ML20116E250
| ML20116E250 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 04/27/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20116D885 | List:
|
| References | |
| FOIA-96-237 NUDOCS 9608050170 | |
| Download: ML20116E250 (16) | |
Text
jo auoq'o 8
UNITED STATES g
[
g NUCLEAR REGULATORY COMMISSION L
j WASHINGTON, D. C. 20555
\\...+/
ENCLOSURE 1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION j
RELATING TO AMENDMENT NO.
TO FACILITY OPERATING LICENSE NO. DPR-69 BALTIM0RE GAS AND ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT 2 DOCKET NO. 50-318 l
1.0 INTRODUCTION
By letter dated February 7,1989, the Baltimore Gas and Electric Company (BG&E or the licensee) submitted a request for an amendment to its operating license 4
i for Calvert Cliffs Unit No. 2 to allow operation for a ninth cycle at a 100%
rated core power of 2700 MWt. The licensee also submitted proposed modifications to the Technical Specifications (TS) for Cycle 9.
Cycle 9 will l
have a 24 month cycle length as did the previous cycle. The Unit 1 Cycle IC I
design is the reference design for Unit 2 Cycle 9.
t The NRC staff has reviewed the application and the supporting documents and has prepared the following evaluation of the fuel design, nuclear design, l
[
thermal-hydraulic design, transient and accident analysis, and TS changes.
2.0 EVALUATION OF FUEL DESIGN 2.1 Fuel Assembly Description The Cycle 9 core consists of 217 fuel assemblies. Ninety-two fresh (unirradiated) Batch L assembifes manufactured by Combustion Engineering (CE) will replace previously irradiated assemblies. Four of these Batch L assemblies will contain erbia (Er 0 ) as the burnable poison material instead 23 of B 0.
These assemblies will have an initial enrichment of 3.81 weight 4
percent U-235 and are being placed in the Cycle 9 core as lead test essemblies 9608050170 960731 PDR FOIA DINICOL96-237 PDR
.m m
2 to aid in qualification of the use of erbia as an acceptable burnable poison for use in 24-month cycle cores.
2.2 Mechanical Design The mechanical design of the CE Batch L reload fuel is essentially identical to the Batch M fuel approved for use in Cycle 10 of Calvert Cliffs Unit l'with the following exceptions:
(a) The fuel rod plenum spring in the Batch L fuel has been redesigned to maximize the available rod internal void volume. This modification helps reduce high end of cycle (EOC) internal gas pressures.
(b) The overall length of the Batch L B C burnable poison rod has been 4
increased so that the poison rod length is now the same as the fuel rod
-j length. This allows the same type cladding tube to be used for both rod types.
l (c) The size and number of crimp holes in the upper end of each of the five guide tubes of each Batch L assembly have been modified. This design chc 1ge allows the fuel assembly upper end fitting guide tube posts to be reusable if the assembly must be disassembled for fuel rod reconstitution.
(d) The Batch L lower end fitting, flow hole configuration has been modified to a new smaller hole, more debris-resistant design.
In this design, nine small, chamfered holes replace each of the larger holes in the reference cycle design, thus forming a smaller diameter flow path more restrictive to the intrusion of reactor coolant system debris into the fuel assembly.
The staff finds these design changes have been adequately considered in all aspects of the nuclear, mechanical, thermal-hydraulic, and transient safety P
.=.
i 3
analyses for Cycle 9.
In addition, all CE fuel to be loaded for the Cycle 9
)
core was reviewed to ascertain that adequate shoulder gap clearance exists.
Analyses were performed with NRC approved models and the licensee concluded 1
that all shoulder gap and fuel assembly length clearances are adequate for Cycle 9.
2.3 Thermal Design The thermal performance of the CE fuel in Cycle 9 was evaluated using the FATES 3B fuel evaluation model which has been approved by the NRC for BG&E licensing submittals. The licensee analyzed a composite standard fuel rod that enveloped the various fuel batches in Cycle 9.
