ML20116E054
| ML20116E054 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 04/12/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20116D885 | List:
|
| References | |
| FOIA-96-237 NUDOCS 9608050054 | |
| Download: ML20116E054 (14) | |
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UNITED STATES 1
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NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D. C. 20555
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ENCLOSURE 1 SAFETY EVALUATION BY THF 0FFICE OF NUCLEAR REACTOR REGULATION RELATING TO CYCLE 10 RELOAD BALTIPORE GAS AND ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT UNIT 1 DOCKET NO. 50-317
1.0 INTRODUCTION
By letter dated February 12, 1988, Baltimore Gas and Electric Company (BG&E),
the licensee, submitted a request for an amendment to its operating license for Calvert Cliffs Unit No. I to allow operation for a tenth cycle at 1001 rated core power of 2700 MWt (Ref. 1). The licensee also submitted proposed modifications to the Technical Specifications for Cycle 10. Cycle 10 will l
have a 24 month cycle length as compared to 18 months for tLe previous cycle.
The NRC staff has reviewed the application and the supporting documents (Refs.
2 & 3) and has prepared the following evaluation of the fuel design, nuclear l
design, thermal-hydraulic design, and Technical Specification changes.
2.0 EVALUATION OF FUEL DESIGN 2.1 Fuel Assembly Description The Cycle 10 core consists of 217 fuel assemblies. Ninety-six fresh (unirradiated) Batch M assemblies will replace previously irradiated assemblies. Of these 96 fresh assemblies, 9? will be manufactured by CombustionEngineering(CE)andfourbyAdvancedNuclearFuels(ANF)
Corporation, and are placed in the Cycle 10 core as an aid in qualifying AFF fuel for 24 month cycle operation. The 92 fresh CE assemblies will consist of 16 unshimed Batch M assemblies and 76 Batch M* assemblies each containing 12 i
B C rods for neutronic shiming and having an initial assembly average 4
9608050054 960731 g,, phic 6-237 PDR
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enrichment of 4.08 weight percent (w/o) U-235. The four ANF Batch MX demonstration assemblies contain ]? fuel bearing Gd 0 rods for shimming and 23 have an initial assembly average enrichment of 3.85 w/o U-235.
2.2 Mechanical Desion The mechanical design of the CE Batch M reload fuel is identical to the Batch K fuel previously inserted in Calvert Cliffs Unit 1.
All CE fuel to be loaded for the Cycle 10 core was reviewed to ascertain that adequate shoulder gap clearance exists. Analyses were performed with approved models and the licensee concluded that all shoulder gap and fuel assembly length clearances are adequate for Cycle 10. The replacement control element assembly (CEA) to be used in the center location of the core will have the same reconstituted features as the replacement CEA installed in the reference cycle.
The mechanical design features of the ANF lead fuel assemblies are described i
in Reference 3.
Most of the assembly and core interface dimensions are identical to the CE fuel assemblies. Differences in the upper and lower end fitting height and overall assembly height should not affect the performance of either 'uel assembly. ' Experience with similar ANF fuel designs coresiding ad.facent to CE reload fuel in the Maine Yankee, Fort Calhoun, and St. Lucie Unit I cores have caused no unexpected problems or operational difficulties.
Therefore, the staff finds the ANF lead assemblies to be mechanically compatible with the co-resident CE fuel during Cycle 10.
2.3 Themal Design The thermal performance of the CE fuel in Cycle 10 was evaluated using the FATES 3B fuel evaluation model (Ref. 4). The staff issued an SER (Ref. 5) approving the use of FATES 3B for BG&E licensing submittals. The licensee analyzed a composite, standard fuel pin that enveloped the various CE fuel batches in Cycle 10. The analysis modeled the power and burnup levels representative of the peak pin at each burnup interval. Although the burnup range analyzed for the peak pin was greater than that expected at the end of
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Cycle 10 approximately 0.3% of the fuel pins will achieve burnups greater than the 57,000 MWD /T value approved for CE fuel (Ref. 6) if Cycles 9 and 10
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are operated to their maximum burnups.
