ML17194A667
| ML17194A667 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 04/29/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17194A668 | List: |
| References | |
| NUDOCS 8205130074 | |
| Download: ML17194A667 (22) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 63 TO FACILITY OPERATING LICENSE NO. DPR-25 COMMONWEALTH EDISON COMPANY DRESDEN NUCLEAR POWER STATION UNIT 3 DOCKET NO. 50-249
. - Authors:
J. Holonich S. Wu
- L. Lois R. Audette I.. Introducti.on By letter dated January 11, 1982, as supplemented by letters dated January 21, February 23, and March 22 and 29~ 1982, Commonwealth Edison Company (licensee) proposed changes to the Technical Specifications (TS}
for Dresden Nuclear Power Station Unit 3 to support operation using reload fuel supplied'by, and the associated analyses perfoniled by the Exxon Nuclear Company (ENC or Exxon}.
Previous plant operation utilized fuel and analyses provided by the General Electric Company (GE).
In support of the Cycle 8 reload, the 1 icensee sutmitted plant-speCific reports which describe the steady-state reload analysis {Reference 17), the pla~t tr~nsient_analysi~ (References 18 and 19), and ~he lri~~-of-coolant an~lysis (Reference 16). This Safety Evaluation addresses the acceptability of those reports in support of the licensee's proposed TS changes.
II. Background Dresden 3 Cycle 8 operation will represent the first reload of a jet pump BWR using fuel supplied by ENC.
The reload, designated as XN-1, will consist of 224 reload fuel assemblies fabricated by ENC and identified as type XN802.69... 5.
During Cycle 8 operation the ENC bundles will be placed with
- the 500 GE fuel assemblies presently in the core.
III. Evaluation We have evaluated the infonnation in the plant-specific reports sutmitted by the licensee in support of Cycle 8 operation describing the steady-state reload analysis (Reference 17), plant transient analysis (Reference 18), and LOCA analaysis (Reference 16).
The acceptance criteria used during our evaluation consisted of the following:
Sections 4.2, 4.3 and 4.4 of the Standard Review Plan (NUREG-0800} for our evaluation of the steady-state reload analysis, Section 15 of the Standard Review.Plan for our evaluation of the plant transient analysis, and 10 CFR Part 50.46 and Appendix K to 10 CFR Part 50 for our evaluation of the LOCA analysis.
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- 1. 0 FUEL DESIGN 1.1 Mechanical Design The Cycle 8 core is composed of 224 Exxon XN-1 8x8,. -200 GE P8x8R, and 300..
GE 8x8 fuel assemblies.
The GE fuel types were approved for operation in previous cycles.
The Exxon XN-1 8x8 fuel design in the Dresden 3 Cycle 8 Reload Report, XN-NF-81-76, (Reference 17), is referred to as ~he generic jet pump BWR fuel design and is described in detail in XN-NF-81-21 (P)
(Reference 15).
This report (Reference 15) has been reviewed and approved by the staff and the associated Safety Evaluation documenting our findings
... will be issued under separate cover.
Table 4.1 of XN-NF-81-76 shows the pertinent data for the XN-1 fuel.
Each Exxon assembly contains five burnable poison rods of 3.0 weight percent GD203 blended with 2.25 weight percent U-235 to reduce the initial reactivity~
The Exxon XN-Lfuels are designed to fit into the existing GE channel boxes.
Amore detailed description can be found in Table 4.Tand the design drawings of XN-NF-81-21(P).
On the basis that the generic jet pump BWR fuel design described in Reference 15 has. been accepted by the staff, we approve the use of Exxon XN-1 fuel for Cycle 8 reload.
Thus with respect to mechanical design, we find Exxon XN-1 8x8.fuel for Dresden 3, Cycle 8 r~load is acceptable.
1.2 Rod Pressure The Exxon design. calls for no rod interna.1 pressure exceeding the system pre$SUre during nonnal operation.
Since Exxon uses the RODEX2 code (Reference 10), which has not yet been approved by the staff, to demonstrate compliance with the design basis, the NRC staff has requested Exxon to redo the analysis by using the approved GAPEXX code (Reference 28).
An Exxon reanalysis (Reference 43) for a burnup up to 10,000 MWD/MTU shows that the internal rod pressure will not.exceed the system pressure.
Because the peak assembly-averaged. burnup of Exxon fuel is expected to be 12,000 MWD/MTU at the end of Cycle 8, additional GAPEXX calculations will *be required for rod pressure after 10,000 MWD/MTU if RODEX2 is not approved at that time.
Accordingly, we have modified the appropriate Technical Specification to require the licensee to submi.t additional analyses for NRC approval using an approved code if use of ROD~X 2-code has not been approved prior to exceeding 10,000 MWD/MTU burnup.
Based on the.lice!'ls_e.e's an-alyses. -
using an approved code for the first 10,000 MWD/MTU burnup, we find the fuel design with respect to rod pressure to be acceptable.
- l. 3 Fuel Centerline Melting The design basis for Exxon fuel center:..line temperature is.~hat no_f1Jel m{;!lting should result from nonnal operation including transient* occurrences.
Exxon has generated a fuel temperature history for the Dresden 3 reactor using RODEX2 at a. peak rod power of 13.93 kW/ft.
Fuel temperatures under this condition envelope maximwn temperatures expected at any exposure for XN-1 fuel
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The. maximum calculated temperature is 3909 °F and occurs at a rod exposure of 21,200 MWD/MTU~ -. By raising the power to 120% overpower
{16.72 kW/ft) at this exposure to. simulate a limiting transient condition, RODEX2
_ gives a fuel centerline temperature of 4607 OF.,i:The fuel melting tenperature at 21,200 MWD/MTU according to an acceptable correlation is 4959 Of. *Thus a significant margin to centerline melt exists for 120% overpower.
