Rejection Without Bypass) 1.24 8x8 0.24 (Load Rejection Without Bypass 1.31 and Control Rod Withdrawal Error) 8x8R 0.24 (Load Rejection Without Bypass 1.31 and Control Roa Withdrawal Error)
Thus, when the reactor is operated in accordance with the above operating limit MCPRs the 1.07 SLMCPR will not be violated in the event of the most severe abnormal operational transient.
This is acceptable to the staff per the finding of the previous section.
On this basis, operating limit MCPR Technical Specifications have been established.
In the analysis of the rod withdrawal error (RWE), flow biased upscale rod block monitor (RBM) setpoints are established to assure that the safety limit MCPR is satisfied. 'Therefore, this setpoint is specified in the Technical Specifications. On the basis of the acceptance of RWE analysis methods in Reference 8, we find the calculated CPR reduc-tion and RBM setpoint for the RWE acceptable.
Fuel loading errors which could result in violation of the MCPR limit have been analyzed (Reference 1) by methods described in Reference 6 and approved by the staff in Reference 8.
Reference I states that these analyses show that no fuel loading error could lead to a MCPR less than 1.07.
Therefore, based on the licensee's analyses of anticipated transients and fuel loading errors using acceptable methods and input, and con-sidering the incorporation of an appropriste rod block monitor set-point into the Technical Specifications, we conclude that the operating limit MCPRs proposed for Cycle 7 are acceptable.
8
. 2.3 Accident Analysis _
2.3.1 LOCA Analyses and MAPLHGR Limits The introduction of 8x8R fuel into the Dresden 2 core for the time has required addition of MAPLHGR limits for the 8 DRB 265L fue Concurrent with the intro-to the plant Technical Specifications.
duction of 160 8x8R fuel bundles, the licensee has proposed to raise MAPLHGR limits for some of the other fuel types in the core based on the improved reflood characteristics of BWR/3 The new limits have been provided in Reference 2 and 6.
Dresden 2 is one of four Commonwealth Edison BWRs w Quad Cities Unit 1, which plant analyses for the Duane Arnold BWR/4.is another of the fuel bundles and requested new MAPLHGR limits Staf f acceptance of the Refer-analyses (Reference 3) as Dresden 2.
ence 3 MAPLHGRs for Quad Cities 1 is documented in Refe Except for the dif fering numbers of 8x8R fuel bun Because the with regard to ECCS and LOCA response characteristics.
Reference 3 analyses assumed 156 8x8R drilled b BWRs, these limits are expected to be conservative for core > with better reflooding characteristics (i.e., for Quad Cities 1 with 192 drilled fuel bundles or for Dresden 2 with 160).
The Reference 3 analyses and the proposed MAPLHGR limits are based on the assumption that the worst single f ailures during small and large break LOCAs for loop selection logic plants such as Dresden 2 are the f ailure of a HPCI pump and the f ailure of a LPCI syste jection valve respectively.
to staff concerns that the loss of a DC power source could be a more limiting single failure for either large or small break LOCAs thanG l
the failures previously assumed.
plant analyses, that for Dresden 2 type plants peak clad temperature I
(PCT) during the limiting small break LOCA may be larger if a DC
power source f ails than if a HPCI pump fails, but that the DBA break j
PCT would be limiting overall.
They have also concluded that the limiting (DBA) LOCA PCT with f ailure of a LPCI system injection valve Plant would be higher than that assuming f ailure of a DC power source.
specific analyses for Dresden 2 were not referenced by GE in Refer-l ence 16.
I The staf f is reviewing the acceptability of these conclusions on i
We have also obtained additional information on the a generic basis.
subject related specifically to Dresden 2 (References 19 and 20).
The additional information includes results of Dresden 2 Cycle 7 The specific LOCA analyses which assumed loss of a DC power source.
licensee stated that the ECC systems assumed operable in the Reference 16 analyses would be operable for Dresden 2 if a DC power source should These conclusions depend upon the findings by GE and the licensee fail.
regarding the ECC systems which would remain available upon loss of a DC 3
power source.
g Based on our review of Reference 16 to date, and the additional informa-tion in Reference 19, we conclude that for large break LOCAs the margin between the Reference 3 Dresden 2 PCT and the bounding plant PCT pre-dicted by GE in Reference 16 is large enough that the Reference 3 PCT We also will remain valid af ter completion of our generic review.
conclude that for small break LOCAs at Dresden 2 the PCT will not exceed e
the value obtained in the Reference 20 analyses.