In response to the staff's request, the licensee confirmed that these few high burnup pins will be in low power regions of the Cycle 10 core and the maximum pressure within these pins will i
not reach the nominal reactor coolant system pressure of 2250 psia (Ref. 7).
l Evaluations have been performed to show that the four ANF lead assemblies are thermally compatible with the existing CE fuel assemblies and meet the l
appropriate fuel thermal design criteria required by the staff (Ref. 3).
Based on its review of the infomation discussed above, the staff concludes that the evaluation of the themal design of the CE and ANF fuel for Cycle 10 is acceptable.
3.0 EVAll!ATION OF NUCLEAR DESIGN 3.1 Fuel Management The Cycle 10 core consists of 217 fuel assemblies, each having a 14 by 14 fuel rod array. A general description of the core loading is given in Section P.1 of this SER. The highest ll-235 enrichment occurs in the CE Batch M fuel assemblies which contain an assembly average enrichment of 4.08 w/o U-235.
The Calvert Cliffs fuel storage facilities have been approved for storage of fuel of maximum enrichment of 4.10 w/0 if-235 and, therefore, the fresh Batch M assemblies are acceptable from a fuel storage aspect.
The Cycle 10 core will use a low-leakage fuel management scheme. With the proposed loading, the Cycle 10 reactivity lifetime for full power operation is expected to be 21,400 MWD /T based on a Cycle 9 length of 11,800 MWD /T. The analyses presented by the licensee will accomodate a Cycle 10 length between 20,600 MWD /T and 21,800 MWD /T based on Cycle 9 lengths between 9,800 PVD/T and 11,800 MWD /T.
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3.2 Power Distribution Hot full-power (HFP) fuel assembly relative power densities are given in the 1
reload analysis report for beginning-of-cycle (BOC), middle-of-cycle (MOC),
and'end-of-cycle (EOC)unroddedconfigurations.
Radial power distributions at BOC and EOC are also given for control element assembly (CEA) Bank 5, the lead regulating bank, fully inserted. These distributions are characteristic of the high burnup end of the Cycle 9 shutdown window and tend to increase the radial power peaking in the Cycle 10 core. The four ANF lead test assemblies were calculated to have maximum pin power peaking at least 10% lower than the maximum pin peaking in the core under all expected Cycle 10 operating conditions. The distributions were calculated with approved methods and include the increased power peakino 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 pom r peaking values actually used are higher than those expected to occur at any time in Cycle 10. Therefore, the predicted Cycle 10 power distributions are acceptable.
3.3 Reactivity Coefficients In order to accomodate 24 month cycles, the moderator temperature coefficient (MTC) limit above 70% power is raised from +0.2x10~4 delta rho /* F to a value which varies linearly from +0.3x10-4 delta rho /* F at 100% power to
+0.7x10~4 delta rho /* F at 70% power. The staff has previously expressed concern about the positive MTC effect on the generic anticipated transients without scram (ATWS) assumptions and BG&E has stated that they will address the generic ATWS implications, if any, in the future.
In the interim, the j
staff has approved operation for core designs with allowable positive MTC values provided that the MTC becomes negative at 100% power and equilibrium
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xenon conditions. The licensee has predicted a negative MTC at hot full power, equilibrium xenon conditions of -0.2x10~4 delta rho /* F for Cycle 10 and has committed to a full power negative value at equilibrium xenon conditions (Ref. 7).
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The Doppler coefficient for Cycle 10 is a best estimate value expected to be i
accurate to within 15%. These reactivity coefficient values are bounded by the values used in the safety analyses for the reference cycle (Calvert Cliffs linit 2 Cycle 8). The staff, therefore, finds the values of the MTCs and Doppler coefficients to be acceptable.
