ML20086D213
| ML20086D213 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 11/21/1983 |
| From: | Rubenstein L Office of Nuclear Reactor Regulation |
| To: | Lainas G Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20085K039 | List: |
| References | |
| FOIA-83-722 TAC-51793, NUDOCS 8311300460 | |
| Download: ML20086D213 (16) | |
Text
'
i o
UfJITED STATES l
i 8
1 NUCLEAR REGULATORY COMMI ON
)
WASHINGTON, D. C. 20555
.NOV ; 1 1983 1
MEMORANDUM FOR:
G. C. Lainas, Assistant Director i
for Operating Reactors, DL FROM:
L. S. Rubenstein, Assistant Director for Core and Plant Systems, DSI
SUBJECT:
SAFETY EVALUATION FOR TURKEY POINT UNITS 3 AND 4 TECHNICAL SPECIFICATION CHANGES RELATED TO USE OF 0FAs AND WABAs (TACS 51793 AND 51794)
Plant Name:
Turkey Point Units 3 and 4 Docket Numbers:
50-250 and 50-251 Licensing Stage:
Operating Reactors Responsible Branch:
Operating Reactors Branch No.1 Project Manager:
D. Mcdonald Description of Review:
Technical Specification Change SER Review Status:
Complete Enclosed is the Core Perfonnance Branch safety evaluation of a proposed amendment to the Technical Specifications contained in Appendix A of Facility Operating License Nos. DPR-31 and DPR-41 for Turkey Point Units 3 and 4.
We conclude that reasonable assurance has been provided that (a) the planned design change from the Westinghouse 15X15 low parasitic (LOPAR) design to the 15X15 Optimized Fuel Assembly and (b) the planned use of Wet Annular Burnable Absorber (WABA) rods will not pose a threat to the health and safety of the public and that the proposed Technical Specification changes are, thersfore, acceptable.
D /,3
((-
c-ukiuu u e m L. S. Rube tein, Assistant Director for Core and Plant Systems, DSI
Enclosure:
As stated cc:
R. Mattson D. Eisenhut -
ISR'McDonard Vapge'4 D.
R. Capra Contacts:
M. Dunenfeld, DSI:CPB G. Hsii, DSI:CPB S. Wu, DSI:CPB X-28097 X-29473 X-29476 M ' 5HM%Tto paa h//
dd O
~
\\
SAFETY EVALUATION FOR TURKEY POINT UNITS 3 AND 4 TECHNICAL SPECIFICATION CHANGES FOR OPTIMIZED FUEL AND WET ANNULAR BURNABLE ABSORBER RODS (TACS 51793 AND 51797)
1.0 INTRODUCTION
By letters dated June 3,1983, and supplemented on November 16, 1983, Florida Power & Light Company submitted a request (Ref.1) for an amendment of the Technical Specifications contained in Appendix A of Facility Operating Licenses DPR 31 and 41. The Technical Specification changes are intended to accommodate: (1) a planned fuel design change from the Westinghouse (H) 15X15 low parasitic (LOPAR) design to the 15X15 Optimized Fuel Assembly (0FA) and (2) use of Wet Annular Burnable Absorber (WABA) rods.
2.0 FUEL MECHANICAL DESIGN Turkey Point Units 3 and 4 have been operating with all W 15X15 low-parasitic (LOPAR) fuel. The Unit 3 Cycle 9 and Unit 4 Cycle 10 cores will received W 15X15 0FAs resulting in a 1/3 0FA-2/3 LOPAR mixture.
Subsequent reloads are expected t.. eventually contain only 0FA fuel.
Although the W 15X15 0FA fuel is a new design, it is very similar to the W 15X15 standard low parasitic (LOPAR) fuel design. The major change intro-duced by the 15X15 0FA design is the use of 5 intermediate Zircaloy grids replacing 5 intemediate Inconel grids in the LOPAR fuel. The Zirca':,y grids have thicker and wider straps than the Inconel grids in order to closely. match I
the Inconel grid strength.
Furthermore, the 15X15 0FA Zircaloy grid design is similar to the W 17X17 0FA grid design, which was described in WCAP-9500-A.
l This report has been reviewed and approved by the NRC staff (Ref. 2).
l In perfoming our review of the 15X15 0FA fuel for Turkey Point Units 3 l
and 4, we have relied upon the D. C. Cook Unit 1 Cycle 8 reload report l
(Ref. 3) that the design criteria and evaluation methods used for 17X17 l
l
1
[ -
0FA in WCAP-9500-A were also used for 15X15 0FA. This information is also applicable to Turkey Point Units because identical fuel is used.
