ML20236C424
| ML20236C424 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 03/13/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20236C411 | List: |
| References | |
| NUDOCS 8903220143 | |
| Download: ML20236C424 (10) | |
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'o UNITED STATES
- "g NUCLEAR REGULATORY COMMISSION o
E WASHING TON, D. C. 20565 t
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q SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 57 TO FACILITY OPERATING LICENSE N0 J PF-29 l
SYSTEM ENERGY RESOURCES, INC., et al.
GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416
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1.0 INTRODUCTION
By letter dated December 6, 1988, as supplemented December 30, 1988 and
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January 31, 1989 (Reference 1),
System Energy Resources, Inc. (the j
licensee), requested an amendment to Facility Operating License No. NPF-29 for the Grand Gulf Nuclear Station, Unit 1 (GGNS-1).
The proposed anxndment would change the Technical Specifications (TS) as required for the reload and operation of Cycle 4.
The requested TS changes and reports discussing the reload and analyses to support and justify Cycle 4 operation were enclosed in the December 6,1988 submittal (References 2-4).
The January 31, 1989 submittal provided a non-proprietary version of a j
report previously submitted and did not alter the action noticed, or affect the initial determination published, in the Federal Register on February 8, 1989.
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This reload has the following features:
(a) it will make the reactor core
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the first total 8x8 Advanced Nuclear Fuels (ANF) fuel core, (b) the fresh fuel is designed for higher discharge exposures than the previous reloads, thur it has higher enrichment and hi inclades four lead test assemblies (gher gadolinia loading, and (c) it LTA) of ANF 9x9-5 fuel. The reload methodology is based on the approved ANF topical reports XN-NF-80-19(A),
Volumes 1-3 (References 9, 10, 13 and 14). Additional information has been submitted in plant specific reports.
The scope of the proposed TS changes includes:
1.
Addition of the maximum axial planar linear heat generation rate'(MAPLHGR) curve for the new 8x8 fuel.
2.
Revis tor: of the single loop MAPLHGR curve.
3.
Change of flow and power dependent thennal limits for off-rated condition transients.
i 4.
Addition of linear heat generation rate (LHGR) and MAPLHGR curves for the 9x9-5 LTAs.
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The specific TS changes are as follows:
1.
\\ Bases 2.1.1 and 2.1.2 - Editorial changes to substitue the name
" Advanced Nuclear Fuels Corporation" for " Exxon Nuclear Company."
2.
TS 3/4.2.1 - Revisions to delete references to the replaced fuel, to reference MAPLHGR curves and information for the new fuel and the 9x9-51LTAs.
A MAPLHGR curve for the new 8x8 and the 9x9-5 fuel has been added. The single loop operation limit is added for the new 8x8 fuel and the 9x9-5 lead assemblies.
Figures 3.2.1-1, 3.2.1-2, 3.2.3-1, 3.2.3-2 and 3.2.4-1 reflect these changes.
3.
TS 3/4.2.4 - Editorial changes to delete references to GE fuel which is no longer present in Cycle 4.
i 4.
Bases 3/4.2.1 - References, text revisions and discussion were added to cover the revised MAPLHGR single loop operation.
Description of flow runout was deleted. Text revision for MAPFAC was made.
p 5.
Bases 3/4.2.3 - The transient evaluation description for the GE core has been eliminated. The discussion of MCPR for the loop manual and non-loop manual operaticn has been revis,d.
The cycle specific e
NCPR B3/S.discussionisdeleted. An editorial change was made to Figure 2.3-1.
6.
TS 5.3.1 - Editorial changes were made to generalize the description l
for all Cycle 4 fuel types.
l 2.0 EyALUATION
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2.1 Relo,a,d Description The Cycle 4 core will consist of 800 8x8 fuel assemblies of all ANF manufacture. Ofthese,236aretwiceburned(firstreload),288areonce l
burned (second reload), 272 are fresh (third reload) and 4 are 9x9-5 LTAs.
The core arrangement is the conventional scatter load with the
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lowest reactivity bundles placed in the peripheral regions of the core.
l The loading pattern is designed to maximize the cycle energy while i
minimizing the peaking factors. The Cycle 4 core is, estimated to provide i
1,698 gigawatt days (GWd) of energy compared to an estimatt:d 1,455'GWd for Cycle 3.
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s 2.2 Fuel Design The mechanical design of the fuel for the entire. Cycle 4 loading is de. scribed in XN-NF-85-67(P)(A), Revision 1 (Reference 5).