3.4 Control Requirements The CEA worths and shutdown marcin requirements at the most limiting time for the Cycle 10 nuclear design, that is, for the E00, are presented in Reference 7.
These values are based on an E00, hot zero power (HZP), steamline break accident. At EOC 10, tie reactivity worth with all CEAs inserted is 9.0%
delta rho. An allowance of 1.1% delta rho is made for the stuck CEA which yields the worst results for the EOC HZP steamline break accident. An allowance of 2.0% delta rho is made for CEA insertion in accordance with the
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powerdependentinsertionlimit(PDIL). The' calculated scram worth is the
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total CEA worth less the worth of the stuck CEA and less the worth of CEA insertion to the PDIL and is 5.9% delta rho. Deducting 0.8% delta rho for physics uncertainty and bias yields a net available scram worth of 5.1% delta rho. Since the Technical Specification FOC shutdown margin at zero power is 5.0% delta rho, a margin of 0.1% delta rho exists in excess of the Technical Specification shutdown margin. Therefore, sufficient CEA worth is available to accomodate the reactivity effects of the steam line break event at the worst time in core life allowing for the most reactive CEA stuck in the full withdrawn position. The staff concludes that the licensee's assessment of reactivity control is suitably conservative and that adequate negative reactivity worth has been provided by the control system to assure shutdown capability assuming a stuck CEA that results in the worst reactivity condition for an EOC, HZP steamline break accident.
3.5 Safety Related Data l
l Other safety related data such as limiting parameters of dropped CEA reactivity worth and the maximum reactivity worth and planar power peaks
6 associated with an e,iected CEA for Cycle 10 are identical to the values used in the referenc3 cycle and are, therefore, acceptable.
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4.0 EVALUATION OF THERMAL-HYDRAULIC DESIGN 4.1 DNBR Analysis Steady state themal-hydraulic analysis of CE fuel for Cycle 10 is perfomed usina the approved core thennal-hydraulic code TORC and the CE-1 critical heat flux correlation (Ref. 8). The core and hot channel are modeled with the approved method described in CENPD-206-P-A (Ref. 9). The design thennal margin analysis is performed using the fast running variation of the TORC code, CETOP-D (Ref. 10), which has been approved for Calvert Cliffs with the
. appropriate hot assembly inlet flow starvation factors to assure its conservatism with respect to TORC. The enoineering het channel factors for heat flux, heat input, rod pitch and cladding diameter are combined
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statistically with other uncertainty factors using the approved extended i
statistical combination of uncertainties (ESCU) method described in CEN-348(B)-P (Ref. M) to arrive at an eouivalent DNBR limit of 1.15 at a 95/95 probability / confidence level.
DNBR analyses were also perfomed to assess the perfomance of the ANF lead assemblies (Ref. 3) using the XCOBRA-III code (Ref.12) and the ANF approved thermal-hydraulic methodology for mixed fuel cores (Ref.13). The XNB DNR correlation (Ref.14) has been shown to be applicable to co-resident CE and ANF fuel (Refs. 14 & 15)'and the staff concludes that it is acceptable to apply it to the mixed Cycle 10 core containing the four ANF lead fuel assemblies. The results indicate that the ANF lead assemblies exhibit higher MDNBRs than the hot CE assembly due to the 5% lower assembly power at which the ANF lead assemblies were simulated. Since the insertion of the ANF lead assemblies does not significantly affect the MDNBR of the hot CE assembly, which establishes the core MDNBR, the staff concludes that the core PDNRR is essentially unchanged by insertion of the four ANF lead assemblies and thus i
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7 the design criterion on DNBR is satisfied by the mixed core containing ANF lead assemblies.
4.2 Fuel Rod Bowing The fuel rod bow penalty accounts for the adverse impact on MDNBR of random variations in spacing between fuel rods. The methodology for detemining rod bow penalties for Calvert Cliffs was based on the NRC approved methods presented in the CE topical report on fuel and poison rod bowing (Ref.16).
