ML15112B116
| ML15112B116 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 08/03/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML15112B115 | List: |
| References | |
| NUDOCS 8308290016 | |
| Download: ML15112B116 (11) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 122 TO FACILITY OPERATING LICENSE NO.
DPR-38 AMENDMENT NO. 122TO FACILITY OPERATING LICENSE NO.
DPR-47 AMENDMENT NO. 119TO FACILITY OPERATING LICENSE NO.
DPR-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS NOS. 1, 2 AND 3 DOCKETS NOS.
50-269, 50-270 AND 50-287 1.0 Introduction By letter dated May 19, 1983 (Ref. 1), Duke Power Company (Duke or the licensee) made application to modify the Oconee Nuclear Station Technical Specifications in support of Cycle 8 operation of Unit 1. The analysis performed and the resulting modifications to the Station's common Technical Specifications are described in the Unit 1 Cycle 8 Reload Report (Ref. 2).
The application also includes a modified version of the Oconee Nuclear Station Generic Startup Physics Test Program Report (Ref. 3).
The safety analysis for the previous seventh cycle of operation at Oconee Unit 1 is being used by the licensee as a reference for the proposed eighth cycle of operation. Where conditions are identified as limiting in the seventh cycle analysis, our previous evaluation (Ref. 4) of that cycle continues to apply.
1.1 Description of the Cycle 8 Core
._.__The OconeeUnit 1 Cycle 8 core will consist of 177 fuel assemblies, each_ of which is a 15x15 array containing 208 fuel rods, 16 control rod guide tubes, and one incore instrument guide tube. Cycle 8 will operate in a bleed-and-feed mode with core reactivity control supplied mainly by soluble boron in the reactor coolant and supplemented by 61 full length control rod assemblies (CRAs) and40 burnable poison rod assemblies (BPRAs).
In addition, eight axial power shaping rods (APSRs) are provided for additional control of the axial power distribution.
The length of Cycle 8 is expected to be 410 effective full power days (EFPD) of operation, somewhat lower than the 427 EFPD accumulated during Cycle 7. Due to the shorter cycle length, the average burnup for Cycle 8 will also be lower than the previous cycle, 12,858 MWd/MtU as compared to 13,363 MWd/MtU. The licensed core full power level remains at 2568 MWt.
2.0 Evaluation of the Fuel System Design 8308290016 8300 PDR ADOCK 05000269 P
0PDR
-2 2.1 Fuel'Assembly Mechanical Design The 65 Babcock and Wilcox (:B&W) Mark-B4 fuel assemblies loaded as Batch 10 at end of Cycle 7 (EOC 7) are mechanically.interchangeable with Batch 8 and 9 fuel assemblies loaded previously at Oconee Unit 1. The Mark-B4 fuel assembly has been previously approved (Ref. 4) by the NRC staff and is utilized in other B&W nuclear steam supply systems. The oldest fuel in the core (designated Batch 8C) consists of 44 assemblies of the standard Mark-B4 design.
Batch 9 consists of 64 assemblies of the standard design (designated Batch 9A) and also includes four once-burned Mark-8Z fuel assemblies (designated Batch 9B).
The Mark-BZ design is similar to the standard Mark-B4 except that six imtermediate Inconel spacer grids have been replaced with Zircaloy grids. The design (Ref. 5) of these demonstration assbmblies was reviewed and approved by the NRC staff (Ref. 4) for the previous cycle (Cycle 7) of operation at Oconee 1.
We continue to find the use of these demonstration assemblies acceptable.
We are aware of a number of other recent changes to the B&W 15x15 fuel assembly design (e.g., a larger fuel assembly holddown spring,fuel pellets manufactured by an alternate supplier).
These changes have been approved for use in other operating B&W 177-fuel-assembly plants on a limited basis and may be incorporated into future cycles of operation at Oconee Unit 1. However, for the current cycle of operation, the licensee has identified no other changes in the fuel assembly mechanical design. We find-this acceptable.