Inasmuch as the RODEX2 is unapproved yet, and the limiting* condition occurs at 2J,200 MWD/MTU, which exceeds the Cycle 8 burnup for Exxon fuel, we conclude.
that reasonable assurance has. been provided from GAPEXX calculations that the melting of Exxon XN-1 fuel will not occur during normal operation and transient events for burnups up to 10,000 MWD/MTU. *However, additional calculatfons will
- be needed for center 1 i ne me 1t i ng after 1 0, 000 MWD/MTU if: RODEX2 is not approved at that time.
Accordingly, we have modified the appropriate Technical Specifi-cations to require the licensee to sul:mit additional* analyses for NRC approval using an approved code if the use of RODEX2 has not been approved prior to exceeding 10,000 MWD/MTU burnup.. Based on the use of an approved. code for the first 10,000 MWD/MTU of burnup, we. find the fuel design with respect to fuel centerline melting to be acceptable.
J.4 Cladding Swell.ing and Rupture We have been evaluating on a generic basis three fuel material models that are used in the ECCS analysis.
Those models predict cladding rupture temperature, cladding burst strain (ballooning.), and fuel assembly flow blockage {used only in PWR analyses).
We have discussed its evaluation with
.vendors and other industry representatives {Reference 31), published NUREG-0630 (Reference 32), and required. licensees to confinn that their operating reactors
- c*\\\\'aiild'continue to be in*confonnancewith the ECCS Acceptance Criteria of 10 CFR
- Part 50.46 if the NUREG-0630 correlations were substituted for the present materials models in their ECCS evaluations and certain other compensatory model*
changes were allowed (References 33 and 34) to offset any*penalties incurred due to the use of the NUREG-0630 correlations.
Since.the Cycle 8 reload fuel for Dresden 3 is supplied by Exxon instead of GE, the original NSSS vendor,_ an ECCS reanalysis (Reference 35); was perfonned **
an~ in acco~ance with the requirenents discussed in the preceeding paragraph. __ _
Th1s*analys1s was augmented by a supplenental ECCS assessment (Reference 45}
that addressed the pred.icted effect of NUREG-0630 correlations on the Dresden analysis.
For the Dresden specific range of conditions ( i ~ e., cladding temperature - ramp cladding hoop stress, and cladding rupture tenperature}, Exxon found that the ENC model (Reference 14) either agrees with or is conservative with respect. to NUREG-0630 correlations.
We thus conclude that the inclusion of the NUREG-0630 correlations into the Dresden ECCS analysis would not significantly reduce the presently predicted
- margins to the ECCS Acceptance Criteria and that the licensee's design with respect to cladding swelling and rupture is, therefore, acceptable.
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- 1.5 Conclusion We have reviewed the Dresden 3 fuel design, including the Exxon XN-1 8x8 fuel, and analyses for the Cycle 8 reload.
On the basis of analyses using the approved GAPEXX code and the approved generic report, XN-NF.:81-~l (P),
we conclude the application is acceptable as submitted. However, the licensee will be required to submit additional calculations for NRC approval of rod pressure and fuel centerline temperature with acceptable methods prior to reaching 10,000 MWD/MTU unless the use of RODEX2 code is approved without modifications.
The appropriate Technical Specifications have been modified to reflect this requirement *
... 2.0 Thennal-Hydraul ic Design The review of the thermal-hydraulic aspects of the Cycle 8 reload consisted of the foll owing: (a) the compatibility of the ENC and GE fuel bundles; (bl the fuel cladding integrity safety limit; (c) the operating safety limit minimum critical power ratio (OLMCPR); (d) the amount of bypass flow associated with
- the different fuel designs; (e) thennal-hydraulic stability; and (g} the:
proposed technical specifications.
The objective of the review was to confirm that the thermal-hydraulic design of the reload core was accomplished using acceptable analytical methods, provided an acceptable margin of safety from conditions which would lead to fuel damage during normal operation and anticipated operational occurrences and is not susceptible to thennal-hydraulic instability.
2.1 Hydraulic Compatibility
- Since a BWR core*is a series of parallel flow channels connected to a common lower and upper plenum, the total*prE:ssure drop across the bundles will be equal.
However, differences in the hydraulic res.istances of the fuel designs may cause variations in axial pressure drop profiles across the bundles.
- The 1 icensee provided (Reference 43)' a figure of pressure drop versus axial length for both fuel types.
The calculation of these pressure drops was perfonned using the methodology documented in Reference 2 and hydraulic resistance factors obtained from single-phase flow tests conducted on the fuel bundles.
The results of
- these* analyses showed that the difference -in local pressure drop is the AP across the lower tie plates for the two fuel designs. wliich results in a different total flow for the different assembly types.
The APs across the rodded region are of the same magnitude. and the total pressure drop equalizes once the upper tie plate is accounted for in the calculations.
Additional analyses of the effects of hydraulic compatibility on thennal margin were presented in the reload report (Reference 17). The results of these analyses showed that the Cycle 8 reload had a mor:e. limiting core config~ration than previous.
reloads and that the most limiting assembly for Cycle 8 in tenns of CPR is the GE 8x8Rdesign.
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Based on our review of the information presented in the Cycle 8 reload report
. and the additional information sut:mitted in Reference 43, we conclude that the GE and Exxon fuel types *have similar hydraulic resistances and are therefore, hydraulically compatible.
2.2 Minimum and Operating Limit Critical Power Ratio The minimum. critical power ratfo (MCPR) for the Cycle 8 reload was determined by the 1 icensee to be 1.05 for all fuel types using the XN critical power correlation described in Reference 1.2 and the methodology described in Reference
- 13. This 1.05 MCPR is established on a plant-specific cycle-specific basis such that 99.9% of the fuel rods are expected to avoid boiling transition.