In summary, we believe that the MAPLHGRs proposed for Cycle 7 adequately take into account the ef fects of drilled lower tie plates in the 8x8R fuel as addressed in Reference 15, and that all requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50.46 will be met when Dresden 2 is oper-ated in accordance with these MAPLHGRs.
We, therefore, find that the proposed changes to the Technical Specification MAPLHGRs are acceptable, under the condition stated in the previous paragraph.
2.3.2 Control Rod Drop Accident The analysis of the control rod drop accident (CRDA) has been per-formed on a generic (counding analysis) basis.
In our safety evalua-tion (Reference 6) of GE's generic reload methods (Reference 6), we concluded thi.t the bounding analysis basis is acceptable with the provision that the key input parameters for a plant-specific reload are conservatively bound by the analysis assumptions.
In the plant-specific reload application (Reference 1), the licensee has shown that the maximum incremental control rod worth is conservatively represented in the bounding analysis, which is acceptable.
8
. 2.3.3 Fuel Loading Error The effect of postulated fuel loading errors on fuel thermal limits has been analyzed by the licensee using the new ana!ytical methods developed by GE (Ref erences 10 and 11) and approved by the staf f (Reference 12).
These analyses indicate that the safety limits on MCPR and LHGR will not be violated by the most severe fuel misloading which could occur during Dresden 2 Cycle 7 operation.
Based on the use of accepted methods and the result that safety limits will not i
I be violated, we find the operation of Dresden 2 during Cycle 7 ac-ceptable with regard to fuel loading errors.
2.4 Overpressure Analysis The overpressure analysis for the MSIV closure with high flux scram, which is the limiting overpressure event, has been performed in accordance with the requirements of Reference 8.
As specified in Reference 8, the sensitivity of peak vessel pressure to f ailure of one safety valve has also been evaluated.
We agree that there is sufficient margin between the peak calculated vessel pressure and the design limit pressure to allow for the failure of at least one valve. Therefore, the limiting overpressure event as analyzed by
)
the licensee is considered accepta'ble on the bases outlined in j
Reference 8.
2.5 Thermal-Hydrau.jg Stjbility A thermal-hydrauli t etability analysis was performed with the methods described in Reference 6.
The results show that the channel hydro-dynamic and reactor cor; decay ratios at the least stable operating state (corresponding to the intersection of the natural circulation curve and 100% rod line on the powcr-flow map) are below the 1.0 Ultimate Performance Limit decay ratio proposed by GE.
The staf f has expresssed generic concerns regarding reactor core thermal-hydraulic stability at the least stable reactor condition.
This condition could be reached during an operational transient from high power if the plant were to sustain a trip of both recirculation The concerns are motivated by increasing pumps without a reactor trip.
decay ratios as equilibrium fuel cycles are approached and as reload fuel designs change. The staf f concerns relate to both the conse-quences of operating at a decay ratio of 1.0 and the capability of the analytical methods to accurately predict decay ratios.
- The General Electric Company is addressing these staff concerns through Although a meetings, topical reports and a stability test program.
final test report has not as yet been received by the staf f for review, it is expected that the test results will aid considerably in resolving the staff concerns.
For a previous operating cycle, the staff, as an interim measure, added a requirement to the Technical Specifications which restricted ontinuation of c
planned operation in the natural circulation modc this restriction will also provide a significant increase in the reactor core stability operating margins for the current cycle so On the basis that the decay ratio is (1.0 in all operating modes.
of the foregoing, the staff considers the plant thennal-hydraulic stability characteristics to be acceptable.
2.6 Startup Testing A startup test program for Cycle 7 has been submitted ay the licensee (Reference 5).
The purpose of the startup t :,t; in to provide assur-
'.' h e l i -
ance that the reload core has been assembled as inte.'ded.
censee's proposed test program and acceptance criteria provide such assurance and are acceptable for Cycle 7 operation.