The analysis modeled the power and burnup levels representative of the peak rod at each burnup interval and bounds the erbia bearing fuel rods. Although the burnup range analyzed for the peak rod was greater than that expected at the end of Cycle 9, approximately one percent of the fuel rods will achieve burnups greater than the 52,000 MWD /T value approved for CE fuel if Cycles 8 and 9 are operated to their maximum burnups.
In order to limit the maximum internal hot gas pressure throughout Cycle 9 to a value that is less than the nominal reactor coolant system (RCS) pressure of 2250 psia, the allowable peak linear heat generation rate (PLHGR) in the peak rod was reduced to 15.2 Kw/ft. The licensee has confirmed that the maximum relative power density of any rod which exceeds 52,000 MWD /T will be at least 30% below the single rod peak in the core and the maximum pressure within any of these rods also will not reach the nominal RCS pressure.
Based on the above and on the fact that evaluations performed by the licensee have shown that the four erbia lead test assemblies are thermally compatible with the other fuel assemblies and meet all the appropriate fuel thermal design criteria required by the staff, the staff concludes that the thermal design of the Cycle 9 fuel is acceptable.
2
d i
4 a
)
_ 3.0 EVALUATION OF NUCLEAR DESIGN
+
3.1 Fue1 Nanagement 1
I The Cycle 9 core consists of 217 fuel assemblies, each having a 14 by 14 fuel rod array. The highest U-235 enrichment occurs in the non-erbia bearing Batch i.
L fuel assemblies which contain an assembly average enrichment of 4.30 weight
= percent U-235. The Calvert Cliffs fuel storage facilities have been approved' for storage of fuel of maximum enrichment of 5.0 weight percent U-235 and,.
j
+
therefore, the fresh Batch L assemblies are acceptable from a fuel storage
. aspect.
i l
The Calvert Cliffs refueling procedures allowed placement of fuel assemblies l
in intennediate positions during core alterations. During an analysis of j
refueling configurations, the licensee discovered that the potential has cxisted for placing several 4.3 weight percent fresh highly reactive fuel assemblies together and losing some of the required'5 percent shutdown margin, f
cr in the extreme, having an inadvertent criticality. Since a significant f
amount of reactivity could be added to a suberitical geometry by placing a j:
single fresh 4.3 weight percent fuel assembly in certain intermediate n p ht i
1 y k gn t or c n r 1 i
assemblies (CEAs)'andassumingthattheminimumrefuelingboronconcentration exists, it was determined that an inadvertent criticality could occur under j
the extreme conditions of grouping a number of such highly reactive assemblies together. Therefore, BG8E issued a 10 CFR 21 written report to the NRC on I
March 15, 1989, concerning the potential loss of shutdown margin during l
refueling.
i 1
As documented in the letter from George C. Creel (BG&E) to NRC dated April 21
~
j 1989, BG&E has revised their refueling procedures to ensure that a fuel assembly j
will not be placed in an intermediate position during core alterations without i
first verifying its potential reactivity. Fuel will only be positioned in intermediate core locations'which will contain fuel of equal or greater s
5 reactivity in the final core configuration.
In order to prevent inadvertent misplacement of fuel assemblies, the revised procedures also require operators to identify a fuel assembly as new or irradiated by its appearance. An irradiated assembly will have an oxide layer and appear black.
In addition, all core locations will be verified prior to insertion and after insertion.
)
These additional procedures should ensure that a new fuel assembly is picked up and placed where a new fuel assembly is supposed to go. The staff finds this to be an acceptable means for assuring that required shutdown margin is i
maintained at all times during refueling and that an inadvertent criticality could not occur.
The Cycle 9 core will use a low-leakage fuel management scheme. With the proposed loading, the Cycle 9 reactivity lifetime for full power operation is expected to be 20,650 MWD /T based on a Cycle 8 length of 18,300 MWD /T. The analyses presented by the licensee will accommodate a Cycle 9 length between 20,400 MWD /T and 21,500 MWD /T based on Cycle 8 lengths between 17,000 MWD /T and 19,000 MWD /T.