We asked ENC to explain why this li~it of 1.05 was less than the 1.07 and 1.06 values reported for* GE 8x8 and GE 8x8R fuel designs, respectively, in the previous reload application.
Exxon sutmitted information. (Reference 44) to show that the methodology used to determine the nuclear uncertainties and the resultant code uncertainties account.for the increase in thermal margin.
The smaller uncertainties are due to the use of a wider data base in the ENC analyses. This larger data base results in less scatter of the data about the mean and thus less uncertainty.. We have reviewed the Exxon response and have determined that the reduced nuclear uncertainties do result in an increased thermal.margin; however, when coupled with the uncertainty associated with the XN-3 correlation a net increase of only 0.01.in MCPR is justified. Hence, the MCPR for the GE 8x8R fuel will decrease to 1.06 rather than 1.05.
The additional margin of 0.01 does result in a MCPR of 1.05 for the GE 8x8 fuel design *
. The methodology for determining the nuclear uncertainties and the uncertainties
- thenselves are reported in References 3, 12, and 13.
We have reviewed and approved the above methodologies.
The Safety Evaluations documenting_ our acceptance are. being issued under separate cover. Based on these facts,
- we conclude that a MCPR of 1.05 is acceptable for the ENC and GE 8x8 fuel designs; however, the MCPR for the GE 8x8R fuel will be 1.06.
Various transients could reduce the MCPR below the intended safety limit. The most limiting of these operational transients_ have been analyzed by the 1 icensee to determine which event could potentially induce the largest reduction in the*initial critical power ratio (~CPR). Table 1 contains the results of these analyses. The transient which resulted in the. greatest -~CPR was the load rejection. without bypass.
- The ~CPR. for the load rejection. without bypass was c~lculated using the statistical methodology described in XN-NF-81-22(P}, (Reference 9}.
Based on this analysis the licensee has proposed a.~CPR of 0.25 at a. 95% probability level.
The results of the transient analyses are discussed in XN-NF-81-78.
We have reviewed and approved XN-NF-81-78 in Section 5.0 of this SER; therefor~, the response surface presented is valid for Cycle 8 operation. Licensee submittals in support of Cycle.9 operation must contain a new response. surface or justification of why the present response surface is valid.
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In detennining. the operating 1 imit critical power ratios. (OLMCPR) for Cycle 8 operation, biases.assoctaied with the ENC plant transient code were not considered. Since these biases could affect the probability of exceeding the thenual-hydraulic design basis.
we asked Exxon to show how:-these biases were accounted for in the OLMCPR calculations.
In response (Reference 44),
calculations were perfonned to evaluate the effect of variations in steam line dynamics, core average gap conductance, the length to area ratio of the steam separator and dryer, the steam relief valves' perfonnance, and turbine control valve full stroke.
The *calculations were perfonned by taking the nominal plus 2a*values for each of the predictor variables used in the response surface methodology and the licensing basis calculational values of the five parameters discussed above.
The results of this caleulation yielded a ACPR of 0.219.
Exxon then changed the value of each of the response surface predictor variables to its nominal.value and the values of the.five parameters listed above to its best-estimate value~ The calculation was repeated using these best-estimate values and. the resultant ACPR.was detennined to be 0.175.
Table 2 contains the results of the ENC ACPR sensitivity study and the list of parameters that were varied. The &CPR propo_sed for the Dresden 3 Cycle 8 reload is 0.25.
Although the ENC sensitivity study {Reference 44) demonstrates that there is conservatism in its licensing basis calculation, we are not convinced that this conservatism sufficiently accounts for uncertainties in the Exxon plant transient code. While calculational comparisons of the ENC plant transient code and the Peach Bottom test results are generally. conservative, the data base is too
- 1 imited to assure that the code uncerta int.ies do not exceed the conservatism a.chiev.ed through calculational methods *. We are presently reviewing the uncertain-ties associated with the ENC plant transient code on a generic basis in conjunction with our evaluation of XN-NF-79-71 and XN-NF-81-22.
Until we* complete. our generic review of XN-NF-79-71 and XN-NF-81-22 we require that code uncertainties be accounted for using the method~ discussed in our safety evaluation report on the GE ODYN computer code (Reference 45). This methodology is as follows:
where;
. ({ACP~/ICPRc) + (ACPRu/ICPRu)J (ACPRc)
(A CPRc/ I CPRc)
&CPRc = ENC Calculated ACPR = 0. 25 for Cycle 8 ICPRc = Initial CPR or OLMCPR = 1. 30 for ENC and GE 8x8 fuel
= 1.31 for GE 8x8R fuel
&CPRu/lCPRu =Value associated with*code uncertainties.
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7 As part of our evaluation of the ODYN code, we perfonned an independent sensitivity analysis of the input parameters and detennined that the value of t.CPRu/ICPRu was 0.044. Since this study was perfonned using an independent code, this: audit is also applicable to the ENC code.
The analysis was perfonned by perturbing an.*input parameter and calcula.ting the change 1n t.CPR/ICPR due to this perturbation. Two of the.input parameters that were varied in our study were the void coefficient and the scram reactivity parameters, which comprise a large fraction of the total code uncertainty of* 0.044.
Exxon *has considered the uncertainties as~ociated with void reactivity' and rod worth since these were two of the predictor variables used in the construction of the:*response surface.
In addition,: Exxon intends to statistically convolute the code uncertainty by. treating it as one of the response surface variables. The
- using the detenninist.ic method desert bed above.
We anticipate completing our generic review of the ENC code uncertainties prior to the end of Cycle 8 operation, which is the most limiting operational state.