The penalty at 45,000 MWD /T burnup is 0.006 in MDNBP. 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.
5.0 EVALUATION OF SAFETY ANALYSES 5.1 Non-LOCA Events j
For the non-LOCA safety analyses, the licensee has determined that the key j
input parameters for the transient and accident analyses lie within the bounds of those of the reference cycle (Unit ? Cycle 8). As noted in Section 6.0, j
the shutdown maroin Technical Specification is being changed from a singular value to a variable ranging from 3.5% delta rho at BOC to 5.0% delta rho at E00. The EOC shutdown margin requirement is determined by the steam line rupture event and a reevaluation of this event at EOC 10 with the revised shutdown margin has indicated that it is less limiting than the reference analysis. The staff, therefore, concludes that the non-LOCA transient and accident events for Cycle 10 are bounded by the reference analyses.
E.2 LOCA Events The large break loss of coolant accident (LOCA) has been reanalyzed for Cycle 10 to demonstrate that a peak linear heat generation rate IPLHGR) of 15.5 kw/ft complies with the acceptance criteria of 10 CFR 50.46 for emergency core
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8 cooling systems (ECCS) for light water reactors. The Cycle 10 analysis, as the reference cycle analysis, was performed with the 1985 CE evaluation model which was approved in Reference 17. The Cycle 10 analysis showed that the
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double ended guillotine pipe break at the pump discharge with a discharge coefficient of 0.6 (0.6 DEG/PD) gave the highest peak clad temperature. Table 8.1-1 of the reload report provides the input parameters for the fuel for Cycle 10 and the reference cycle. Table 8.1-2 presents the results of the analysis for the limiting break for Cycle 10 and the reference cycle. The results for the limiting Cycle 10 break show that (1) the peak clad temperature is 1983* F which is well below the acceptance criterion of 2200* F and (2) the maximum local and core wide oxidation values are 4.14% and less than 0.51%, respectively, and these are well below the acceptance criteria of i
17% and 1%, respectively. The analysis considered up to 500 plugged tubes per steam generator and a 40 second safety injection pump response time. Since the Cycle 10 large break LOCA ECCS analysis has shown that both the peak clad temperature and clad oxidation meet the acceptance criteria of 10 CFR 50.46, the operation of Cycle 10 at an allowable PLPGR of 15.5 kw/ft is acceptable.
The licensee reports that analyses have confirmed that small break loss of coolant accident (SBLOCA).results for Calvert Cliffs Unit 1 Cycle 8, which is the reference cycle for SBLOCA, bound the Calvert Cliffs Unit 1 Cycle 10 results. Unlike the large break LOCA analysis, the SBLOCA considered only 100 plugged tubes per steam generator. The increased safety injection pump response time considered in the large break analysis also was not evaluated for the SBLOCA analysis. Since the acceptance criteria for the SBLOCA are met, the operation of Cycle 10 at an allowable PLHGR of 15.5 kw/ft, with up to 100 plugged tubes per steam generator, is acceptable.
6.0 TECHNICAL SPECIFICATIONS As indicated in the staff's evaluation of the nuclear design, provided in Section 3, the operating characteristics of Cycle 10 were calculated with approved methods. The proposed Technical Specifications are the results of
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the cycle specific analyses for, among other things, power peaking and control rod worths. The analyses perforined include the implementation of a low-leakage fuel shuffle pattern with fuel enrichments and burnable. poison loadings and distributions chosen to provide a cycle length of 24 months.
Some of the reauested Technical Specification changes involve changes to both Unit I and Unit 2 Technical Specifications.
Each proposed change is discussed below.
1.