3.2 Power Distribution Hot full power (HFP) fuel assembly relative power densities are provided in the reload analysis report for beginning-of-cycle (BOC), middle-of-cycle (MOC),
and end-of-cycle (EOC) unrodded configurations. Radial power distributions at B0C and EOC are also provided for CEA Bank 5, the lead regulating bank, fully inserted. These distributions are characteristic of the high burnup end of the Cycle 8 shutdown window and tend to increase the radial power peaking in the Cycle 9 core. The four Batch LE lead test assemblies (Er 0 ) were calculated p3 to have maximum pin power peaking at least 10% lower than the maximum pin peaking in the core under all expected Cycle 9 operating conditions. The distributions were calculated with approved methods and include the increased power peaking which is characteristic of fuel rods adjacent to water holes.
In addition, the safety and setpoint analyses conservatively include uncertainties and other allowances so that the power peaking values actually used are higher
~
'i 6
than those expected to occur at any time in Cycle 9.
Therefore, the predicted Cycle 9 power distributions are acceptable.
3.3 Moderator Temperature Coefficient (MTC) j
~
The Cycle 9 moderator temperature coefficient (MTC) positive limit varies 1
linearly from +0.3x10~4 delta rho /*F at 100% power to +0.7x10~4 delta rho /*F at 70% power and below. The negative limit is -2.7x10~4 delta rho /*F.
The NRC has previously expressed concern about positive MTC effects on the generic anticipated transients without scram (ATWS) assumptions and BG8E has stated that they will address the generic ATWS implications, if any, in the future.
In the interim, the NRC has approved operation for core designs with allowable positive MTC values provided that the MTC becomes negative at 100% power and equilibrium xenon cenditions. The licensee has agreed to this commitment and has predict <,d a negative MTC at HFP equilibrium conditions ranging from
-0.04x10'# delta rho /*F at B0C to -2.3x10~4 delta rho /*F at E0C for Cycle 9.
3.4 Control Requirements 1
The CEA worths and shutdown margin requirements for Unit 2 Cycle 9 are most limiting at EOC. The licensee's assessed shutdown margin requirements for Cycle 9 are based on the results of the EOC, hot zero power (HZP), steamline break event. After consideration of all reactivity uncertainties and biases, a worst case assessment for Cycle 9 results in a 0.4% delta rho margin in excess of the proposed 5.0% delta rho EOC Technical Specification requirement.
l Therefore, sufficient CEA worth is available to accommodate the reactivity effects of the steamline break event at the worst time during Cycle 9 allowing for the most reactive CEA stuck in the fully withdrawn position. Thus, the control requirements are satisfactorily met.
3.5 Incore Monitoring 4
The incore detector measurement constants to be used in evaluating the reload cycle power distributions will be calculated in the same manner as those for 1
e 7
the reference cycle. However, changes to the on-line incore limiting condition for operation (LCO) monitoring system have been proposed for Cycle 9.
The INCA computer code, which is currently used for power distribution surveillance, would be replaced by the CECOR 3.3 code. Since CECOR 3.3 calculates a full core solution and can be used to obtain 3-dimensional power distribution data as compared to INCA which gives an octant solution, this change should result in more accurate monitoring. CECOR 3.3 would be used on-line and is virtually identical to and gives the same results as the CECOR 2.0 code which is currently utilized off-line. The use of CECOR 3.3 in an on-line network to monitor compliance with the linear heat rate (LHR) and departurefromnucleateboiling(DNB)LC0TechnicalSpecificationswould incorporate the present separate monitoring systems, f.e., Alarm Limit System for LHR and Better Axial Shape Selection System (BASSS) for DNB, into one linked system. BASSS would use the CECOR 3.3 calculated radial peaking factor instead of the presently required Technical Specification Ifmit. CECOR 3.3 would also supply BASSS with a core average axial shape index based on the full core solution which would be used together with the Bank 5 rod position and the unrodded radial. peaking factor to calculate the available DNB cverpower margin and alarms as at present. BASSS would provide the capability T
to monitor the LC0 on total planar adial peaking factor, F and total xy integrated radial peaking factor, F. If the Technical Specification limits r
on these were exceeded during normal operation, BASSS would activate an alarm and would calculate the proper trade-off with maximum allowed power to ensure that the axial power distribution and thermal margin / low pressure (TM/LP) trips remain conservative. The proposed Technical Specifications are worded to support either the full core CECOR 3.3 measured power distribution or the present INCA octant measured power distributions.