Taking this into consideration and based on our review of the above. infonnation, we have concluded that an ENC code uncertainty value of 0.022 t.CPRu/ICPRu, applied deterministically, should appropriately bound the expected result from our generic review of this topic.
When this value is included in the t.CPR calculations, the results are a t.CPR of 0.28 for all fuel types.
When this t.CPR is added to the MCPRs, the resultant OLMCPRs are 1.33 for the ENC and GE 8x8 fuel designs and a 1.34 for the GE 8x8R fuel. After our review of the code uncertainties in References 2 and 9 is complete, results of that review may be applied to the Dresden core.
Table 3 contains the ENC pro posed OLMCPRs and the staff imposed OLMCPRs for Cycle 8 and those reported for Reload 6..
The staff imposed OLMCPRs account for code uncertainties and will assure that the safety limit MCPR is not violated in the event of any anticipated transients. Therefore, we have modified the proposed TS changes regarding OLMCPRs to incorporate these limits.
Another variable that was statistically convoluted was the scram rod insertion.......
time.
The licensees analyses employed the Dresden 3 known average scram rod insertion times of 2.58 seconds; however, the TS limit is 3.5 seconds. Since the maximum cycle average scram for 90% rod insertion of 3.5 seconds is not bounded by the statistical distribution, the 1 icensee has proposed that an increase in t.CPR be required once the. average measured scram *time exceeds 2.58 _seconds:
The value of 2.58 seconds is a one a bound on a 95/95 mean.
A.95/95 bound on the same distribution is 2.66 seconds, hence 2.58 is conservative.
We agree that an increase in t.CPR is required if the average measured scram time exceeds 2.58 seconds.
The increase in t.CPR is calculated using the following equation; where; d Ct.CPR) = {O. 0544 h - 0.14 5(t.CPR) = The increase in t.CPR required if the average*meas~red scram time exceeds 2. 58 seconds.
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't = *a.verage measured scram insertion time.
We have therefore modified the 1 icensee' s proposed TS changes by incorporating
- the above equation to account fo.r an increase in-average scram rod insertion times
- between that analyzed and the TS 1 imi t of 3. 5 seconds.
Since the*inc~ease in 4CPR using.the method described above will preclude violation of-the safety limit MCPR, we find this approach acceptable.
2.3 Bypass Fl cw The amount of bypass flow calculated for the end of life of Cycle 8 is 10.4%.
This value is consistent with the 10.46% value reported by General Electric in its generic reload document, NEDE-24011, which has been previously approved.
Therefore, we conclude that the amount of bypass flow reported for the Cycle 8
- mixed core is acceptable.
- 2.4 Thennal-Hydraulic Stability The thennal-hydraulic stability of the Cycle 8 core was analyzed using approved methods described in Refer~nce 3.
The ca1culated decay ratio at the natural circulation - 100% rod line intersection.(which is the least stable physical attainable point of operation) is 0.45.
The calculated decay ratio for Cycle 7 was 0.53.
The use of two different calculatfonal methods precludes a direct comparison of the Cycle 7 and Cycle 8 decay ratios. However, the smaller decay ratio reported for Cycle 8 operation may.be partially attributed to the thicker rods.associated with the XN-1 design.
Based on the fact that jet pump BWRs are not. permitted to operate in the natural circulation mode and
- the fact that the calculated decay ratio shows a large margin of stability, we conclude. that the stability analysis for the Cycle 8-core is acceptable
- 2.5 Conclusion We have completed our review of XN-NF-81-76 suanitted in support of Cycle a*
operation of Dresden Unit 3. This review concentrated on comparing the differences in operational limits during Cycle 8 and Cycle 7 and the Exxon XN-1 fuel design.
Based on our review, we.conclude that Cycle 8 operation is acceptable provided that:
and the OLMCPR of 1.34 for the GE 8x8R fuel design are. incorporated
. into the technical specifications *
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(2) *The 11censee will' be required to sul::mit additiona*l calculations for NRC appro.. val of rod pressure an*d fuel centerline temperature with acceptable methods prior to reaching 10,000 MWD/MTU unless the use of RODEX2 code is approved without modifications. * * *
(3)
An increase in the ~CPR be calculated in accordance with the equation specified in section 2.3 of the SER in the event the average measured scram*
rod insertion time exceeds 2.58 seconds.
The licensee has su!:Jnitted proposed changes to the Dresden 3 Technical Specifications. *These changes modify the'.MCPRs and incorporate the ENC curves for detennining the OLMCPR for core flows less than rated. Based on our review of Reference 17, we have detennined that with the incorporation of the modifications stated above,.. the 1 icensee.' s proposed
- TS changes are acceptable. :Although our rev1ew of ENC' s analysis for reduced flow operation (Reference 19) is incomplete, we have detennined that the existing Technical Specifications pertaining to reduced flow operation are still applicable and are conservative. Therefore, the existing TS governing reduced flow operation shall remain in effect pending completion of our review.
We have modified the proposed TS to incorporate the above three modi-fications. Thus, we conclud~ that the licensee's proposed TS chan~es to support Cycle 8 operation are acceptable.
3.0 NUCLEAR CHARACTERISTICS The 224 f;esh 8x8 ENC fuel bundles (type XN802. 69-5) wil f have an enrichment of 2.69 weight percent (w/o) of U-235 with a six inch natural urani~m blanket at
' each end.
The average enrichment of the central region of the fuel bundle is
- 2.87 w/o.
In addition, 72 GE 8x8 assemblies of 2.. 50 w/o enrichment, 228 GE*
8x8 assemblies of 2.62 w/o, and 200 GE 8x8R 2.65 w/o enrichment will be loaded.