The Lalance of our review thus focused on those plant-specific issues identified in the SER for WCAP-9500-A insofar as they are applicable to Turkey Point Units 3 and 4.
Our evaluation of those issues follows.
2.1 Cladding Collapse l
The licensee uses an approved method described in WCAP-8377 (Ref. 4) to analyze cladding collapse. The result for Turkey Point shows that no cladding collapse is expected up to 40,000 EFPH (about 51,200 mwd /MTU peak-rod average burnup) for the new W fuel design. We conclude, therefore, that no cladding collapse is expected for the proposed and subsequent cycles of operation.
i 2.2 Fuel Themal Conditions
{
The Turkey Point submittal is based, in part, upon fuel themal analyses generated with a revised (Ref. 5) version of a previously approved W code called PAD (Ref. 6). This revision has been approved for generic reference in W fuel design (including 0FA) calculations.
2.3 Cladding Swelling and Rupture j
For large break loss-of-coolant accident analysis, the licensee used the approved 1981 large break ECCS evaluation model (Ref. 7), which includes 3
approved cladding swelling and rupture models. The use of this ECCS-model obviates the need for supplemental ECCS calculations mentioned in 4
the SER for WCAP-9500-A (Ref. 2). We thus find that cladding swelling l
and rupture have been adequately treated in the ECCS analysis.
l i
2.4 Seismic and LOCA Loads In 1975 asymmetri-blowdown forces on PWRs during LOCA was identified. As a result, NRC issued NUREG-0609 (Asymmetric Blowdown Loads on PWR Primary l
Systems) to address this concern and required all PWRs to submit'such an l
analysis for evaluating fuel assembly structural adequacy.
i 4
i
. >,. - -. -. ~. -.
a
- ~,,, - -
a
e J Westinghouse A-2 Owners Group, including Turkey Point Units 3 and 4, submitted two reports, WCAP-9558, Revision 2 and WCAP-9787 (Ref. 8), for staff review in response to PUREG-0609. They stated that a rapid blowdown is very un-likely because the stainless steel primary piping would leak before it breaks during a LOCA; therefore, the reports argue that the requirements of NUREG-0609 can be waived.
Although the review of W A-2 Owners Group reports has not yet been completed, no structural responte analysis of combined seismic and LOCA loads is presently being required from any A-2 Owner.
Although the issue of combined seismic and LOCA loads need not be resolved at l
this time, the analysis requirement remains for the seismic event alone. This is particularly so because the new 0FA fuel assemblies and the existing LOPAR fuel assemblies have slightly different structural properties as a result of incorporation of the new Zircaloy grid. Since OFA and LOPAR fuels will both be loaded in mixed configurations during the next few operating cycles, the licensee analyzed several mixed configurations for structural adequacy.
Generic nethods (WCAP-9401), which were previously reviewed and approved by NRC, were used for this analysis. Results show at least 20 percent margin relative to allowable limits for the spacer grid and more than a factor of 3 margin for other fuel assembly components including the functionally important thimble tubes. Based on the finding of adequate margins and the use of approved methods, we conclude that fuel assembly structural adequacy has been demonstrated.
2.5 Wet Annuler Burnable Absorbers Turkey Point Units 3 and 4 will utilize a new bjrnable poison design, the WABA, in the future cores. The WABA rod design consists of annular pellets of aluminum oxide and boron carbide (A10 -B C) burnable absorber material 23 4 encapsulated within two concentric Zircaloy tubings. The reactor coolant flows inside the inner tubing and outside the outer tubing of the annular rod. The topical report describing the WABA design (Ref. 9) has been recently reviewed and approv?d (Ref.10), and the utilization of W4BA rods in both units would thus be automatically approved subject to certain con-ditions described in the NRC staff's approval of the generic topical report
. (those conditions concern surveillance and the analysis of ccre bypass flow). The WABA surveillance is discussed in Section 2.7 and the analysis of core bypass flow is discussed in Section 4.0 of this SER.