Figure 2.2-2 Thermal Margin / Low Pressure Trip Setpoint-Part 1 Figure 2.2-2 is modified due to a revision in the curve fit for the TM/LP trip setpoint to accommodae.e the implementation of the extended statistical combination of uncertainties methodology. The setpoint analysis uses this methodology and.the licensee has determined that acceptat'.e results are obtained for Cycle 10. The changes to Figure 2.2-2 are, therefore, acceptable.
2.
Figure 2.2-3 Thermal Margin / Low Pressure Trip Setpoint-Part ?
Figure 2.2-3 is modified for the same reason as Figure 2.2-2 and the change is acceptable for the same reason, j
3.
Bases 2.1.1 and 2.2.1 The text is modified to replace a specific minimum DNBR value with the phrase DNB SAFDL. The use of a phrase in place of a specific minimum DNBR value was recommended in the extended SCU methodology (Ref.11) and approved by the staff (Ref. 18). The change is, therefore, acceptable.
4.
Technical Specification 3.1.1.1 Shutdown Margin Two modifications are proposed for this specification. First, the shutdown margin is changed from a constant value to text which refers to a new Figure 3.1-1b which presents shutdown margin as a function of time
4 10 in cycle. Since the required shutdown margin varies throughout the cycle due to fuel depletion, boron concentration and moderator temperature and i
this variation with cycle time has been incorporated in all the.
appropriate safety analyses for Cycle 10, this change is acceptable, l
The shutdown margin at EOC is increased from 3.5% delta k/k to 5.0% delta k/k. The_ analysis of the Cycle 10 steam line rupture analysis, which is limiting at hot zero power EOC conditions, supports this change and it is, therefore, acceptable.
.5.
Technical Specification 3.1.1.4 Moderator Temperature Coefficient The MTC limit above 70% power is being raised from +0.2x10~4 delta rho /* F to a value which varies linearly from +0.3x10~4 delta rho /* F at 100% power to +0.7x10-4 delta rho /* F at 70% power. This change is being implemented to accommodate 24 month cycles and to facilitate initial i
reactor startup at the beginning of the cycle. The licensee has committed to a negative MTC at hot full power, ecuilibrium xenon conditions. As mentioned in Section 3.3, this value has been predicted to be -0.2x10~# delta rho /* F.
The feedline break analysis which supports this char.ge is applicable to Cycle 10 and, therefore, the proposed change is acceptable.
6.
Figure 3.1-2 CEA Group Insertion Limits i
The transient insertion limit between 90% and 100% power is being increased from an allowed insertion limit which varies linearly from 35%
j for Bank 5 at 907 power to 25% at 100% power, to a constant value of 35%.
This change, which is being made to enhance the ability to control axial i
oscillations near EOC, has been incorporated into all of the Cycle 10 1
physics, safety and setpoint analyses and is, therefore, acceptable.
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11 7.
Figure ?.?-1 Axial Power Distribution Trip LSSS Figure ?.2-1 is modified to increase the positive and necative axial shape index (ASI) regions below 70% power. The setpoint analysis uses the modified results given by Figure 2.2-1 and the licensee has determined that acceptable results are obtained for Unit 1 Cycle 10 and Unit 2 Cycle 8.
The changes to Figure 2.2-1 are, therefore, acceptable for both units.
8.
Fioure 3.2-2 linear Heat Rate Axial Flux Offset Control Limits Figure 3.2-4 DNB Axial Flux Offset Control Limits These Figures are modified to increase the negative ASI limits below 50%
power. The licensee has evaluated the effect of the proposed new limits on the Unit 1 Cycle 10 and Unit 2 Cycle 8 transient analyses, margin to fuel centerline melt limits, margin to DNB limits, margin to LOCA PLHGR limit, core power versus planar radial peaking factor LCO TM/LP LSSS, and core power versus integrated radial-peaking factor LCO and has determined that acceptable results are obtained. The changes are, therefore, acceptable for Unit 1 Cycle 10 and Unit 2 Cycle 8.