There will be no 7x7*bundles in this loading. Approximately 31%
of the fuel assemblies will be fresh fuel, distributed in a 2-out-of-4 scatter p~ttern with the exception of the c9re axes as described in Reference 17.
The effective *multiplication factor keff, will be equal to or greater than O. 90 for nonna 1 storage con di ti ons provided that k... is less than 1. 31. (Reference *20).
Similarly, based on keff less than."95 for the spent fuel pool for._GE fabricated racks when k is less.than 1.31, ENC has calculated that: its fuel is _slightly less reactiv~ than the GE fuel within the applicable exposure and temper~ture
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-* :1 10 range; hence, the technical specification requirements for the storage of fuel at Dresden are met (Reference 20).. For the high density fuel storage, analyses performed by ENC demonstrated that the Nuclear Service Corporation designed racks will meet the keff less than.95 requirement.
The maximum lattice k= in the reactor core 1s 0.244 at peak reactivity as described in Reference 17.
The shutdown margin of the reconstituted core meets the technical specification *
- requirement that the core be at least 0.25% 6k subcritical in the worst reactive condition throughout the cycle when the highest worth control rod is*
- rully withdrawn and all other rods are fully inserted. *For Cycle 8 the licensee cakulated that the keff with the strongest rod out is 0~984 resulting in a shutdown margin of 1.6% 6k {Reference 17). The effect of settling of s4c in.the absorber tubes will not have a significant effect on the shutdown margin.
The standby liquid control sys~em is capable of bringing* the reactor from full power to a subcr:itical condjtion assuming none of the withdrawn control rods are inserted. The 600 ppm boron concentration in the reactor will bring the reactor subcritical to keff = 0.994 at cold xexon free conditions (Reference 17).
The calculated shutdown margin {6k) of the liquid control system is 0.56 which satisfies the required minimum value in the technical specification of 0.03.
Fuel misloading and fuel misorientation analyses were performed using the approved methodology d~scribed in XN-NF-80-19, Volume 1, Supplements 1 and 2.
This-analysis covered both.the ENC XN-1 and all the GE 8x8 fuel type$.
The largest 6CPR calculated was 0~016 for a misloading error. This result is bounded by the result based on the load rejection without bypass, which
- is the limiting transient.
The control rod withdrawal error, i.e., inadvertent withdrawal of a high control blade was calculated using the approved methodology described in XN-NF-80-19, Volume 1, Supplements 1 and 2. 'For a rod block setting of 110% for.Cycle 8
- the greatest 6CPR was 0.15.
(Rod block setting has bee~ increased from the current value of 107%.)
Thi.s condition is also bounded by the load rejection without bypass transient.
- A control rod drop accident analysis was performed using the approved parametric values developed in XN-NF-80-19, Volume 1, Supplements. 1 and 2.
The ca~culated deposited enthalpy of 151 cal/g is well within the allowable
- limit of 280 cal/g as stated in the acceptance _criteria {NUREG-0800, Section 4.2).
Based on our review of the licensee submittal {Refere~ce 24), the plant specific analysis (Reference 17) and the approved topical.report (Reference 3) which dealt with the analysis methodology, we have determined that the nuclear characteristics and the expected performance of the reload core for Cycle.8 are acceptable *
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11 4.0 LOSS OF COOLANT ANALYSIS From an earlier generic BWR-3 analysis (Reference 36) conducted by the Exxon Nuclear Company (ENC), it was found that jet-pump BWRs of the Dresden 3 design would demonstrate a limiting or Design Basis LOCA for a double-ended guillotine break of a recirculation loop pump suction pipe~ To detennine the allowable Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) over the Cycle 8 reload core life in Dresden 3, ENC has conducted a burnup effects study for the Design Basis Accident (OBA) LOCA using its EXEM LOCA analysis model (Reference 16).
The burnup effects analysis consists of detennining the system blowdown and reflood response to the DBA-LOCA break followed by a series of hot channel analyses where varying fuel initial conditions detennined by fuel burnup effects are considered.
The hot channel study uses the. hot channel thenno-hydraul ic boundary conditions detennined from the single DB-LOCA calculations to determine the Peak Clad Temperature (PCT)* and Metal-Water Reaction (MWR) variation with burnup.
Conformance to all of the acceptance criteria for Emergency Core Cooling Systems in 10 CFR 50.46 for ea.ch burnup case serves to estab.lish the allowable MAPLHGR over core life.
- 4. l Evaluation Use of ENC's approved EXEM methods for LOCA analyses provides conformance of the analytical model used for these studies to Appendix K of 10 CFR 50 for ECCS Evaluation Models.
Results of the computed PCT variation for the DBA-LOCA an~lysis over core life is shown on Figure l as an overlay to the MAPLHGR curve presented in Reference 16.
The PCT peaks at 2156°F at a burnup of 20 GWD/MTM due to dimensional conditions in the fuel at this burnup level allowing a 44°F margin to the 10 CFR 50.46 limit. Metal-water reaction also peaks at 4.1% over the 18 to_20 GWD/MTM range of burnup remaining well below the 17% limit in 10 CFR 50.46.
In performing the MAPLHGR analyses for the Dresden 3 Cycle 8 reload core, ENC determined initial fuel conditions for the DBA-LOCA hot channel transients with its newly developed RODEX2 code (Reference 38).
- This code is under review by the staff and is not expected to be fully accepted before restart with the Cycle 8 reload core.
To provide an interim analysis using accepted models, ENC recomputed hot channel transient temperatures for the DBA-LOCA using the originally approved GAPEX code contained in the approved EXEM (Reference 37) model for determining initial fuel conditions as a function of burnup.