2.6 Guide Thimble Diameter Reduction The 15X15 0FA guide thimbles are similar in design to those in the LOPAR fuel assemblies except for a 13 mil reduction in the ID and OD of the guide thimble above the dashpot. Because guide thimble tube fretting wear has been observed in some PWR designs, the NRC staff questioned the potentici for increased wear in the 0FA 15X15 design due to the reduction in clearance for the control rods.
However, W has shown in other cases (Ref.11) that results of the analysis for 0FA 15X15 guide thimble tube wear, using an approved technique (Ref.12),
were unchanged in the predicted guide tube wear compared to the }] 15X15 standard design. Based on the information presented, we agree with the licensee that the reduction in guide thimble to RCC rodlet clearance should have no adverse effect on the extent of guide tube wear and, consequently, there is reasonable assurance that (a) the structural integrity of the 15X15 0FA will be maintained with respect to load carrying capability of the guide thimble tubes, and (b) "scramability" will be maintained.
2.7 Post-irradiation Surveillance As indicated in SRP Section 4.2.II.D.3, a post-irradiation fuel surveillance progran should be established to detect anomalies or confinn expected fuel performance.
For a new fuel design, such as the 15X15 0FA, we normally request that a fuel surveillance program be developed for the first two lead plants utilizing the new design. Since two other operating reactors (other than Turkey Point Units 3 and 4) have been identified as lead plants for the 15X15 0FA design, we conclude that no special surveillance requirements are necessary for this fuel design change at Turkey Point.
l l
J L
. As for the WABAs, the licensee has agreed to have a supplementary surveillance program as described in Reference 10 if Turkey Point is the first or second lead plant to discharge the WABAs. We find this
[
acceptable.
2.8 Summary We have reviewed the fuel assembly mechanical design for Turkey Point Units 3 and 4.
We conclude that the fuel mechanical design, which includes the W 15X15 0FAs and the WABAs, is acceptable.
3.0 NUCLEAR DESIGN The proposed Technical Specification changes will allow the transition from lor:-parasitic 15X15 (LOPAR) assemblies to 15X15 0FA assemblies.
These OFA assemblies are identical to the LOPAR assemblies except that five of the interior Inconel grids have been replai i by Zircaloy grids.
The physics characteristics for the OFA fuel are on, slightly different from those of the LOPAR. These differences are within the nomal range of i
variations seen from cycle to cycle. They are due primarily to fuel manage-
)
ment considerations and not due to the fuel assembly design, i
The 15X15 0FA has features similar to the 'd 17X17 0FA which has been I
generically approved by NRC (Ref. 2). There has been considerable experience with the OFA fuel and recently its use was approved for the D. C. Coui Unit 1 l
Cycle 8 core and for the Zion _ Units 1 and 2 Cycle 8 cores. ~ The standard calculational methods as described in Reference 13 continue to apply.
Each reload core will be evaluated to assure that design and safety limits are l'
satisfied according to the reload methodology.. On this Lasis we approve
~
f use of the OFA for Turkey Point Units 3 and 4.
Acceptability of the WABA
~
l design is discussed in Section 2.5.
I
.i
')
[
i
a 6-4.0 THERMAL HYDRAULIC EVALUATION Since the Turkey Point Units 3 and 4 cores will be refueled with the 15X15 0FA fuel and the WABA rods, these cores will-have LOPAR-0FA mixed core configurations during the transition fuel cycles. The 15X15 0FA fuel has design features similar to the 15X15 LOPAR fuel except for the I
use of 5 intermediate Zircaloy grids for the OFA fuel to replace the 5 intemediate Inconel grids used in the LOPAR fuel. The Zircaloy grids have thicker and wider grid straps which result.in the OFA fuel assembly having approximately 4.5 percent increase in hydraulic resistance compared to the LOPAR assembly. Westinghouse has perfomed hydraulic tests at its fuel assembly test system facility to evaluate the hydraulic effects of the OFA-LOPAR mixed core. The tests were perfomed with a side-by-side OFA and LOPAR fuel assembly arrangement under hydraulic flow conditions approximating the reactor conditions. The results show that they are hydraulically compatible with the pressure drops within 3.5 percent of each other.