2.2 Fuel Rod Design Although the 65 fresh fuel assem5ltes in Batch 10 are all of the Mark-84 design (and externally similar), five assemblies (denoted Batch 10A and 108) will contain fuel pellets containing both urania (U02) and gadolinia (Gd203) as described in Reference 6. These five lead test assemblies (LTAs) are part of a joint Duke Power/Babcock & Wilcox/Department of Energy program to develop and demonstrate an advanced fuel design incorporating U02-Gd2O3 for extended burnup in pressurized water reactors (PWRs).
Gadolinium is a so-called burnable poison. That is, it contains isotopes whi.ch have large absorption cross sections and are converited to isotopes of low absorption cross section as the result of neutron absorption.
Thus the increase in reactivity accompanying the burnup of the poison compensa-tes to some extent for the decrease in reactivity due to fuel burnup and the accumulation of fission product poisons.
Gadolinia is commonly used in..boiling water reactor fuel designs (Ref. 7-9), but its use in PWRs has been limited. In addition to changing the neutronic properties (e.g., radial power distribution) of the fuel,.the introduction.of gadolinia is known to affect the physical properties (e-.g.,
therma:1-conductivity, melting point) as well.
.The effects of -gadoliraia on irradiation properties (e.g., fuel swelling and fission gas release), parti cularly at the-higher concentrations proposed for PWR designs, is not wiell
- known,
Under the -provisions of 10 CFR 50.59, a licensee may conduct tests or experiments (i.e., incorporate LTAs in a reload core design) without prior NRC notification or approval.
One of these provisions is that the proposed test or experiment does not involve a change in the plant Technical Specifications. In.the case of Oconee Unit 1, the power-to-incipient-centerline melt limit (Page 2.1-2 of the Technical Specifications) has been modified to account for the lower power-to-melt values calculated for the gadolinia rods.
However, a limited number of gadolinia-bearing rods are present in each LTA and the power peaking in those rods (for Cycle 8) is significantly less than the expected power peaking of gadolinia-free rods in either the LTAs or any other fuel assembly in the core. Thus, the licensee has concluded (Ref. 6) that the loading of the five extended-burnup lead test assemblies in the Oconee 1 Cycle 8 core will not adversely affect either the nuclear, mechanical, or thermal-hydraulic character of the reactor, or the existing safety anaylsis. Since the addition of gadolinia to the Oconee 1 Cycle 8 core appears to change only the design, as opposed to operating limits in the Technical Specifications, it is debatable whether formal review of this application iT required. Nevertheless, we have reviewed this application and find it acceptable for the reasons discussed below.
We have, however, examined the licensee's report (BAW-1772 -.Ref. 6) and noted that the evaluation performed includes all, fuel system, nuclear and thermal hydraulic desian analyses, as well as the transient and accident evaluations, considered in a standard reload safety analysis. With the exception of the gadolinia properties (i.e., physical, neutronic and irradiation behavior),
which were not covered in the report, the evaluation utilized methods previously reviewed and approved by the NRC. Because of the exception, we are unable to confirm the licensee's findings.
However, the results appear similar to those submitted by other manufacturers of gadolinia-bearing fuel and the report provides a reasonable basis for our approval of the LTA irradiations.
Our approval is limited to Cycle 8 (as the analyses presented were limited to Cycle 81 and should not be construed as an approval of this.design for full
-scale applications (because of the use of unreviewed.properties and other information).
We will pursue the issue of the continued irradiation.of these lead test assemblies at the time the Oconee 1 Cycle 9 reload application is made.
7 -
cladding-stress, strain and collapse analyses for-the standard fuel designs in the core are bounded by conditions previously analyzed for Oconee Uhit 1 of were analyzed specifically for Cycle 8 using methods and limits previously reviewed and approved by the NRC.
We find that no further review of these areas
=is necessary.
2.2.1 Rod Internal Pressure Section 4.2 of the Standard Review Plan (SRP) (Ref. 10) addresses a number of acceptance criteria used to establish the design bases and evaluation of the fuel system. Among those which may affect the operation of the fuel rod is the internal pressure limit.
The-acceptance criterion (SRP 4.2,Section II.A.1(f))
is that the fuel rod internal gas pressure should remain below normal system pressure unless otherwise justified.