A complete reload lifetime calculation was not recomputed but does address the initial 10 GWD/MTM of Cycle 8 life. If RODEX2 is not approved before 10 GWD/MTM burnup.is achieved, additional analyses with GAPEX must be provided to verify the continued acceptability of the proposed MAPLHGR limits.
Accordingly, we have modified the appropriate TS. to require the licensee to submit additional ~nalyses for NRC approval using an approved code if use of RODEX2 code has not been approved prior to exceeding 10. GWD/MTM burnup.
- -:--**-::--_~----,.... - --*-.
.;..... *---*'**-:0.-*-~-.*---***~---~-_:__-~..._. __ *--*--------~--"-*:.......:..*:.. ~ *.....:.~--~.----'--*-** -. ------~---.... *--*- *-----*-'---.
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Results of the GAPEX calculations using the same MAPLHGR conditions assumed for theRODEX2 calculations are also shown on Figure 1.
Comparison of PCT results on this figure for the two fuel codes over the initial 10 GWD/MTM shows the influence of stored energy differences at the start of the LOCA on the PCT which occurs some time later during the reflood stage. This relationship to stored energy is predicated on the assumption that energy removal during the blowdown will be nearly identical for each case when using EXEM in the hot channelmode.
As a result, residual clad temperature differences at the end of blowdown resulting from initial stored energy differences will also be reflected throughout the adiabatic heatup phase 4ntil core midplane quench occurs. *As quench time is detennined on the basis of core average channel midplane reflood, the hot channel PCT results at quench will linearly reflect initial stored heat conditions.
- ~tTs obtained for either fuel initialization codes are on the order of 300°F below the 10 CFR 50.46 limit over the initial 10 GWD/MTM burnup which is the burnup for which the licensee has received o~r approval.
4.2 Conclusion Identification of the Dresden 3 Design Basis LOCA on the basis of the generic BWR-3 break location and spectrum studies previously reviewed and apgroved is considered acceptable, particularly in view of the large margin (300 F) to the limit value. Analysis of the LOCA was perfonned with reviewed and accepted codes for the initial 10 GWD/MTM of operation. Acceptance of the studies perfonned using RODEX2 will.be contingent on future staff acceptance without modification of the ~ode for fuel burnup effects over the remainder of core life~. Changes required for acceptance of RODEX2 will a,lso require reevaluation of the.. PCTs shown on Figure 1, and potential respecification of the allowable MAPLHGR limits shown on this figure.
Based on the analysis perfonned with the approved version of EXEM over the initial 10 GWD/MTM,. the MAPLHGR results shown on Figure 1.1 of Reference 16
. can appropriately be used as the basis for establishing Technical Specification limits on linear power generation rates for Dresden 3 operation with the Cycle 8 reload core for the first 10 GWD/MTM of operation.
If RODEX2
, approval is not available prior to that time, additional GAPEX*analyses will be required to verify the acceptability of the proposed MAPLHGR 1 imi ts *
. Accordingly, we have modified the appropriate TS to require the licensee to submit additional analyses for NRC approval using an approved code if use.of RODEX2 code has not. been approved prior to exceeding 10 GWO/MTM.
5.0 PLANT TRANSIENTS ANALYSIS The plant transients analyzed for the Dresden 3 Cycle 8 reload were perfonned by the Exxon Nuclear Company (ENC) (Reference 18) to assess hot channel thermal margin to transition boiling and system pressure margin to safety limits for*
limiting thennal and pressurization transients. Evaluation of the thermal margin to transition boiling was assessed for three transients whfch had been
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. r found to provide bounding thermal transients in a generic BWR-3 study (Reference 39) in which eight categories of potential transients had been considered.
The bounding thermal transients identified in the generic study and analyzed for Dresden 3 were:
- l. Generator Load Rejection Without Bypass
- 2.
Increase in Feedwater Flow
- 3.
loss of Feedwater Heating Similarly, for the bounding overpressurization transient, the generic study of Reference 39 indicated that the containment isolation event involving rapid closure of all main steam line isolation valves assuming no direct reactor
, scram and no steam line relief valve operation would produce the maximum overpressurization transient for assessment of compliance to ASME Pressure Vessel Code requirements.
5.1 Thermal Margin Assessment The thermal *margin assessments for the three bounding transients analyzed for Dresden 3 were performed with approved (Reference 40) ENC plant system simulation models.
The COTRANSA (Reference 41) formulation of ENC's approved plant simulator was used to analyze the first two transients due to the potential for core void collapse and consequent axial power redistribution before scram during these transients. The loss*-~of feedwater heating was analyzed using the PTSBWR3 (Reference 42) formulation of the approved model as void collapse is minimal for this transient. The principal difference between the two simulator formulations used for these analyses concerns the use of one-dimensional neutron kinetics in COTRANSA to allow consideration of axially dist~ibuted neutronics, and the use of point kinetics in PTSBWR3 which allows only changes in power level with a fixed axial power distribution shape.
Results computed by COTRANSA for the generator.load rejection transient were
- obtained using nominal design data for all plant parameters.
To assess the.
. effect of parametric uncertainties on the principal mitigating action for the transient (reactor scram) results of the COTRANSA calculations, obtained with nominal input data, were used in ENC's statistical methodology (Reference 13) for computing uncertainties in the critical power ratio change (~CPR). The
~CPR uncertainty computed by the statistical method was based on the use of one standard deviation for void reactivity, scram worth, control rod insertion speed, and scram time delay *. The ~CPR obtained with a 95% certainty was found
,to be 0.25 for the generator load rejection transient. With a Minimum Critical Power Ratio * (MCPR) limit of 1.05 for Dresden-3 as determined by ENC's CPR Methodology (Reference 13) the operating MCPR limit for this transient.was 1.30.