The 8x8 ANF fuel assembly contains 62 fuel rods and 2 water rods. The fuel rods are
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pressurized and use a diametral pellet to clad gap, which is smaller on
.the interior high enrichment rods than on the remaining rods in the j
bundle, to improve ECCS MARGIN. The scope of the mechanical design analyses included: cladding steady-state strain, transient stresses, j
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fatigue damage, creep collapse, corrosion buildup, hydrogen absorption, fuel rod maximum internal pressure, differential fuel rod growth, creep l
bow and the grid spacer spring design. All parameters meet their l
respective design limits for a batch average burnup cf 34,000 megaw'att days per metric ton of uranium (Mid/MTU).
This average burnu) is about l
4,000 mwd /MTU higher than the Cycle 3 burnup. The peak assem)ly exposure l
is 39,000 mwd /MTV (Reference 6).
The mechanical fuel design is essentially the same as the generic ANF design; thus, the majority of the specific features are covered by generic i
mechanical design reports, The few analyses which have been extended use approved methodology.-
The mechanical response of the fuel assembly during loss of coolant accidents (LOCA) or seismic events is the same as the response of a GE assembly, because the physical properties and bundle natural frequencies are similar. The seismic-LOCA analyses for the GE' fuel showed that the resultant loading would not exceed the fuel design limits (Reference 7).
l Seismic-LOCA analyses for the ANF fuel showed large design margins compared to the GE fuel (Reference 8). Because of the similarity of the I
fuel types and the large margin calculated for the ANF fuel, we find the mechanical design of the ANF fuel assemblies'to be acceptable.
2.3 Thermal Hydraulic Design Analysis The methodology used for the thermal hydraulic design analysis is described in the /iNF approved topical report XN-NF-80-19(A), Volume 3, Revision 2 (Reference 9). The thermal hydraulic design criteria, which were used in the determination of the fuel cladding integrity safety l
limits and bypass flow, were defined as described in the approved ANF l
topical report XN NF-80-19(A),' Volume 4, Revision 1, (Reference 10).
i The uncertainties used in the minimum critical power ratio (MCPR) safety l
limit calculation are provided in the approved topical report SN-NF-524(A)
(ReferenceII). The specific inputs for GGNS-1 and the results of the calculations for Cycle 4 are giventin Reference 3.
i The operating limit MCPR (OLMCPR) values are determined by the limiting transients.
To confirm, and if needed revise, the thennal limits for the all-ANF Cycle 4 core, the following transients were analyzed: load rejection without bypass (LRNB), feedwater controller failure (FWCF) and loss of feedwater heating (LFWH).
In addition, it was established that the generic analysis for the control rod withdrawal error is applicable to GGNS-1, Cycle 4.
All these cases have shown that Cycle 4 is less:
restrictive than Cycle 3. _ Therefore, it is reasonable to conclude that t.he less restrictive transients will continue to be protected. The LFWH transient analysis covered the conditions from beginning to end of cycle l
and the maxim m extended operating domain (ME0D)..The calculation assumed L
a conservative reduction of 100*F.in the feedwater temperature. The l
results showed that the LFWH OLMCPR for all operating conditions'is 1.17.
The LRNB event is the most limiting pressurization transient.
In this transient, the load rejection causes fast closure of the turbine i
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valves resulting in a vessel compression wave and reactor scram.
In the analysis, condenser bypass is'not allowed.. The power spike due to void collapse is terminated by the scram and the recirculation pump.
The maximum steam dome pressure is 1,280 psig, which is less than 1,325 psig j
the required limit.
A flow transient is 'also analyzed to determine the flow dependent thermal
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limits MCPR and the MAPFAC The transients analyzed assume failure of:
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therecircufatingcontrolsfs. tem which results in.a flow increase equal to i
the maximum physically attainable flow.
Two operational modes were 4
assumed, i.e., flow excursion of one pump (designated loop. manual) and of bothpumps(designatednon-loopmanual). For both events, the recirculation system capacity was set at 110% of rated.
For both cases, calculations
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show that the MCFfe and MAPFAC, values are conservative compared to the t
Cycle 4operatingf1mitsthathavebeenestablishedforCycle4(Reference' 4
4). Core flows for the one loop were iniHalized from 30% to 100% of 1
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rated flow and for seven burnup values M ugh Cycle 4.
The MCPR safety I
limit for all types of: fuel in Cycle 4 remain at 1.06 (References 10 and i
11). However, the licensee decided to retain the existing 1.18 as the MCPR operating limit.
The previous most limiting value was 1.17.