The themal hydraulic analysis of the mixed core is perfomed using the same methods described in the FSAR for the 15X15 LOPAR fuel except that the WRB-1 1
critical heat flux correlation is used for the OFA and the W-3 L-grid CHF correlation is used for the LOPAR fuel. The staff evaluation of the themal hydraulic analysis is summarized in the following.
(a) The WRB-1 correlation (Ref.14) was approved for the 17X17 0FA, and l
17X17 and 15X15 standard LOPAR fuel assemblies with DNBR limit of
(
l.17 for R-grid. No CHF test data is available for the 15X15 0FA and, therefore, the application of the WRB-1 correlation to the 15X15 0FA is of concern.
In response to staff questions during the D. C. Cook Unit 1, Cycle 8 reload review, _W provided the 14X14 0FA CHF test data and additional proprietary infomation regarding the design of the 15X15 0FA. The 15X15 0FA design is virtually identical l
e r
,m.._
e
. to the 15X15 R-grid design. A scaling tachnique was used in the 15X15 0FA grid design to ensure that the DNB performance is not affected by the 0FA grid. This scaling technique has also been used for the design of the 17X17 and 14X14 0FA grids.
In order to evaluate the effect of the geometry change on the accuracy of the WRB-1 correlation, W also perfomed a statistical analysis using the T-test and F-test for the 17X17 standard /0FA data and the 14X14 standard /0FA data. The results show that the null hypothesis, that the WRB-1 correlation pre-dicts the DNB behavior of the 0FA geometry with the same accuracy as the standard R-grid geometry, cannot be rejected at a 5 percent significance level.
For the case where the F-test rejects the null hypothesis, the OFA data have an appreciably lower variance which is indicative of better correlation accuracy. Therefore, even though no 15X15 0FA CHF data is available, the statistical analysis per-formed by W_ has provided the basis for the applicability of the WEB-1 correlation on the 15X15 0FA.
(b) The themal hydraulic analysis of a transitional mixed core has been previously reviewed by the staff (Ref.15) and approved with a con-dition requiring a penalty on DNBR to account for the uncertainty associated with the interbundle cross-flow in the mixed core.
The licensee has perfomed an analysis to detemine the required penalty factor in the same manner approved for the 17X17 0FA/LOPAR mixed core analysis. The result shows that a 3 percent penalty is required on the OFA for the transitional mixed core. The penalty will not be required for the full core OFA fuel.
l l
(c) The W WABA poison rod design is described in Wl,c'-10021, Revision 1 (Ref. 9) which has been approved by the staff.
In order to ensure no violation of the total core bypass flow limit, the total number
. of WABA rods in the core should be less than the upper limit es-tablished in Table 7.2 of WCAP-10021, Revision 1.
The licensee has indicated that a total of 160 WABA rods will be used in the Turkey Point Unit 3 Cycle 9 core. This number is far below the allowed limit and is, therefore, acceptable.
For other reload cores, the number of WABA rods will be required to be within the allowed limit.
(d) Using the approved method for rod bow penalty calculation described in the staff review of WCAP-8691 (Ref.16) the licensee indicated that the maximum rod bow penalty is 14 percent of DNBR corresponding to 85 percent gap closure. The staff independent calculation using the approved interim rod bow method (Ref.17) with the revised rod bow coefficients (Ref.18) has detemined a gap closure of 85.7 percent at 33,000 MWD /MTV. Because the physical burndown effect at higher burnup is greater than the rod bowing effect which would be calculated based on the amount of bow predicted at those burnups, the 33,000 MWD /MTU represents the maximum burnup of concern for rod bow penalty calc ~ulation.
(e) For the LOPAR fuel, DNBR is calculated with the W-3 L-grid CHF correlation with the design minimum DNBR limit of 1.30.
This value is 4.8 percent higher than the allowable DNBR limit of 1.24 derived from the 15X15 L-grid CHF test data. The analysis contains an inherent DNBR margin of 18.0 percent resulting from the use of conservative values of thennal diffusion coefficient and pitch reduction, the use of a conservative fuel densification model (Ref.19) and the difference in the design and allowable DNBR limits. This DNBR margin is more than enough to compensate for the rod bow penalty of 14.9 percent.
