ML20154J798
| ML20154J798 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 05/16/1988 |
| From: | Office of Nuclear Reactor Regulation |
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| Shared Package | |
| ML20154J789 | List: |
| References | |
| NUDOCS 8805270136 | |
| Download: ML20154J798 (10) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION o
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WASHINGTON, D C. 20555
,o SAFETY EVALUATION BY THE OcFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENCPENT t10. 130 TO FACILITY OPFRATit!G LICENSE N0. OPR-53 BALTIMORE GAS AFD ELECTPTC COMPANY CAI."EP' CLIFFS NUCLEAR POWER PLANT, Ut!!T 1 POCVET NO. 50-317
1.0 INTRODUCTION
Py letter dated February if,19E8, as supplemented on March 01, ?' arch 25 (2 letters! anc April 14, 1988, the Baltimore Gas and Electric Company (RGAE or the licenseel submitted a request for an amendment to its cperatino license for Calvert Cliffs Unit No. I to allow operation for a tenth cycle at a 100% rated core pcver of 2700 MWt (Ref. 1).
The licensee also submitted proposed modifi-cations to the Technical Specifications (TS) for Cycle 10. Cycle 10 will have a 24 month cycle length as compared to IP months for the previous cycle.
The licer.see submitted a final camera-reedy copy of the previously requested TS on April 14, 1988.
The supplerents to the February 12, 1988 submittal dio not affect the, proposed TF change noticed in the Federal Register on April 15, 1988, with correcticri en April 79, 1988, and did not affect the staff's proposed no significant bezards determination.
The NFC staff has reviewed the application ard the supporting documents (Refs.
2 & 3) and has prepared the followino evaluation of the fusi design, nuclear design, thermal-hydraulic desien, and TS changes.
?.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 asserblies, 92 will be manufactured by Combustion Engineering (CE) and four by Advanced fluclear Fuels (ANF) Corporation, and are placed in the Cycle 10 ccre as an aid in qualifying ANF fuel fer ?a month cycle operation.
The 92 fresh CE assemblies will consist of 16 unshimred Patch M assemblies and 76 Batch M*
assemblies each containinc 1? B C rods for neutronic shiming anc having an 4
initial assembly averace enrichment of 4.08 weight percent (w/o) U-235. The 8805270136 000316 PDR ADOCK 05000317 P
PDR i
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I four ANF Batch MX demonstration assemblies c6ntain 12 fuel bearing Gd,0, rods for shimming and have an initial assembly average enrichment of 3.85 4/d U-235, 2.2 Mechanical Design The mechanical design of the CE Batch M reload fuel is identical to the Batch l
X 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 in Reference 3.
Most of the assembly and core interface dimensions are identical to the CF. fuel assemblies.
Diffe' ences in the upper and lower end fitting height and overall assembly height should not affect the performance of either fuel assembly.
Experience with similar ANF fuel designs co-residing adjacent to CE reload fuel in the Maine Yankee, Fort Calhoun, and St. Lucie Unit 1 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 Thermal 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 analyzec for the peak pin was greater than that ex;;ected at the end of I
Cycle 10, approximately 0.3% of the fuel pins will achieve burnups greater than the 52,000 MWD /T value approved for CE fuel (Ref. 6) if Cycles 9 and 10 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 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 appropriate fuel thermal design criteria required by the staff (Ref. 3).
Based on its review of the information discussed above, the staff concludes that the evaluation of the thermal design of the CE and ANF fuel for Cycle 10 is acceptable.
1
l o, 3.0 EVALUATION OF NUCLEAR DESIGN 1
3.1 Fuel Manaaement 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 2.1 of this SER.
The highest U-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/o U-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 accommodate a Cycle 10 length between 20,600 MWD /T and 21,800 MWD /T based on Cycle 9 lengths between 9,800 MWD /T and 11,800 MWD /T.
3.2 Power Distribution j
Hot full power (HFP) fuel assembly relative power densities are given in the reload analysis report for beginning-of-cycle (BOC), middle-of-cycle (M0C),
and end-of-cycle (E0C) unrodded configurations.
Radial power distributions at BOC and ECC 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 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 than those expected to occur at any time in Cycle 10.
Therefore, the predicted Cycle 10 power distributions are acceptable.
j 3.3 Reactivity Coefficients In order to accommodate 24 month cycles, the moderajor temperature coefficient (MTC) limit above 70% power is rais delta rho /* F to a value which vagies linearly from +0.3x10 gd from +0.2x10-delta rho /* F at 100% power to
+0.7x10- delta rho /* F at 70% power.