Finally, the power dependent MCPR and MAPFAC for off-rated condition l
operationduringanticipatedoperEtionaloccuErenceshasbeendetermined 1
by adding the delta-CPR for the limiting event to the calculated safety limit MCPR. The MAPFAC is'used to protect against fuel melting and excessive clad strain by, setting conservative LHGR limits consistent with the MAPLHGR and consideration of the maximum local peaking factors.
1 The results showed that above 40% power the Cycle 3 MCPR bounds the Cycle l
values are conservative with respect to the $ed.
4 results.
Therefore, the existing limit remains unchan Similarly
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the MAPFAC existing values (Referencep3) and require no change.
In summary, the themal hydraulic design analysis has been performed with approved methods and conservative data. The resulting proposed TS power distribution limits are either within the limits of existing analyses 'or within the operating limits set for Cycle 4 and, therefore, are acceptable.
j 2.4 Core Stability Analysis The 8x8-2 ANF fuel assemblies used in Cycle 4 are hydrodynamically similar to the previously used GE fuel assemblies.
For Cycle 4, all fuel assemblies '
will be ANF assemblies, whereas for Cycle 3 about one. third of the core-was loaded with GE fuel assemblies.
The licensee's analyses performed for
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Cycle 4 confirmed that stability analyses and tests performed for Cycle 3' i
are applicable to Cycle 4.
The four 9x9-5 ANF lead test assemblies in the.
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Cycle 4 core will be placed in low-power regions of the core.
In this
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position the probability of channel instability in these assemblies is minimal.
The overall stability characteristics-of.the core will,- in any.
case, be determined by the 8x8 fuel and not by these four. assemblies.
1 We conclude that the core stability. analysis for Cycle. 4 is. acceptable, j
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2.5 Nuclear Design Analysis The methodology used for the nuclear design and analysis is contained in the NRC approved topical reports XN-NF-80-19A, Volume 1, and Supplements 1 and 2 (Reference 13). The core description and the results of the core reactivity characteristics are given in ANF-88-149 (Reference 2). The results are within the range of those usually encountered in BWR reloads.
In particular, the shutdown margin is 1.094% delta-k/k at the beginning of the fuel cycle.
This value is the minimum value bec6use the reactivity defect is (0.0% delta-k/k) by a large margin.
The shutdown margin is greater than the reqdired value in the TS (0.38% delta-k/k) by a large margin.
Similarly, the standby liquid control system reactivity at 660 ppm boron concentration, for cold xenon free condition is k effective equals 0.96215, which provides adequate shutdown margin.
The end of Cycle 4 core exposure is estimated to be 22,308 mwd /MTV with a maximum value of 23,130 mwd /MTU.
The nuclear design analysis was performed with previously approved methods and the results fall within expected ranges and with i
I adequate margin.
Therefore, we find the nuclear design acceptable.
l 2.6 Transient and Ac'cident Analyses For Cycle 4 the most limiting anticipated operational occurrences are:
load rejection without bypass, feedwater controller failure and loss of l
feedwater heating. Our discussion and evaluation of these occurrences is provided in Section 2.3 above.
A fuel loading error analysis has been performed for Cycle 4 using the l
l methodology described in the approved topical report XN-NF-80-19(A) l (References 9, 10 and 13). The results of the analysis show that the maximum linear heat generation rate (LHGR) is less than the TS limit for l
Cycle 4 and the MCPR is 1.17; i.e., the same limiting vclue as for the l
limiting LFWH transient. Therefore, we find the fuel loading error analysis results are acceptable.
The control rod withdrawal error was analyzed generically (Reference 12) and found to be applicable for Cycle 4.
Finally, the full ANF core was analyzed for reduced flow and power operation to establish MCPR[c MCPR,
MApFAC and MAPFAC limits.
These limits were established in C P
and3$ndhavebeeRrevisedforcycle4.
we find these limits, as 1
l proposed in the TS changes, to be acceptable.
i To support the Cycle 4 operation, the results of LOCA and rod drop r
accident analyses were provided.
The LOCA methodology is based on approved ANF topicals in References 14-16. The analysis confirmed that the peak cladding temperature remains below the 2,200*F limit of 10 CFR 50.46 for all types of fuel present in Cycle 4.
Similarly, the local zirconium-water reaction remains below 17% and the core wide hydrogen production below 1.0%, the required 10 CFR 50.46 limits (Reference 2).
L Accordingly, the analyses are acceptable.
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6 The rod drop accident was analyzed using ANF's generic parametric methodology for the fuel enthalpy rise during a postulated rod drop at.cident (References 9, 10, and 13).
The results listed in Reference 2 show that the maximum deposited fuel rod enthalpy is 172 cal /gm, which is much lower than the required limit of 280 cal /gm and are acceptable.