The staff finds these proposed changes acceptable and appropriately included in the Cycle 9 safety analyses and proposed Technical Specification modifications.
8 4.0 EVALUATION OF THERMAL-HYDRAULIC DESIGN I
4.1 DNBR Analysis Steady state Departure from Nucleate Boiling Ratio (DNBR) analyses of Cycle 9
-at the rated power level of 2700 MWt have been performed using the approved core thermal-hydraulic codes TORC and CETOP, and the CE-1 critical heat flux correlation. The cycle specific TORC and CETOP models used for designing Cycle 9 account for the small flow hole configuration used in the lower end i
fitting of the Batch L fresh fuel. Engineering hot channel factors and conservatisms are combined statistically with other uncertainty methods using theapprovedExtendedStatisticalCombinationofUncertainties(ESCU) methodology to arrive at an equivalent DNBR limit of 1.15 at a 95/95 probability / confidence level. The Cycle 9 DNBR analyses bound the four Batch LE lead test assemblies without crediting ESCU methods since the maximum single fuel rod peak is at least 10% below the maximum single fuel rod peak in the core. Therefore, the hottest fuel rod in'the core is never located in any of the lead test assemblies at any point throughout Cycle 9.
The DNBR analysis for Cycle 9 is, therefore, acceptable.
F_uel Rod Bowing 4.2 ue The fuel rod bow penalty accounts for the adverse impact on minimum DNBR of random variations in spacing between fuel rods. The methodology for determining rod bow penalties for Calvert Cliffs was based on NRC approved methods. The penalty at 45,000 MWD /T burnup is 0.006 in MDNBR. This penalty is included in the ESCU uncertainty allowance discussed above. For those assemblies with average burnup in excess of 45,000 MWD /T, sufficient margin exists to offset rod bow penalties. The staff, therefore, concludes that the analysis of fuel rod bow penalty is acceptable.
l 9
5.0 EVALUATION OF SAFETY ANALYSES 5.1 Non-LOCA Events All of the key input parameters for the transient and accident analyses are bounded by (or are conservative with respect to) those of the reference cycle (Unit 1 Cycle 10) with the exception of the following:
1.
Cycle 9 Batch L fuel utilizes a small flow hole debris resistant design.
2.
The maximum auxiliary feedwater (AFW) flow assumed in'the safety analyses was increased from 1300 gpm for the reference cycle to 1550 gpm for Cycle j
9 in anticipation of larger measured AFW flow.
3.
The maximum assumed number of plugged U-tubes per steam generator was increased to 500 for all but the small break LOCA Cycle 9 analyses.
The licensee has determined by reanalysis or reevaluation that the results of all events affected by these input parameter changes remain bounded by those of the reference cycle. The staff, therefore, concludes that the non-LOCA transient and accident events for Cycle 9 are bounded by the reference analyses and the results of the non-LOCA safety analyses are acceptable.
5.2 LOCA Events The large break LOCA has been reevaluated for Cycle 9 to demonstrate that a peak linear heat generation rate (PLHGR) of 15.5 Kw/ft complies with the acceptance criteria of 10 CFR 50.46 for emergency core cooling systems (ECCS) for light water reactors. Blowdown hydraulic calculations have shown that the slight increase in pressure drop due to the Batch L debris-resistant fuel design is insignificant.