The staff has previously expressed concern about the pssitive MTC effect on the generic anticipated transients without scram (ATVS) assumptions and BGLE has stated that they will address the generic ATWS implications, if any, in the future.
In the interim, the staff has approved operation for core designs with allowable positive MTr values provided that the MTC becomes negative at 100% power and equilibr+um 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).
d-The Doppler coefficient for Cycle 10 is a best estimate value excected to be accurate to within 15%.
These reactivity coetticient values are bounded by the values used in the safety analyses for the reference cycle (Calvert Cliffs Unit 2 Cycle 81 The staff, therefore, finds the values of the PTCs and Doppler coefficients to be acceptable.
3.4 Control Reauirements The CEA worths and shutdown margin requirencnts at the most limiting time for the Cycle 10 nuclear desion, that is, for the ECC, are presented in Refererce 7.
These values are based on an ECC, het 7ero power (HZPI, steamline break accident.
At EOC 10, the reactivity wcrth with all CEAs inserted is 9.0%
delta rho.
An allowance of 1.15 delta rbo 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 'cr CEA insertion in accordance with the power dependent insertion limit (PDIL'.
The calculated scram worth is the total CEA worth less the worth c' the stuck CEA and less the worth of CEA insertion to the PDil and is 5.0% delta rho.
Geoucting 0.8% delta rho for physics uncertainty and bias yields a ret available scram worth of 5.1% delta rho.
Since the TS EOC shutdown margin at zero pcwer is 5.0", delta rho, a margin of 0.1% delta rho exists ir excess of the TS shutdown margin.
Therefore, sufficient CEA worth is available to accommodate the reectivity effects of the steam line break event at the worst time in core life allowing for the most reactive Cf.A stuck in the full 1,thdrawn position.
The staff cercludes that the licen.<ee's assessment of reactivity control is suitably conservative and that adec.Mte r.egotive reactivity worth has been provided by the control system to assure shutdown capability assuming a stuck CEA that resu'tr in the worst reactivity cer.dition for an E0C, HZP steamline bre49 cccident.
Thus, the control requirements are acceptable.
3.5 Safety Delated Data Other safety related data such as limiting parameters of dropped CEA reactivity worth and the maximum reactivity worth and planar power peaks asscciateo with an ejected CEA for Cycle 1.0 are identical to the values usec in the reference cycle and ere, therefore, acceptable.
4.0 EVALUATION OF TFERMAL-HYORAULIC DESIGN a,1 DNBR Analysis Steady state thermal-hydraulic analysis of CE fuel for Cycle 10 is cer'ormed usina the approved core thermal-hydraulic coce TCRC and the CE-1 critical heat flux correlation (Ref. 8).
The core and hot channel are ecdeled with the approved method described in CENPD-206-P-A (Ref. 9).
The design thermal maroin analvsis is performed using the fast running varietion of the TORC code, CETOP-D (Ref.10), which has been approved fer rahert Cliffs with the appropriate hot assembly inlet flow starvatien factors to assure its conservatism with respect to TORC.
The engineering hot channel factors for beat flux, heat input, rod pitch and cladding diameter are combined statistically with other uncertainty
- actors using the approved extenced statistical combination of urcerteinties IFSCU) method described in CEM-3aE(b)-P (Ref. 11) to arrive at an ec.uivalent departure from nucleate boiling ratic (CFBP) limit of 1.15 at a 95/95 probability / confidence level.
ChBR analyses were also performed to assess the cerformance of the ANF leae assemblies (Ref. 3) using the XCOBRA-III code (Ref. 17) and the ANF approved thermal-hydraulic methodology for mixed fuel cores (Ref.131 The XNB departure fro.m nucleate hoiling 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 teceptable to apply it to the mixed Cycle 10 core containing the four ANF lead fuel asserrblics.
The results indicate that the ANF lead assemblies exhibit hiaher MDNBDs than the hot CE assembly due to the 5% lower asstrbly power at which the ANF lead assemblies were sirruleted.
Since the insertion of the ANF lead assenhlies dces not significantly affect the mirirrum DNPR (MDNBR) of the hot CE asserr.bly, which establishes the core k'CUBR, the staff concludes that the ccre PCNGR is essentially unchar.ced by ir.sertion of the four ANF lead assemblies and thus the design critericn on CNBR is satisfiec by the mixed core containina ANF leed asser:blies.