In summary, we conclude that the transient and accident analyses were i
performed with approved methodology,-the results are within acceptable limits and are acceptable.
1 2.7 9_x9-5 Lead Test Assemblies As mentioned previously Cycle 4 includes four ANF 9x9-5 lead test I
assemblies (LTA).
The LTA are to be placed in low-power locations in the core and are designed to have improved thermal performance.
Therefore, the LTA have ample margin to the operating limits.
The licensee analyzed the performance of the LTA.
The analysis methodology used was the same as for the 8x8 fuel assemblies (References 9, 10, and 13).
The licensee's analysis confirmed that the LTA mechanical design meets the no-centerline melting and the 1.0% clad strain criteria.
The analysis also determined that the 9x9-5 LTAs are hydraulically compatible with the 8x8 regular assemblies over the full range of the expected operating conditions. Analyses showed that no reduction in thermal margin will take place.
The nuclear design of the 9x9-5 assemblies is similar to the 8x8-2 assemblies and, therefore, has similar reactivity characteristics. The transient and accidant analyses, the shutdown margin,'the liquid boron control and the LTA loading error analyses were explicitly modeled for the four LTAs and demonstrate that their power is conservatively predicted.
Because the LTA will be placed in a low-power region, the licensee's analysis showed that during anticipated operational occurrences the bundle power will be lower than that required to reach transition boiling.
Therefore, we agree that operational limits are adequate for the LTA.
LOCA analyses for the LTA demonstrated that these assemblies perform better than the 8x8-2 assemblies, thus, meeting the 10 CFR 50.46 limits by alargermargin(Reference 17). The consequences of the rod drop accident are governed by the rod worth.
In the vicinity of the LTA, the reactivity J
is not different from regions loaded with all 8x8-2 assemblies. Therefore, i
the results are similar to the previously analyzed regions and within the i
required limits.
In summary, approved analysis methods for the LTA performance in Cycle 4 showed that the LTA meet the operational and accident limits.
Thi-s is i
because the LTA are located in positions of low power and because they are neutronically and hydraulically similar to the 8x8-2 assemblies.
Therefore.
we find the use of the four 9x9-5 LTAs in Cycle 4 is acceptable.
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2.8 Fuel Handling Accident The licensee has requested authorization to allow fuel burnup to 39,000 megawatt l days per metric' ton (MWD /MT) from 33,000 MWD /HT.
The staff evaluated the potential impact of burnup up to 60,000 MWD /MT on the radiological assessment of design basis accidents (DBA), which were previously analyzed in the licensing of GGNS-1.
The staff reviewed the licensee's submittils and also reviewed a pub-lication, " Assessment of the Use of Extended Burnup Fuel in Light-Water Reactors," NUREG/CR-5009, February 1988, prepared for the NRC. The NRC contractor, the Pacific Northwest Laboratory (PNL) of Battelle Memorial Institute, examined the changes to NRC DBA assumptions (described in the various appropriate SRP sections and/or Regulatory Guides) that could result from the use of extended burnup fuel (up to 60,000 MWD /MT).
The contractor concluded, and the NRC staff agrees, that the only DBA that could be affected by the use of extended burnup fuel, even in a' minor way, would be the potential thyroid doses that could result from a fuel handling i
accident.
PNL estimates that I-131 fuel gap activity in the peak fuel' rod with 60,000 MWD /MTU burnup could be as'high as 12%. This valve is 20%
higher than the 10% value normally used by the staff in evaluating fuel handling accidents (Regulatcry Guide 1.25, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling _and Pressurized Water Reactors").
The staff, therefore, reevaluated the fuel handling accidents for GGNS-1 with an increase in iodine gap activity in the fuel damaged in a fuel handling accident. The maximum thyroid dose from a fuel handling accident within the secondary containment as shown in the operating licensing Safety Evaluation Report dated September 1981 is 2.3 rem at the exclusion area boundary. The recalculated thyroid dose (increased by 20%) possible with extended burnup fuel of 60,000 MWD /MT is 2.8 rem.
We conclude that the only potential increased dose potentially resulting from DBA with extended fuel burnup to 60,000 MWD /MT is the thyroid dose resulting from fuel handling accidents.
This small calculated increase is insignificant, in that theue' doses remain well within the 300 rem thyroid exposure guideline value of 10 CFR Part 100.
Therefore, the requested extended burnup to 39,000 MWD /MT is acceptable.