-4 The licensee has stated that the fuel rod internal pressure will not exceed nominal system pressure during normal operation for Cycle 8. This analysis.is based on the use of the B&W TAFY-3 code (Ref. 11) rather than one of the newer B&W codes, TACO-1 (Ref. 12) and TACO-2 (Ref. 13).
Although all of these codes have been approved for use in safety analysis, we believe (Ref. 14) that only the newer TACO series of codes are capable of correctly calculating fission gas release (and therefore rod pressure) at very high burnups. Babcock & Wilcox has responded (Ref. 15) to this concern with an analytical comparison between the TAFY-3 and TACO-1 codes. In this response, they have stated that the fuel rod internal pressure predicted by TACO-1 is lower than that predicted by TAFY-3 for-fuel rod exposures of up to 42,000 MWd/MtU. Although we have not examined this comparison, we note that the analyses exceed the maximum expected exposure (40,238 MWd/MtU) for all fuel rods in the Oconee Unit 1 core at the end of Cycle 8. We conclude that the rod internal pressure limits have been adequately considered for Cycle 8 operation.
2.3 Fuel Thermal Design The thermal behavior of all fuel in the Cycle 8 core, with the exception of the gadolinia-bearing lead test assemblies, is virtually identical.
In general, the thermal analysis was performed with the approved version of TACO-2. We find this acceptable.
For the Loss of Coolant Accident (LOCA) analysis (Section 7.2 of the Reload Report), the average fuel temperature as a function of linar heat rate and the lifetime pin pressure data were calculated with the older TAFY-3 code. The licensee has stated that the fuel temperature and pin pressure data used in the generic LOCA analysis are conservative compared with those calculated for Cycle 8 at Oconee Unit 1.
As mentioned previously, B&W currently has several fuel performance codes which are-approved and could be used to calculate LOCA initial conditions.
The older TAFY-3 code was used for the generic LOCA analysis cited in the Oconee Unit 1 Cycle 8 Reload Report.
Information obtained y the NRC staff (Ref.
- 16) indicates that the TAFY-3 code predictions do not produce higher calculated peak cladding temperatures in the generic LOCA analysis than the newer TACO-1 or TACO-2 codes as suggested by the licensee. The issue involves excessive fuel densification and lowered fuel rod internal gas pressures at beginning of life. Babcock and Wilcox has proposed a method of resolving this issue which has been adopted.by Duke Power Company (Ref. 17).
The method relies on reduced peak linear heat rate (PLHR) limits at low core elevations for the first 26 EFPD of operation based on comparison of TAFY-3 and TACO-2 calculated LOCA initial conditions.
The method is similar to an older TAFY-3/TACO-1 comparison (Ref. 2) used in the original Oconee Unit 1 Cycle 8 safety analysis. However, the resulting PLHR reduction is different for each code.
In addition to the issue of initial fuel temperatures and rod internal pressures u-sed in-the LOCA analysis, a second issue involving cladding swelling and rupture models has affected the proposed Cycle 8 operating limits for Oconee 1.
-5 In 'late 1979, the NRC staff reviewed Emergency Core Cooling System fuel cladding models in light of new data. Adequacy of the models then in use was questioned and new models, developed as Appendix K acceptance criteria, were presented in NUREG-0630 (Ref. 18).
Each fuel vendor was then asked to shoa how, in light of the new models, the plants analyzed with their analytical methods continued to meet the applicable LOCA limits.
The B&W response (Ref. 19) concluded that the impact of the NRC models was small and did not result in analytical results in excess of the LOCA limits.
A more recent B&W calculation (Ref. 20), however, found that the cladding swelling and rupture models presented by the staff have a non-trival effect on LOCA peak cladding temperatures in B&W 177 fuel assembly plants. Because this calculation was applicable to all B&W plants, the licensee was requested (Ref. 21) to provide supplemental calculations for Oconee Unit 1 similar to those provided in Reference 20. The licensee's responses (Refs. 22, 23 and 24) culminated in the supplemental calculation (Ref. 17) cited previously. This calculation, which considers both fuel densification (TAFY-3/TACO-2) and cladding swelling and rupture effects, results in low core elevation PLHR limits which are more restrictive than those which consider only fuel densification. The licensee has proposed (Ref. 17) modification to the Oconee Station Technical Specifications which account for these reduced PLHR limits.