Thus, the results of the DNBR analysis are acceptable.
4.2 Fuel Rod Scwing The fuel rod bcw per.alty accounts for the adverse irpact en MDNBR of randem variations in spacina between fuel rods.
The methodology for determining rod bcw cenalties for Calvert Cliffs was based cr the NPC approved trethods presented in the CE topical repcrt on fuel and poison rod bowing (Fef. 16).
The penalty at 45,000 MWD /T burnup is 0.006 in MCNPR.
This penalty is included in the ESCU uncertainty allowance discussed above.
Fcr those assemblies with average burnup in excess o# d5.000 MWD /T, sufficient margin exists to offset rod bow penalties.
The staff, therefore, concludes that the analysis of fuel rod bow penalty is acceptable.
5.0 FVALUA'!ON Or SAFETY ANALYSES 5.1 fon-LOCA F.vonts For the non-LOCA safety analyses, the licensee has determined that the key input parameters for the transient and accident analyses lie within the bounds of thcse c# the reference cycle (Unit 2 Cycle 81 As noted in Section 6.0, the shutdown margin TS is being changed from a singular value to a variable ranging frem 3.5% delta rho at POC to 5.0% delta rho at E00.
The E0C shutdown traroin recuirement 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 limitino than the reference analysis.
The staff, therefore, concludes that the non-LCCA transient and accident events for Cycle 10 are bounded by the refeience analyses and, therefore. the results of the non-LOCA safety analysis are acceptable.
5.2 LOCA Events The large break loss of coolant accident (LOCA) has been reanalv)ed for Cycle 10 to demonstrete that a peak linear heat generation rate (PLMR of 15.5 kw/ft conolies with the acceptance criteria of 10 CFP E0.46 for errergency core cooling systems (ECCS) for licht water reactors.
The Cycle 10 ariclysis, as the reference cycle analysis, was perforced with the 1985 CE evaluation medel which was approved in Reference 17.
The Cycle 10 analysis showed that the double ended guillotine pipe break at the pump discharge with a discherge i
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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 limitirig Cycle 10 break shcw 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 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 PLHGR 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 j
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 1
Section 3, the operating characteristics of Cycle 10 were calculated with approved methods.
The proposed TS are th'e results of the cycle specific analyses for, among other things, power peaking and control rod worths.
The analyses performed 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 reouested TS changes involve changes to both Unit 1 and Unit 2 TS.
Each proposed change is discussed below.
6.1 Figure 2.2-2 Thermal Marcin/ Low Pressure Trip Setooint-Part 1 Figure 2.2-2 is modified due to a revision in the curve fit for the TM/LP trip setpoint to accommodate the implementation of the extended statistical combination of uncertainties methodology.
The setpoint analysis uses this methodology and the licensee has determined that acceptable results are obtained for Cycle 10.
The changes to Figure 2.2-2 are, therefore, acceptable.
6.2 Figure 2.2-3 Thermal Margin / Low Pressure Trio Setooint-Part 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.
. 6.3 Bases 2.1.1 and 2.2.1 The text is modified to replace a specific MDNBR value with the phrase DNB SAFDL.
The use of a phrase in place of a specific MDNBR value was recomended in the extended SCU methodology (Ref.11) and approved by the staff (Ref.18).
The change is, therefore, accoptable.
6.4 Technical Specification 3.1.1.1 Shutdown Margin Two modifications are proposed for this TS.
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 in cycle.
Since the required shutdown margin varies throughout the cycle due to fuel depletion, boron concentration and moderator temperature and this variation with cycle time has been incorporated in all the appropriate safety analyses for Cycle 10, this change is acceptable.
The shutdown margin at E0C 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.
6.5 Technical Soecification 3.1.1.4 Moderator Temperature Coefficient The MTC limit above 70% power is being raisgd from +0.2x10'# delta rho /
F to
+0.7x10'ghich varies linearly from +0.3x10' a value del ta rho /* F at 100% power to delta rho /* F at 70% power.
This change is being implemented to accomodate 24 month cycles and to facilitate initial reactor startup at the j
beginning of the cycle.
The licensee has committed to a negative MTC at hot full power, equilibrium xenon condition value has been predicted to be -0.2x10'g.
As mentioned in Section 3.3, this delta rho /* F.
The feedline break analysis which supports this change is applicable to Cycle 10 and, therefore, the proposed change is acceptable.