In addition, a reduction of 260 gpm in low pressure safety injection (LPSI) flow was assumed in order to establish an increased margin between analysis flow values and actual measured flow values. The licensee confirmed that there remains adequate safety injection flow to
~.. - - -. - - -._ _ _ _
~
10 t
maintain a full downcomer with the reduced flow and, therefore, the reduction in LPSI flow does not affect the results. Since all other fuel related parameters for Cycle 9 remain bounded by those of the reference cycle, the staff concludes that the large break LOCA is bounded by the reference cycle analysis. Therefore, operation of Unit 2 Cycle 9 at a PLHGR of 15.5 Kw/ft and a power level of 2754 MWt (102% of 2700 MWt) is in compliance with the.10 CFR 50.46 acceptance criteria. The allowable PLHGR is being decreased to 15.2 Kw/ft for Cycle 9 to accommodate the more limiting fuel performance data associated with this second 24 month cycle in Unit 2.
4 The licensee reports that analyses have confirmed that small break 1oss of coolant accident (SBLOCA) results for Calvert Cliffs Unit 1 Cycle 8, which is the reference cycle for SBLOCA, bound the Unit 2 Cycle 9 results. Unlike the large break LOCA analysis which considered 500 plugged tubes per steam generator, the SBLOCA considered orly 150 plugged tubes per steam' generator.
This effect as well as the effect of the reduction in LPSI flow and the incorporation of the debris-resistant fuel design were evaluated by the licensee and were found to have no significant impact on the SBLOCA.
Therefore, acceptable SBLOCA ECCS performance is also demonstrated at a PLHGR of 15.5 Kw/ft and a reactor power level of 2754 MWt. The allowable PLHGR is being dec eased to 15.2 Kw/ft for Cycle 9 to accommodate the more limiting fuel performance data associated with this second 24 month cycle in Unit 2.
6.0 TECHNICAL SPECIFICATIONS The following paragraphs summarize the proposed changes to the Technical Specifications requested to support operation of Unit 2 Cycle 9.
Some of these changes have already been implemented for Unit 1 Cycle 10. Others involve specific Unit 2 Cycle 9 changes such as the use of the new on-line incore LCO monitoring system discussed in Section 3.5 of this SER.
l Figure 2.2-2 is modified to accommodate the implementation of the ESCU 1.
methodology used originally in the thermal-hydraulic analysis of the reference cycle (Unit 1 Cycle 10) and now in the Unit 2 Cycle 9 reload
)
11
' design. Use of the ESCO methodology requires changes to the coefficients of the PVAR equation presented in Figure 2.2-2, ensuring agreement of the Thermal Margin / Low Pressure trip setpoint with the Unit 2 Cycle 9 DNBR analysis. The changes are, therefore, acceptable.
2.
Figure 2.2-3 is modified for the same reason as Figure 2.2-2 and the change is acceptable for the same reason.
3.
The text of B 2.2.1 is modified to identify in the Technical Specification bases the DNBR Specified Acceptable Fuel Design Limit (SAFDL)valueof1.15. This change is resultant from use of the ESCU methodology and is identical to that approved for the reference cycle.
Therefore, it is acceptable.
4 The text of 3.1.1.1/4.1.1.1.1 (with the addition of new Figure 3.1-Ib) is modified to establish a shutdown margin limit for Unit 2 Cycle 9 as a function of time in cycle. The actual required shutdown margin varies throughout the cycle due to changes in the core such as fuel depletion, boron concentration, and moderator temperature. The Technical Specification references a'new Figure 3.1-lb, which_provides a shutdown margin limit line varying from 3.5% delta rho required shutdown margin at BOC to 5.0% delta rho shutdown margin at EOC. The Unit 2 Cycle 9 safety analyses consider the Ifmits of Figure 3.1-Ib and are bounded by the reference cycle. This change is identical to the shutdown margin change approved for the reference cycle and is, therefore, acceptable.