In general, the supplemental calculation utilizes previously approved methods except for the substitution of the NRC cladding models. However, there are segments of the analysis (e.g. THETAl-B - Ref. 25) that are currently undergoing NRC review. The licensee has also presented results from a calculation using a new FLECTSET heat transfer correlation (Refs. 26 and 27).
This correlation appears to offset the NUREG-0630 penalties.
The licensee has not yet claimed these FLECTSET benefits, however, because the benchmarking and other final evaluations of FLECTSET have not been completed and provided to the NRC for review.
Considering the above, we conclude that the licensee's-proposed Technical Specification changes are both appropriate and necessary. Since these operating limits are more restrictive than those previously used at the Oconee Station,. since they are only needed for a brief time period, and since potenti albut
-unused compensating benefits may exist, we, therefore, conclude that tne operating restrictions imposed on an interim basis are acceptable for inzorporating the 1V1REG-0630 penalties until our final evaluation of FLECSET is completed.
3.0 Evaluation of the Nuclear Design
- 31. Physics.Characteristics The nuclear characteristics of the Oconee 1 Cycle 8 core have been computed by methods previously used and approved for B&W reactors.
Comparisons are made between the physics parameters for Cycles 7 and 8. The differences that exist between the parameters are due to the decreased cycle length and the higher average burnable poison enrichment which tends -to decrea-se values of critical boron concentrations. Changes in the radial flux and burnup distributions between cycles also account for the differences in control rod worths, including ejected and stuck rod worths.
All safety criteria are still met.
Beginning-of-cycle radial power distributions show acceptable margins to limits.
-6 Shutdown margin calculations for Cycle 8 include the effects of poison material depletion, a 10% calculational uncertainty, and flux redistribution as well as a maximum worth stuck rod. Beginning and end-of-cycle shutdown margins show adequate reactivity worth exists above the total required worth during the cycle. The required shutdown margin is 1.00%4k/k, the shutdown margin at the beginning and end-of-cycle is 3.11% Ak/k and 2.28% 4k/k, respectively.
Based on our review, we conclude that approved methods have been used, that the nuclear design parameters meet applicable criteria and that the nuclear design of Oconee 1 Cycle 8 is acceptable.
3.2 Gadolinia Lead Test Assemblies Four of the unirradiated Mark-GdB fuel assemblies in the Oconee 1 Cycle 8 core will have an initial enrichment of 4.0 weight percent U-235. Storage of this maximum enrichment in the Oconee Unit spent fuel storage racks has been reviewed and approved by the staff (Ref. 28).
The effect of these higher enriched fuel assemblies on the nuclear design has been taken into account for Cycle 8 and the design continues to meet all criteria including those applicable to radial power peaking, ejected rod worths, and shutdown margin.
3.3 Startup Physics Test Program A modified version (Ref. 3) of a report entitled "Oconee Nuclear Station Generic Startup Physics Test Program" has been reviewed.
The modifications consist of:
- 1. Changing the names of the 40% FP, 75% FP and 100% FP Power Distribution Tests to low, intermediate and full power core mapping.
- 2. Establishing power ranges for this mapping and for the core symmetry test.
- 3. Changes to the critical boron concentration tests.
Establishing power ranges for the mapping permits operating flexibility with no loss of information from the tests.
Establishing a lower power range for the symmetry test:takes advantage of a planned change to the computer software and will allow earlier identification of core power distribution problems. The changes to the critical boron concentration tests are minor.procedural changes.
We have reviewed these changes and find them acceptable.
-4,0 Evaluation of the Thermal-Hydraulic Design In Section 6 of the B&W report BAW-1774 (Ref..2), the licensee has described the thermal-hydraulic design. Cycle 7 is used as the reference cycle for -the thermal-hydraulic evaluation.
Table 1 shows a comparison-of the maximum design conditions for Cycles 7 and 8. As seen from the table,- the maximum design
-conditions are unchanged from Cycle 7.
-7 Cycle 8 fuel includes four Mark BZ-demonstration assemblies and five LTA demonstration assemblies. These assemblies have a design peak of 1.61 (6%
peaking reduction) to ensure that'they are not thermally limiting.