6.6 Figure 3.1-2 CEA Group Insertion Limits The transient insertion limit between 90% and 100% power is being increased from an allowed insertion limit which varies linearly from 35% for Bank 5 at 90% 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 oscillations near EOC, has been incorporated into all of the Cycle 10 physics, safety and setpoint analyses and is, therefore, acceptable.
6.7 Figure 2.2-1 Axial Power Distribution Trip LSSS Figure 2.2-1 is modified to increase the positive and negative 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.
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. 6.8 Fiaure 3.2-2 Linear Feat Rate Axial Fluy Offset Control Limits And Fiaure 3.2-4 DNB Axial F e y. Offset Control Limits These Figures are modified to increase the negative ASI limits below 50% power.
The licensee has evaluated the effect o' the preposed new limits on the Unit 1 Cycle 10 and Unit ? Cycle 8 transient enalyses, margin to fuel centerline melt limits, margin to DNB limits, margin to LOCA PLHGR limit, core pcwcr versus planar radial peaking factor LCO, TM/LP LSSS, and core power versus integrated radial peaking 'acter LCO and has determined that acceptable results are obtained.
The changes are, therefore, acceptable for Unit 1 Cycle 10 and Unit ? Cycle C.
'C FI'PMARY The staff has reviewed the fuel system design, nuclear desien, thermal-hydraulic desicr, and the transient and accident analysis information presented in the Calvert Cliffs Unit 1 Cycle 10 relcad submittals.
Rased on this review, which is described above, the staf f concludes that the proposed Cycle 10 reload and associated modified TS are ecceptable.
This conclusicn is further based en the followina:
(1) previcusly reviewed and approved methods were used in the analyses; '?i the results of the safety analyses show that all safety criteria are met; and /3) the proposed TS are consistent with the reload safety analyses.
P.0 ENVIRONMENTAL CONSIDEPATION These amendments involve a change in the,nstallation or use of the facilities' comperents located within the restricted areas as define ( in 10 CFR 20 and chances in surveillance recuirements.
The staff has determired that these amendments invelve re significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has treviously issued a proposed finding that these emer.dments involve no significant ha:aros consideration and there has been no public comment en such finding.
Accordingly, these amencments meet the eligibility criteria for categorical exclusion set.forth in 10 CFP Soc 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or envircnmental assessment need be prepared in connection with the issuance of these amendments.
I 9.0 CONCLUSICN We have concluded, based on the considerations discussed abcse, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed ranner, and (2) such activities will be ecoducted in compliance with the Ccncission's regulations and the issuance i
of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Cated:
May 16, 1988 PRINCIPAL CONTR N'TCP:
L. Koop
v.
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-g-REFERENCES 1.
Letter from J. A. Tiernan (BG&E) to US NRC, B-88-011. "Calvert Cliffs Nuclear Power Plant Unit Nos. 1 and 2; Docket Nos. 50-317 and 50-318 Request for Amendment Unit 1 Tenth Cycle License Application; Unit Two Axial ~ Shape Index Region Enlargement," February 12, 1988.
2.
Attachment to B-88-011 Calvert Cliffs Unit Cycle 10 License Submittal.
3.
Appendix to B-88-011 Calvert Cliffs Unit 1 Cycle 10 License Submittal, ANF-88-019.
4 "Improvements to Fuel Evaluation Model," CEN-161(B)-P, Supplement 1-P (proprietary), April 1986.
5.
Letter from Scott A. McNeil (NRC) to J. A. Tiernan (BG&E), dated February 4, 1987.
6.
Letter from E. J. Butcher (NRC) to A. E. Lundvall, Jr. (BG&E), "Safety 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 Cycle 10 Response to Request for Additional Information," March 25, 1988.
8.
"Critical Feat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids. Part 1 Uniform Axial Power Distribution " CENPD-162-P-A, April 1975.
I 9.
"TORC Code, Verification and Simplified Fodeling Methods," CENPD-206-P-A, June 1981.
10.
"CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units 1 and 2 " CEN-191(B)-P, December 1981.
11.
"Extended Statistical Combination of Uncertainties," CEN-348(B)-P, January 1987.
12.
XCOBRA-IIIC:
A Computer Code to Determine the Distribution of Coolant During Steady-State and Transient Core Operation," XN-NF-75-21(P)(A),
January 1986.
13.
"Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," XN-NF-82-?.1(P)(A), Rev.1. September 1983.
14 "Exxon Nuclear DNB Correlation for PWR Fuel Designs," XN-NF-621(P)(A),
Rev. 1, September 1983.
e 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.
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