5.
Figure 3.1-2 is modified to allow greater insertion of Group 5 CEA's between 90% and 100% power. Specifically, Group 5 insertion is changed from 25% at 100% power to 35% insertion between 90% and 100% power. This change allows an additional 13.5 inches of CEA insertion of the lead The regulating group for enhanced control of axial power oscillations.
Unit 2 Cycle 9 reload safety analyses assume the proposed limits of Figure 3.1-2.
This change is identical to the Figure 3.1-2 change approved in the reference cycle and is acceptable.
=
1 12 I
6.
The text of B 3/4.1.1.1 and B 3/4.1.1.2 is modified to establish an explanation of the basis for the proposed shutdown margin change of 3.1.1.1/4.1.1.1.1, discussed above. The changes are, therefore, f
acceptable.
l 7.
Figure 2.2-1 is modified on the negative Axial Shape Index (ASI) side to accommodate the increased core average linear heat generation rate (CALHGR) of Unit 2 Cycle 9 which is the second 24-month cycle in Unit 2.
i The CALHGR is increased for Cycle 9 because of the increased number of B C shims.
The modification was considered in the Unit 2 Cycle 9 reload 4
l safety analyses as well as in the setpoint analysis and acceptable i
results were obtained. The change is, therefore, acceptable.
8.
The text of 3.1.3.1 and Figure 3.1-3 is modified to incorporate an increaseinmaximumallowedFffrom1.65to1.70.
TheFfisincreased j
.to accommodate the increased neutron flux peaking associated with this second 24-month cycle for Unit 2.
The setpoint analysis performed in the support of Unit 2 Cycle 9 considers this proposed change in Ff. Also, this Technical Specificaticn reflects the use of the CECOR 3.3/BASSS network whfch is considered in the Unit 2 Cycle 9 setpoint analysis.
Therefore, the changes are acceptable.
l 9.
Figure 3.2-1 is modified to indicate the reduced allowable peak linear heat rate (APLHR), for Unit ? Cycle 9, from 15.5 Kw/ft to 15.2 Kw/ft.
The APLHR is reduced to maintain the maximum Cycle 9 internal pin pressure below reactor coolant system pressure of 2250 psia. This change in APLHR is considered by the Unit 2 Cycle 9 fuel performance analysis.
Also, the LOCA analyses have shown that a PLHR as high as 15.5 Kw/ft complies with the acceptance criteria. Therefore, this change is acceptable.
- 10. Figure 3.2-3b is modified to indicate a reduction in its acceptable value T
region due to a reduction in the 2001 power F value of 1.54 to 1.50.
xy The reduction in the limits of this figure result from the need to t
\\
13 acconnodate the increased core average linear heat generation rate of the second Unit 2 24-month reload core design with its increased number of B C shims. This is acceptable since the setpoint analysis takes credit 4
for this modification in demonstrating acceptable results for Cycle 9.
11.
Implementation of CECOR 3.3/BASSS as the on-line incore LC0 monitoring system requires changes in Technical Specifications 3.2.2.1, 4.2.2.1.2, 4.2.2.1.3, 4.2.2.1.4, 4.2.2.2.2, 4.2.2.2.3, 4.2.2.2.4, 4.2.3.2, 4.2.3.3, 4.2.3.4, 4.2.5.3 and B 3/4.1.3 to ensure they adequately consider the CECOR 3.3/BASSS system. This implementation was shown to be acceptable i
in Section 3.5 of this SER.
Figure 3.2.3ismodifiedtoincreasetheFffrom1.65to1.70to 12.
accommodate the increased nuclear flux peaking associated v;ith this second 24-month cycle for Unit 2 and implementation of the CECOR 3.3/BASSS on-line incore monitoring system. The Unit 2 Cycle 9 setpoint analyses supports this change in Ff and, therefore, it is acceptable.