For the 15X15 0FA fuel, a plant-specific safety analysis DNBR limit of 1.56 is used in the thema1. hydraulic analysis. This safe'.y analysis DNBR limit has a 25 percent DNBR margin compared with the DNBR limit of 1.17 for the WRB-1 CHF correlation. This 25 percent margin is more than enough to account for the rod bow penalty of 14.9 percent, the transitional mixed core penalty of 3 percent and pcssibly small uncertainty associated with the application of the WRB-1 correlation on the 15X15 0FA fuel.
l
, (f) Based on the aforenentioned evaluation, we have concluded that the use of the 15X15 0FA fuel and the WABA rods in the Turkey Plant reloads is acceptable with the condition that the total number ot' WABA rods j
cannot exceed the upper limit imposed in Table 7.2 of WCAP-10021, Revision 1.
i i
3 i
I i
5.0 ACCIDENT AND TRANSIENT EVALUATION The accidents analyzed in the FSAR which could potentially be affected by the OFA design were reviewed.
Since the physics characteristics of the OFA design fall into the normal range of variations seen from cycle to cycle, as discussed in Section 3.0, these do not lead to a need for a reevaluation of the accidents and transients.
liowever, the 15X15 0FA guide thimbles are similar to their counterparts in the LOPAR fuel assemblies except for 13 mil ID and OD reduction in the guide thimble above the dashpot.
Due to the reduced clearance the shutdown and control rod drop time to the dashpot for accident analyses has been detemined to increase from 1.8 seconds for the LOPAR assembly to 2.4 seconds for the OFA. This increase could affect the " fast" transients for which the protection system trips the reactor within a few seconds.
An evaluation of the effect of rod drop time showed that all accidents and transients except the loss of flow, locked rotor and rod ejection are insignificant 1y affected by the increased rod drop time. These
" three accidents were reanalyzed to account for the increased rod drop time.
For the loss of reactor coolant flow accident with the ?.4 second scram time, the flow coastdown, nuclear power, heat flux and ')NBR ratio vs time curves were very similar to the case with the 1.8 seconds scram time.
The minimum DNBR of approximately 1.74 occurred at 3.6 seconds.
The locked rotor was reanalyzed and the figures for core flow coastdown, nuclear power, reactor coolant pressure and fuel clad temperature were similar to the previous ones. Less than 10 percent of the fuel rods exhibited a DNBR less than 1.56.
The peak clad temperature was 1953*F.
When the rod ejection accident was reanalyzed the changes in the maximum fuel centerline temperature, clad average temperature, fuel enthalpy and fuel centerline melt were very small, as can be seen from Table 1.
. The results of the reanalysis for all three accidents thus showed that the safety limits and applicable criteria are satisfied with the OFA. We, therefore, find the increased rod drop time acceptable.
6.0 TECHNICAL SPECIFICATIONS The Technical Specification changes proposed for this amendment involve:
(a)
Pages 3.2-2, B3.2-2 This change permits an increase in the shutdown and control rod drop time.
It is acceptable, as discussed in Section 5.
(b) Page 5.2-1 This change permits the use of WABA rods.
It is acceptable, as discussed in Section 2.6.
(c) Pages B2.1-1, B2.1-2, B2.3-2, B2.3-3, B3.1-1, B3.2-3 and B3.2-8.
The Technical Specification Bases on these pages have been changed by removing the DNBR limit specifically for the W-3 corelation.
These changes are made to allow the use of the WRB-1 correlation DNBR limit for the OFA fuel. Since the 15X15 0FA fuel is acceptable for the Turkey Point plants, as discussed in Section 4.0, the Technical Specification changes are acceptable.
7.0 CONCLUSION
S We have reviewed the proposed changes to the Turkey Point Units 3 and 4 Technical Specifications involving the use of 15X15 0FAs and WABAs and find they pose no significant hazard. The transition to 0FA fuel and WABA rods will not cause any change in the types or increase in the amount of effluents or any change in the authorized power level of the facility. The amendment therefore does not:
l t
1 (a) involve a significant increase in the probability or consequences of an accident previously evaluated; or (b) create the possibility of a new or different kind of accident from any accident previously evaluated; or (c) involve a significant reduction in a margin of safety.
t l
i
1 x.
E-TABLEl 3-r PTP 3 Cycle 8 PTP 3 Cycle P
='
?
Scram Time 1.8 Sec.
Scram Time 2.4 Sec.
BOC'*
.!~
Mx-Povs 0%
502%
0%
102%
e' 102%
Ma -m Fuel Center Temp F"?)