However, based on uncertainties identified by the staff, the safety and operating MCPR limits were later modified as discussed
- in Section 2.2 of this SER to add additional conservatism.
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IV.
14 tiCPRs calculated by the 1 i censee for the feedwater fl ow increase and heater loss transients computed with conservative values for the parametric,
uncertainties demonstrated small er tiCPRs, namely, 0. 21 and 0. 16, so that the generator load rejectjon results provide the bounding thermal margin co'nditions
- to be used for establishing technical specification operating limits.
The thermal margin analyses for the other quite similar fuel types contained in Cycle 8 demonstrated essentially identical behavior to the ENC elements as shown on Table 2.1 of Reference 18 so that the operating MCPR limit bf 1.33 for ENC and GE 8x8 fuel designs and 1.34 for GE 8x8R fuel design are acceptable.
The licensee's proposed TS changes regarding OLMCPR have been modified accordingly.
5.2.Overpressurization Transient Calculation of the containment isolation event resulting in closure of all steam line isolation valves without direct scram and relief valve operation was computed with the COTRANSA version of ENC's Plant Transient Simulator.
This version was used to provide adequate consideration of the core void collapse produced by pressure wave propagation into the reactor vessel before reactor:scram is actuated. Conservative assumptions concerning the neutron flux trip level, control rod insertion time, control rod worth, void reactivity feedback, and end of core life conditions for fuel gap conductance were used to provide a conservative bounding system pressure transient.
- Results of the analyses demonstrated a peak pressure in the 1 ower plenum of 1361 psia or 29 psi below the ASME Pr~ssure Vessel Code Limit of 1390 psia
. for the Dresden 3 vessel (1.10 x Design 3=:1.10 x 1250 psig = 1375 psig).
The corresponding steam* dome pressure is 1324 psig. The TS safety limit is based on dome pressure.
We, therefore, conclude that the licensee's proposed RCS safety limit of 1345 psig is conservative with respect to the above
- evaluation and is therefore acceptable.
5.3 Conclusion 1.j We conclude that-the limiting thermal margin and overpressurization transients analyzed for the Dresden 3 plant have been adequately identified in the generic BWR 3 analyses for both areas of consideration. The analyses of the limiting transients have been performed with reviewed and accepted analytical models, and the assumptions made relative to design parameter values used in the analyses are appropriate for providing bounding transients *. Therefore, we conclude that the results reported in Reference 18 are an acceptable basis for establishing Technical Specifications* governing operating conditions for the Dresden 3 plant using the Cycle 8 reload core.
We have reviewed the licensee's proposed TS changes and with the incorporation of the modified OLMCPRs, we find the licensee's proposed TS-changes to-be acceptable.
SUMMARY
We have reviewed the licensee's plant-specific reports submitted_ in support of Cycle 8 operation for the Dresden Nuclear Power Station Unit 3 using reload fuel supplied by the Exxon Nuclear Company.
Based on our review,
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15 the licensee's analyses were found to be acceptable with the following modifications:
- 1.
The staff imposed OLMCPRs of 1.33 for the ENC and GE 8x8 fuel designs and an OLMCPR of 1.34 for the GE 8x8R fuel design shall be incorporated in the Technical Specifications;
- 2.
The licensee shall confinn the adequacy of the present calculations or
.submit.additional calculations of rod pressure and fuel centerline temperature with acceptable methods prior to reaching 10,000 MWD/MTU unless the use of RODEX2 code is approved without modifications; ~nd
- 3.
An increase in the 6CPR shall be calculated in accordance with the equation specified in Section 2.3 of this SER in the event the average measured *scram __ rod _insertion time~ exceed* 2.58 seconds.
The licensee has submitted proposed changes to the Dresden 3 Technical Specifications. These changes modify the MCPRs and incorporate the ENC curves for determining the OLMCPR for core flows less than rated.
Based on our review, we have detennined that with the. incorporation of the modifications stated above, the licensee's proposed TS changes are acceptable.
Although our review of ENC's analysis for reduced flow operation (Reference 19) is incomplete, we have. detennined that the existing Technical Specifications pertaining to reduced flow operation are still.applicable and are conservative.
Therefore, the existing TS governing reduced flow operation shall remain in
~ffect pending completion of ou~ ~ev1ew.
We have modified the proposed TS to incorporate the above three mod1f1cations.
Thus, we conclude that the licensee* s proposed* TS changes to support Cycle 8 operation are acceptable.
V.
ENVIRONMENTAL CONSIDERATIONS We have determined that the amendment does 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 detennination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 51.5(d)(4) that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of the amendment.
VI. CONCLUSION We have concluded, bas~d on the considerations discussed above, that:
(l) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Conmission's regulations and the issuance Of the amendment Will not be inimical.to the COITD110n defense-and security Of to the health and safety of the public.
Date:. April 29, 1982
. ~
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- - -~TABLE*.1:-:6 CP!rs**For -Anticipated Plant Transients * *"*-* ----*.
- ENC Transient
~ CPRs 6E 8 x 8
-GE 8 x BR.
~~~--...~~--...~--...__,___,_~~--...~__,_~~__,___,_~~~~~~~~--
c:..- ~?ad Rejectio~ Without
_ - ** bypass.
._ Increase in Feedwater *
- Flow Loss of Feedwater
- Heating Contro 1 Rod w; thdrawa 1 Misloaded Bundle
- o.25Cll
- _o.2a<~>
0.21 0.16 0.15 0.16 o.2sf1 >
0.28(2) 0.21 0.16 0.14 0.16 o.2sCl >
-~ a,2at2l o~zi--
- 0.16 0.14 0.16 (l) Detennined using response surface methodology described in XN-NF~Sl-22.
tzl staff ca1~u1ated. values. acco\\mt1ns for eode,m~&!r~ain1t~s and used to calculate OLMCPR *.