T
- 13. Figure 3.2-3c is modified to accommodate the increased F limit for the 7
j same reason as the proposed change to Figure 3.2.3.
- 14. The text of 3.2.5 and Table 3.2-1 is modified by changing " core power" terminology to " thermal power" to maintain consistency with other section Technical Specifications. This is acceptable.
- 15. The text of B 3/4.2.5 is modified to identify in the Technical Specification Bases the DNBR Specified Acceptable Fuel Design Limit (SAFDL) value of 1.15. This change is resultant from use of the ESCU methodology and is acceptable.
- 16. The text of B 3/4.7.1.2 is modified by increasing the maximum allowed Auxiliary Feedwater flow from 1300 gpm to 1550 gpm. An evaluation of increasing this flow was performed by the licensee and it was determined that the results on the safety analyses for Unit 2 Cycle 9 are bounded by
t 14 previously reported and approved analyses. The change is, therefore, acceptable.
- 17. The text of 5.3.1 is nadified to indicate an increase in the maximum enrichment for a reload core from 4.1 w/o to 4.35 w/o U-235. This change is proposed to allow the higher enriched Unit 2 Cycle 9 second 24amonth cycle reload core. All aspects of the Cycle 9 reload core design consider the proposed higher enrichment. The Unit I and 2 fuel storage facilities have been approved for storage of fuel of maximum enrichment of 5.0 w/o U-235. Therefore, this change is acceptable.
7.0 STARTUP TEST PROGRAM The startup testing program proposed for Unit 2 Cycle 9 is similar to that used in the reference cycle. However, a change is proposed in the manner in which core synnetry is verified by using incore power distribution monitoring with only minimal CEA synnetry testing. This monitoring would be done using measured power distributions generated by the new CECOR 3.3 system which does the solution in full core as contrasted to the previous INCA system whose solution is for an eighth core. The staff has approved this new monitoring system in Section 3.5 of this SER. The licensee has also performed an
' analysis which <*emonstrates that the incore monitoring in full core, in conjunction with minimal CEA symmetry testing, will detect any significant fuel assembly misloadings. Therefore, the staff finds this proposed change in core symmetry confirmation to be acceptable.
8.0 SUPMARY The staff has reviewed the fuel system design, nuclear design, thermal-hydraulic design, startup test program, and the transient and accident analysis information presented in the Calvert Cliffs Unit 2 Cycle 9 reload submittals.
In addition, the modifications made to the refueling procedures as a result of the licensee's 10 CFR 21 notification concerning potential loss of shutdown margin were also reviewed.
}
. l
?
15 1
l Based on this review, which is described above, the staff concludes that the proposed Cycle 9 reload and associated modified Technical Specifications are acceptable. This conclusion is further based on the following:
(1) previously reviewed and approved methods were used in the analyses; (2) the results of the safety analyses show that all safety criteria are met; and (3) the proposed Technical Specifications are consistent with the reload safety analyses.
l l
4
.~
ii i
ENCLOSURE 2 SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE FACILITY NAME Calvert Cliffs Unit 2 l
4
SUMMARY
OF REVIEW The Reactor Systems Branch, DEST, reviewed the request by Baltimore Gas and Electric company to reload and operate Calvert Cliffs Unit 2 for Cycle 9.
Based on our review of the safety analyses and proposed Technical Specification changes, we find the request acceptable.
i I
NARRATIVE DISCUSSION OF LICENSEE PERFORMANCE - FUNCTIONAL AREA Review of the submittal indicated that the licensee adequately addressed the technical aspects of the issues. All of the proposed Technical Specification changes were appropriately incorporated in and supported by the Cycle 9 safety analyses. The reason and justification for each change was adequately discussed in the reload amendment.
In addition, the licensee independently discovered the potential for losing some of the required shutdown margin during refueling operations and filed a 10 CFR Part 21 notification to the NRC. They presented timely and acceptable corrective actions and were responsive to the staff's questions and concerns.
AUTHOR:
L. Kopp DATE:
4/4/89