2.9 S_umary We have reviewed the material submitted by Systems Energy Resources, Inc.,
for the GGNS-1 Cycle 4 operation. Based on this review, we conclude that the fuel design analysis, the thennal hydraulic design analyses and the transient and accident analyses are acceptable. The TS changes requested for this reload (listed in Section 1.0 above) reflect the necessary modifi-cations for the operation of Cycle 4 and are acceptable.
3.0. ENVIRONMENTAL CONSIDERATION Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding'of no significant impact have been. prepared and published in the on February 13, 1989 in the Federal Register (53 FR 6629). Accordingly.
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based upon the environmental assessment, the Comission his determined I
that the. issuance of this amendment will not have a significant effect on the quality of the human environment, j
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4.0 CONCLUSION
The Comission made a proposed determination that this amendment involves-1 no significant hazards consideration, which was published in the Federal Register (54 FR 6196) on February 8,1989, and consulted with-the state of Mississippi.
No public consnents or requests for hearing were 1
received, and the State of Mississippi did not have any'coments.-
1 The staff has concluded, based on the considerations discussed above.
that:
(1) there is reasonable assurance that the health and safety of_ the ublic will not be endangered by operation in the proposed manner, and -
p(2) such activities will be conducted in compliance with the Comission's l
4 regulations and the issuance of this amendment will not be inimical to the q
comon defense and the security, or to the health and safety of the public.
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Principal Contributors:
'L. Lois Reactor Systems Branch, DEST j
L. Kintner, Project Directorate II-I, DRPR a
l Dated:
I REFERENCES l
1.
Letter from W. T. Cottle, System Energy Resources, Inc., to USNRC, " Grand Gulf Nuclear Station, Unit 1. Proposed Amendment to the Operating l
License " dated December 6, 1988.
2.
ANF-88-149, " Grand Gulf Unit 1, Cycle 4, Reload Analysis" ANF, dated November 11, 1988.
(Enclosure to Reference 1) 3.
ANF-88-150, " Grand Gulf Unit 1, Cycle 4 Reload Analysis" ANF, dated November 11, 1988.
(Enclosure to Reference 1) 4.
NESDQ-88-003 Rev. O, " Grand Gulf N'.: clear Station, Unit 1 Revised Flow Dependent Thermal Limits" MSU Systems Services Inc., dated November 1988.
(EnclosuretoReference1) 5.
XN-NF-85-67(P)(A), Rev.1 " Generic Mechanical Des"n for Exxon Nuclear det Pump BWR Reload Fuel," Exxon' Nuclear Co., datec eptember 1986.
6.
ANF-88-183(P), " Grand Gulf Unit 1, Reload XN-1,3 Cycle 4 Mechanical Design" ANF Corporation, November 1988.
7.
" Grand Gulf Nuclear Station, Maximum Extended Operating Domain Analysis" General' Electric Company, dated March 1986.-
8.
XN-NF-81-51(A), "LOCA-Seismic Structural Response of an ENC BWR Jet Pump Fuel Assembly" Exxon Nuclear Company, dated May 1986.
L
9
- 9.
XN-NF-80-(A), Vol. 3 Revision 2. " Exxon Nuclear Methodology for Boiling Water Reactors THERMEX:
Thermal Limits Methodology" Exxon Nuclear Company, dated January 1987.
l10.
XH-NF-80-19(A), Vol. 4, Revision 1, " Exxon Nuclear Methodology for l
Boiling Water Reactors: Application of the ENC Methodology to BWR l
Reloaas" Exxon Nuclear Company, dated June 1985.
l 11 XN-f4F-524(A), Revision 1 " Exxon Nuclear Critical Power Methodology for Boiling Water Reactors," Exxon Nuclear Company, dated November 1983, 12.
XN-NF-825A, Supplement 2 "BWR/6 Generic Rod Withdrawal Error Analysis, t
MCPR for Plant Operations Within the Extended Operation Do:nain," Exxon Nuc18arCompany,Inc.,datedOctober1986, 13.
XN-NF-80-19PA, Volume 1, Supplements 1 and 2, " Exxon Nuclear Methodology for Boiling Water Reactors: Neutronic Methods for Design and Analysis,"
Exxon Nuclear Company, dated March 1983.
- 14. XN-NF-80-19(A), Vols., 2, 2A, 28, & 2C, " Exxon Nuclear Methodology for Boiling Water Reactors: EXEM BWR ECCS Evaluation Model,," Exxon Nuclear Company, dated September 1982.
- 15. XN-NF-CC-33(A), Rev.1, "HYXY: A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option," Exxon Nuclear Company, dated November 1975.
16.
XN-NF-82-07(A), Rev.1, " Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, dated November 1982.