7.0 CONCLUSION
S The-staff has reviewed the fuel system design, nuclear design, thermal-hydraulic design, and the transient and accident analysis information presented in the Calvert Cliffs Unit 1 Cycle 10 reload submittals. Pased on this review, which is described above, the staff concludes that the proposed Cycle 10 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.
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8.0 REFERENCES
1 1.
LetterfromJ.A.Tiernan(BG8E)toUSNPC,B-88-011."CalvertCliffs Nuclear Power Plant Unit Nos. I and 2; Docket Nos. 50-317 and 50-318 Pequest for Amendment Unit 1 Tenth Cycle License Application; Unit Two 4
Axial. Shape Index Region Enlargement," February 12, 1988 2.
Attachment to B-88-011 Calvert Cliffs Unit Cycle 10 License Submittal.
j 3.
Appendix to B-88-011 Calvert Cliffs Unit 1 Cycle 10 License Submittal, ANF-88-019.
4 4.
" Improvements to Fuel Evaluation Podel," CEN-161(B)-P, Supplement 1-P (proprietary), April 1986.
5.
Letter from Scott A. McNeil (NRC) to J. A. Tiernan (BG8E), dated February 4, 1987.
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6.
Letter from E. J. Butcher (NRC) to A. E. Lundvall, Jr. (BG&E), " Safety l
Evaluation for Topical Report CENPD-369-P, Revision 1-P " October 10, 1985.
7.
Letter from J. A. Tiernam (BG&E) to NRC, " Unit 1 Cycie 10 Response to Request for Additional Infonnation," March 25, 1988.
8.
" Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids. Part 1. Unifonn Axial Power Distribution," CENPD-162-P-A, l
April 1975, i
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" TORC Code Verification and Simplified Modeling Methods," CENPD-206-P-A, 9.
June 1981.
- 10. "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units-1 and 2," CEN-191(B)-P, December 1981.
i 13 l
11.
" Extended Statistical Combination of Uncertainties," CEN-348(B)-P, January 1987._
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12.
"XCOBRA-IIIC: A Computer Code to Determine the Distribution of Coolant
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During Steady-State and Transient Core Operation," XN-NF-75-21(P)(A),
j January 1986.
1 13.
" Application of Exxon Nuclear Company PWR Thermal Margin Methodology to
)
Mixed Core Configurations," XN-NF-82-21(P)(A), Rev. 1, September 1983.
Rev. 1 September 1983.
15.
" Justification of XNB Departure from Nucleate Boiling Correlation for St.
Lucie Unit 1," XN-NF-83-08(P), February 1983.
- 16. " Fuel and Poison Rod Bowing," CENPD-225-P-A, June 1983.
- 17. Letter from D. M. Crutchfield (NRC) to A. E. Scherer (CE), " Safety Evaluation of Combustion Engineering ECCS Large Break Evaluation Model and Acceptance for Referencing of Related Licensing Topical Reports,"
July 31, 1986.
- 18. Letter from S. A. McNeil (NRC) to J. A. Tiernan (BG&E), ' Safety Evaluation of Topical Report CEN-348(B)-P, " Extended Statistical Combination of Uncertainties,"' October 21, 1987.
s ENCLOSURE 2 CALVERT CLIFFS UNIT 1 SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMAPCE Functional Areas 1.
Managerent Involvement in Assuring Ouality.
Technical review of the submittal indicates that the management reviews are adequate.
Rating: Category 2 2.
Approach to Resolution of Technical Issues from a Safety Standpoint.
The licensee showed a general understanding of the technical issue and used acceptable approaches in most cases.
Rating: Category 2 3.
Responsiveness to NRC Initiatives.
The licensee responded favorably to NPC initiatives.
Rating: Category ?
4.
Enforcement History.
N/A
- 5.
Operational and Construction Events.
N/A 6.
Staffino (includi.g Management).
N/A 7.
Training and Qualification Effectiveness.
N/A
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