2.7 End of Cycle Power Coastdown The licensee has proposed an allowance in the Technical Specifications to operate Dresden 2 in the coastdown mode when operation at full As justification for this proposa!, the power is no longer possible.
licensee has referenced transient analyses performed to justify the Quad Cities Unit 1 Cycle 2 coastdown and the Dresden 2 Cycle 6 coast-down (Reference 19).
These transient analyses (References 17 and 18), performed for non.inal f
recirculation flow and feedwater heating, indicate that decreases in
[
reactor power and in the void reactivity coef ficient during coastdown compensate for the reduction in control rod scram effectiveness with the i
result that potential transient MCPR reductions from full power operation i
tend to be larger than those possible during coastdown.
Based on the trends indicated in the coastdcan node tranalent ana'yres submitted by the licensee, we believe that the MCPR limits accepted in Section 2.2 for rated power and nominal recirculation flow and feedwater heating are also acceptable for coastdown operation at nominal recircula-tion flow and f eedwater heating.
We, therefore, consider the proposed coastdown operation at the end of Cycle 7 to be acceptable.
$1
_9-2.8 Te:hnical Specification Modifications included in Reference 1 are a number of proposed modifications to the Dresden 2 Technical Specifications.
Proposed changes which have been addressed in the preceding sections include adoption of the 1.07 Safety Limit Critical Power Ratio (SLMCPR) (Section 2.2.1),
revision of operating limit MCPRs and adoption of a 107% rod block setpoint (Section 2.2.2), revisions in MAPLHGR limits (Section 2.3.1),
and the license restriction pertaining ;o end of cycle coastdown The remaining proposals are evaluated in this section.
(Section 2.7).
Addition of Limiting Total Peaking Factor for 8x8R Fuel 2.8.1 Because of the addition of 8x8R fuel to the Dresden 2 core, a value for the limiting total peaking f actor (LTPF) for 8x8R fuel has been l
included in the Technical Specifications.
The LTPF for a given fuel type is the ratio of the design linear heat generation rate to the The value (2.98) average linear heat generation rate at rated power.
of LTPF for Hx8R fuel proposed by the licensee has been derived from the BrSR fuel characteristics described in Reference 6, and the design
(!3.4 KW/ft).
We conclude that the adoption of the LHGR for d C fum for 8x8R fuel provides an acceptable limitation on LHGR
(
proposed LTPF consistent with tne requirements of Reference 8.
t Modifications to Low and Low-Low Water Level Setpoints l
2.8.2 The licensee has also proposed changes in the reactor low water level and low-low water level setpoints, and has included an explicit refer-ence point, 360" above vessel zero, as the defined top of the active These changes eliminate any potential anbiguity in the required fuel.
limiting setpoints associated with the dif ferent heights of the fuel The proposed types to be present in the Dresden 2 core during Cycle 7.
water level setpoints do not constitute a reduction in safety margin because the LOCA analyses which are based on these setpoints have always been referenced to vessel zero. Thus, during any LOCA the core uncovery time, or reflood time, or PCT would not be af fected by the We, therefore, find the proposed change acceptable.
change.
E
I l 2.8.3 Reduction in Lowest Safety Valve Setpoint The licensee has proposed to lower the setpoint for opening of the lowest of the safety / relief valves from 1125 to 1115 psig. The new setpoint with the appropriate +1% tolerance was used in the over-On the basis that the pressure transient analyses (Reference 1).
criteria for pressurization transients (References 6 and 8) are met (Section 2.4) with the modified setpoint, we conclude that the proposed change is acceptable.
2.8.4 LHGR Limits for 8x8R Fuel Because of the addition of the 160 8x8R fuel elements, a limit on LHGR for this fuel type has been included in the proposed Technical Specification changes. The limit proposed,13.4 KW/f t, has been discussed in Reference 6 and found acceptable in Reference 8.
As explained in Reference 1, a LHGR spiking penalty associated with densification has not been included in the proposed Technical Speci-fications, but has instead been included in the transient analyses We consider this by lowering the safety limit an equivalent amount.
acceptable.
p l
Reduction in the Reactor Power Level Above Which the SLMCPR 2.8.5
{
Becomes F.ftective Under the proposed Technical Specifications, the SLMCPR will be in i
ef fect above 30% rated power instead of 70% rated power as was the i
case previously.