All other assemblies have a 1.71 design radial times local peak.
We find the incorporation of these four Mark-BZ and five LTA demonstration assemblies acceptable since their design peak is at least 6% less than the remaining assemblies.
A rod bow penalty was calculated using an approved (Ref. 29), interim procedure for calculating departure from nucleate boiling ratio (DNBR) reduction due to rod bow. The licensee used the maximum fuel assembly burnup of the batch that contains the maximum radial-local peak. For Cycle 8, that burnup is 17,511 MWd/MtU in a Batch 1OC assembly. The resultant net rod bow penalty, after inclusion of the 1% flow area reduction credit, is 0.2% reduction in DNBR. We find this acceptable since the thermal-hydraulic design for Cycle 8 includes a margin greater than 0.2% above the minimum DNBR of 1.30.
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=20
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(a)
(a)
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1.014 1.014
!.aw aria, 0.98 0.98 Mit=
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See Tab Is 4-2 --of Re ferenme 2 Cb )3&sed om denstfia lagth of 140.3 Lu.
5.0 Evaluation of Accident and Transient Analysis The key kinetics parameters for Oconee 1 Cycle 8 have been compared to the values used in the Final Safety Analysis Report (FSAR) and densification report. It is shown that in all cases Cycle 8 values are bounded by those previously used. We conclude that the FSAR or previous reference cycle transient and accident analyses are valid.
Three sets of bounding values for allowable LOCA peak linear heat rates are given as a function of core height (Ref. 17).
These limits apply during the periods 0-26 EFPD,26-200 EFPO, and 200 to end-of-cycle. The limits are satisfactorily incorporated into the Technical Specifications for Cycle 8 through the operating limits on rod index, axial power shaping rod limits, and axial power imbalance.
6.0 Evaluation of Technical Specification Modifications We have reviewed the proposed Technical Specifications for Cycle 8. The limiting conditions for operation have been established by previously used and approved methods. The rod withdrawal limits for the various pump combinations and times in life are presented.
On the basis that previously approved methods were used to obtain the limits, we find them acceptable.
7.0 Summary We conclude from the examination of Cycle 8 core thermal and kinetic properties with respect to acceptable previous cycle values and with respect to the FSAR values, that this core reload will not adversely affect-the Oconee Nuclear Station's ability to operate safely during Cycle 8 of Unit 1.
8.0 Environmental Consideration We have determined that the amendments do 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 amendments involve 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 these amendments.
9.0 Conclusion We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted 'in.
compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Dated.:
August 3, 1983 The following NRC staff personnel have contributed to this Safety Evaluation:
J. Suermann, J. Vogelwede, L. Kopp, A. Gill, M. Chatterton.
-10 REFERENCES
- 1. H. B. Tucker (Duke) letter to H. R. Denton (NRC).on "Oconee Nuclear Station Unit 1" dated May 19, 1983.
- 2. "Oconee Unit 1 Cycle 8 Reload Report," Babcock & Wilcox Company Report BAW-1774, February 1983. Attachment 4 to Reference 1 above.
- 3.
"Duke Power-Company Oconee Nuclear Station Generic Startup Physics Test Program," Duke Power Company Report, March 1983. to Reference 1 above.
- 4. P. C. Wagner (NRC) letter to W. 0. Parker (Duke) on Oconee Unit 1 Cycle 7 reload dated November 30, 1981 and transmitting Amendments No. 105, 105 and and 102 to Facility Operating Licenses No.
DPR-38, DPR-47 and DPR-55, respectively.
- 5. "Mark BZ Demonstration Assemblies in Oconee 1, Cycles 7, 8, and 9," Babcock
& Wilcox Company Report BAW-1661, March 1981. Transmitted by J. H. Taylor (B&W) letter to J. Stolz (NRC) dated April 10, 1981.
- 6. "Gadolinia-Bearing Lead Test Assemblies Design Report," Babcock & Wilcox Company Report BAW-1772-P (Proprietary), June 1983. Transmitted by J. H. Taylor (B&W) letter to J. Suermann (NRC) dated July 12, 1983.
- 7. L. D. Gerrald and G. C. Cooke, "Gadolinia Fuel Properties for LWR Fuel Safety Evaluation," Exxon Nuclear Company Report XN-NF-79-56 (P),
Revision 1, August 1979.