2565 5185 2626 Melt
- 3319 4687 Mc:inum Clad Average Teenp PF) 1624 2367 1634 2397 2026 2066
{
Mc:imumFuelErdlialpyCal/gn 84.0 177.0 88.0 181.0 116.9 152.6 h-n Fuel Centerline Melt (%)
0
<lo 0
<10%
0 0
}
i i
- Less sem 10% of fuel enett at fuel rod hot spot
- EOC not reanalyzed for Twleey Paint Unit 3 Cycle 8
i
8.0 REFERENCES
I 1.
R. E. Uhrig (FP&L) letter, L-83-344, to D. C. Eisenhut (NRC),
" Turkey Point Units 3 and 4, Docket No. 50-250 and 50-257, Proposed License Amendment, Optimized Fuel Assembly and Wet Annular Burnable Absorber," June 3,1983.
4 2.
R. L. Tedesco (NRC) letter to T. M. Anderson (W), " Reference Core i
Report 17X17 Optimized Fuel Assembly," May 22, 1981.
3.
R. F. Hering (AEP) letter to H. R. Denton (NRC), August 31, 1983.
4.
V. Stello (NRC) memorandum to R. DeYoung (NRC), " Evaluation of Westinghouse Report WCAP-8377, Revised Clad Flattening Model,"
January 14, 1975.
5.
W. J. Leech, D. D. Davis and M. S. Benzvi, " Revised PAD Code Thermal Safety Model," Westinghouse Electric Corporation Report WCAP-8720, Addendum 2 (Proprietary), October 1982. Submitted by E. P. Rahe, Jr. (W) letter NS-EPR-2673 to C. O. Thomas (NRC),
October 27, 1982 l
6.
J. V. Miller, et al., " Improved Analytical Models Used in' Westing-house Fuel Rod Design Computations," Westinghouse Electric Corpo-ration Repcrts WCAP-8720 (Proprietary), October 1976, and WCAP-8720, Addendum 1 (Proprietary), September 1979.
7.
E. P. Rahe, " Westinghouse ECCS Evaluation Model,1981 Version,"
i WCAP-9220-P-A (Proprietary), Revision 1,1981.
8.
" Mechanical Fracture Evaluation of Reactor Coolant Pipe Containing l
a Postulated Circumferential Through-Wall Crack," WCAP-9558, Revision 2, May 1982; " Tensile and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation,"
l WCAP-9787, May 1981.
9.
E. P. Rahe, Jr. (W), letter to C. O. Thomas (NRC), " Westinghouse 3
WABA Evaluation Report," WCAP-10021, Revision 1, (Proprietary),
l October 18, 1982.
10.
L. S. Rubenstein (NRC) memorandum to F. J. Miraglia, ~"SER of W WABA Design," June 1,1983.
11.
F. G. Lentine (CECO) letter to H. R. Denton (NRC), October 21, 1983.
j 12.
L. S. Rubenstein (NRC) memorandum to T. M. Novak (NRC), June 6,1980.
1 I
13.
F. M. Bordelon, et al. " Westinghouse Reload Safety Evaluation Methodology," WCAP-9273 (Hon Proprietary), March 1978.
- 14. WCAP-8762, "New Westinghouse Correlation WRB-1 For Predicting Critical Heat Flux in Rod Bundles With Mixing Vane Grids," July 1976.
- 15. Memorandum from L. S. Rubenstein (NRC) to T. M. Novak (NRC),
" Supplemental Information on WCAPJ $00 and WCAPJ $01/9402 Mixed Core Compatibility," December 14, 1982.
- 16. Letter from J. Stolz (NRC) to T. M. Anderson (W), " Staff Review of WCAP-8691," April 5,1979.
- 17. Memorandum from D. Ross and D. Eisenhut (NRC) to D. Vassallo and K.
Goller, " Revised Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors," February 16, 1977.
- 18. Memorandum from R. O. Meyer to D. F. Ross, " Revised Coefficients for Interim Rod Bowing Analysis," March 2,1978.
19.
J. M. Hellman, Ed., " Fuel Densification Experimental Results and Model for Reactor Application," Westinghouse Electric Corporation Report WCAP-8218-A (Non-proprietary), March 1975, and WCAP-8218-P-A (Proprietary), March 1975.
l
-