The 301 value is considerably more conservative than the 70% cut-of f which was originally included when minimum critical heat flux ratio (MCHFR) was used as a thermal limit.
We consider the change to be acceptable.
2.8.6 Limits Associated with Control Rod Drop Accidents Two changes in the limiting conditions for operation have been pro-posed to assure tnat if a control rod drop accident should occur the consequences will be within acceptable guidelines.
First, the 280 calorie /gm design basis limit on the peak fuel enthalpy for potential rod drop accidents has been adopted to replace a limit Neither on the maximum reactivity worth of a potential dropped rod.
er
. of these limits is a directly observable quantity. However, the 280 calorie /gm limit is more general than the reactivity worth limit, and the licensee has included the various parametric limits which must be met by the core design in the proposed bases.
Furtherm..te, the bases reference the banked position control rod withdrawal sequences (Reference 13) which indirectly impose the apprcpriate constraints on rod withdrawal through the rod-worth minimizer or administrative con-trols. We, therefore, accept the proposed change.
The second of the proposed limits is the one which requires that ap-propriate control rod withdrawal sequences be followed up to 20% rated power.
The basis for this is that generic analyses have shown that rod drop accidents from power levels above 20% will not exceed the 280 calorie /gm limit (Reference 14).
The 20% cut-off has been accepted generically and we conclude that it is also acceptable for Dresden 2.
3.0 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 environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.
4.0 Conclusions 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 com-liance with the Commission's regulations and the issuance of this amendmt will not be inimical to the common defense and security or to the health and safety of the public.
Date: April 24, 1979
_12 References 1.
" Supplemental Reload Licensing Submittal for Dresden Nuclear Power Station, Unit 2," Reload 4, NE00-24160, November 1978.
2.
Letter, Cordell Reed, Commonwealth Edison (CECO) to NRC, J anuary 15, 1979.
3.
" Loss of Coolant Accident Analysis Report," NEDO 24146, September 1978, Submitted under Docket No. 50-254.
4.
Letter, Cordell Reed, CECO to NRC, March 2,1979.
5.
Letter, Cordell Reed, CECO to NRC, April 6, 1979.
6.
" General Electric Boiling Water Reactor Generic Reload Fuel Application," NEDE-24011-P, May 1977.
7.
" General Electric Boiling Water Reactor Generic Reload Application for 8x8 Fuel," NED0-20360, Rev.1, Supplement 4, April 1, 1976.
8.
Safety Evaluation of the GE Generic Reload Fuel Application (NEDE-24011-P), April 1978.
9.
Status Report on tne Licensing Topical Report " General Electric Boiling Water Generic Reload Application for 8x8 Fuel,"
j NE00-2036L, Revision 1 and Supplement 1 by Division of Technical Review, Office of Nuclear Reactor Regulation, United States Nuclear Regulatory Commission, April 1975.
- 10. Letter, Ronald Engel, GE to Darrell Eisenhut, NRC, Fuel Assembly Loading Error, June 1,1977.
s
- 11. Letter, Ronald Engel, GE to Darrell Eisenhut, NRC, November 30, 1977.
Safety Evaluation of New GE Fuel Loading Error Methods, April 1978.
12.
13.
C. J. Paone, " Banked Position Withdrawal Sequence," NE00-21231, January 1977.
n
. Paone, Stirn, and Wooley, " Rod Drop Accident Analysis for Large 14.
Boiling Water Reactors," NEDO-10527, March 1972.
Amendment No. 50 to Facility Operting License No. DPR-29 for 15.
Unit 1 of Quad Cities Nuclear Power Station, February 23, 1979.
Letter f rom Ronald Engel, GE to Paul Check, NRC, November 1,1978.
16.
Letter f rom R. L. Bolger, CECO to B. C. Rusche, NRC, June 11, 1976.
17.
Letter from R. L. Bolger, CECO to E. G. Case, NRC, June 6,1977.
18.
Letter from Cordell Reed, CECO to Director of NRR, April 12, 1979.
19.
Letter f rom Cordell Reed, CECO to Director of NRR, April 20, 1979.
I 20.