- 8. G. A. Potts, "Urania-Gadolinia Nuclear Fuel Physical and Irradiation Characteristics and Material Properties," General Electric Company Report NEDE-20943-P (Proprietary), January 1977.
- 9. S. 0. Akerlund, et al.,, "The GESTR-LOCA.and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Volume 1:
GESTR-LOCA A Model for the Prediction of Fuel Rod Thermal Performance, Appendix 8-Urania-Gadolinia Nuclear Fuel Material Properties," General Electric Company Report NEDE-23785-1-P (Proprietary),
December 1981.
- 10. U. S. Nuclear Regulatory Commission Standard Review Plan Section 4.2 (Revision 2), "Fuel System Design," U. S. Nuclear Regulatory Commission Report NUREG-0800 (formerly NUREG-75/087), July 1981.
- 11. C. D. Morgan and H. S. Kao, "TAFY - Fuel Pin Temperature and Gas Pressure Analysis," Babcock and Wilcox Company Report BAW-10044, May 1972.
- 12. R. H. Stoudt et al., "TACO: Fuel Pin Performance Analysis," Babcock and Wilcox Company Report BAW-10087P-A, Rev. 2, August 1977
- 13. Y. H. Hsii et al., "TACO2: Fuel Pin Performance Analysis," Babcock and Wilcox Company Report BAW-10141P, January 1979.
- 14. D. F. Ross, Jr. (NRC) letter to J. H. Taylor (B&W) dated January 18, 1978.
- 15. J. H. Taylor (B&W) letter to L. S. Rubenstein (NRC) dated September 5, 1980.
- 16. R. 0. Meyer (NRC) memorandum for L. S. Rubenstein (NRC) on "TAFY/TACO Fuel Performance Models in B&W Safety Analysis" dated June 10, 1980.
- 17. H.-B. Tucker (Duke) letter to H..R. Denton (NRC) on "Oconee Nuclear Station" dated July 13, 1983.
- 18. 0. A. Powers and R. 0. Meyer, "Cladding Swelling Models for LOCA Analysis,"
U. S. Nuclear Regulatory Commission Report NUREG-0630, April 1980.
19..J. H. Taylor (B&W) letter to L. S. Rubenstein (NRC) dated October 28, 1980.
- 20. J. W. Cook (Consumers Power) letter to H. R. Denton (NRC) dated April 2, 1982 and transmitting B&W Report No. 12-1132424, Revision 0, "Bounding Analysis Impact Study of NUREG-0630."
- 21. T. M. Novak (NRC) letter to W. 0. Parker (Duke) dated July 13, 1982.
- 22. H. B. Tucker (Duke) letter to H. R. Denton (NRC) on "Oconee Nuclear Station" dated August 12, 1982.
- 23. H. B. Tucker (Duke) letter to H. R. Denton (NRC) on "Oconee Nuclear Station" dated March 7, 1982.
- 24. H. B. TUcker (Duke) letter to H. R.-Denton (NRC) on "Oconee Nuclear Station" dated June 6, 1982 and transmitting "Bounding Analytical Assessment of NUREG-0630 on LOCA and Operating kW/ft Limits," Babcock and Wilcox Company Document No. 77-1141256-00.
- 25. "Babcock & Wilcox Revisions to THETAl-B, a Computer Code for Nuclear Reactor Core Thermal Analysis (IN-1445) - Revision 3," Babcock & Wilcox Company
- Report BAW-10094, Rev 3, February 1981.
- 26. N. Lee, S. Wong, H. C. Yeh, and L. E. Hochreiter, "PWR FLECHT SEASET Unblocked Bundle, Forced and Gravity Reflood Test Data Evaluation and Analysis Report," NUREG/CR-2256 (EPRI NI-2013 or WCAP-9891), November 1981.
- 27. G. P. Lilly, et al., "PWR FLECHT Skewed Profile Low Flooding Rate Test Series Evaluation Report," WCAP-9183, November 1977.
- 28. L. S. Rubenstein (NRC) memorandum for G. C. Lainas (NRC)-dated May 25, 1983.