'** *. * "i
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. bcPR. SENSITIVITY* RESUt TS FOR DRESDEN UNIT 3.(GENERATOR t.OAD REJECTION)
_._Limit. for Dlj~de!J. ~ ".. - _Bas_e Case(2)
-..... _.. Ope.r.at~Qnl ACPR D.cS (0.248) 4 PREDICTOR VARIABL£S.
- Void *Reactivity Scram React 1v1ty
- Scram* Delay Scracn t'nsertion Rate 1 *
. : 0.22 (0~219). ~
HOM°INAL + :i a 1.1 x Calculated
.9 x Calculated Z98 11sec *.
100 cm/sec 0.175
. ft~lNAt.
- Calculated Calcu1ctted 238 msec 141 QDfscc
~..:*-.
DETERHlHiSTit VARIABLES Steam Line Dynamics
. COHSERVAT1-V£ ftOHINAL 80 mscc bias on sc:ram delay
- Ho bias
- 1 Core ~vg._Gap Conductance 893 BlU/lbm-ft-Of.
1165 *aTU/lbm-ft-OF.
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~------------~-
- 1 so msee *
. o.2s n-1 _
Test Results 250 msec (1) Value derived with predictor var1ables at va~ues such that the ACPR calc;ulated h equal to _or greater _than 951 of possible outcomes.
(2)
- Value with predicto~ variables at nominal Yalues (3) Note (2) applies.. ncn1na1 deten1inistic variables used*
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. ENC Propos*ed
. c 1 '8...
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- ~N/A(Z) 1.'!30 1.30 ;.
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<21*there are no *7 x*7 bundles in the cycle 8 core *
~.
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1.34:..
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. 6.0
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- PCP using GAPEX lU a
ASSEt-!Bl~ A~ERAGE BURNUP, GWD/MTM
\\.
Fi(!Ul'B 1 DRESDEN Unit 3,;., MAPLHGR and PCT vs. Assembly Ave.rage Bumup
' ' ' '8
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References
- 1.
XN-NF-79-59(P), October 1979 Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies
- 2.
XN-NF-79-7l(P), Revision 2, November 1981 Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors
- 3.
XN-NF:..8Q-19(P), Volume 1 (Supplements 1 and 2), May 1980 Exxon Nuclear Methodology for Boiling Water Reactors Neutronics Methods for Design and Analysis
- 4.
XN-NF-80-19(P), Volume 2, Revision 1, June 1981 Exxon Nuclear Methodology for Boiling Water Reactors EXEM:
ECCS 'Evaluation Model, Surranary Description
- 5.
. XN-NF-80-l 9(P), Volume 2A, Revision 1, June 1981 Exxon Nuclear Methodology for Boiling Water Reactors RELAX:
A RELAP4 Based Computer Code for Calculating Slowdown Phenomena
- 6.
XN-NF-80~19(P), Volume 28, Revision 1, June 1981 Exxon Nuclear Methodology for Boiling Water Reactors FLEX:
A Computer Code for Jet Pump BWR Refill and Reflood Analysis
- 7.
XN-NF-80-19(P), Volume 2C, June 1981 Exxon Nuclear Methodology for Boiling Water Reactors
- Verification and Qualification of EXEM 8;
XN-NF-80-19(P), Volume 3, Revision 1, April 1981 Exxon Nuclear Methodology for Boiling Water Reactors
- THERMEX:
Thenna l Limits Methodo 1 ogy, Summary Descri pt i Of'!
- 9.
XN-NF-81-22(P), September 1981 Generic Statistical Uncertainty Analysis Methodology
- 10.
XN-NF-81-58(P), August 1981 RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model
- 11.
XN-NF-81-7l(P), October 1981 Generic Jet-Pump BWR3 LOCA Analysis Using the ENC EXEM Evaluation Model 12~
XN-NF-512(P), Revision 1, March 1981 The XN-3 Critical Power Correlation *
- 13.
XN-NF:..524(P), November 1979 Exxon Nuclear Critical Power Methodology for Boiling Water Reactors
- 14.
XN-CC-33(A), Revision 1, November 1975 HUXY:
A Generalized.Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option
- 15.
XN~NF-81-21 (P),
11Mechanical Design for Exxon Nuclear Jet Pump BWR Fuel.
Assemblies, 11 November 1981
. *... :-:*-.-.----~-.. ~----~ :.... -*.
.. '* (;.-
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- -*~-.. ~
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- 1
- r
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I
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- 16.
XN-NF-81-75, "Dresden Unit 3 LOCA Analysis using the ENC EXEM Evaluation Model--MAPLHGR Results," October 1981
- 17.
- 18.
- 19.
- 20.
- 21.
- 22.
XN-NF-81-76, "Dresden Unit 3 Cycle 8 Reload Analysis," Exxon Nuclear Company, December 1981 XN-NF-81-78, "Dresden 3 Cycle 8 Plant Transient Analysis Report,"
December 1981 XN-NF-81-84, "Dresden Unit 3 Analyses for Reduced Flow Operation,"
November 1981 NEDE-240ll(P}-A, "General Electric Boiling Water Reactor Generic Reload Fuel Application," May 1977 NEDE-24284(P}, "Assessment of Fuel-Rod Bow in General Electric Boiling Water Reactors," August 1980 Paul S. Check (NRC) to T. Novak (NRC) and R. L. Tedesco (NRC), "Safety Evaluation for Qualification of the One-Dimensional Core Transient Model for *Boiling Water Reactors," NED0-24154 and NEDE-24154P, Volumes I, II, and I I I.
- 23.
Themis P. Speis (NRC) to Thomas M. Novak (NRC}, "Dresden 3 Cycle 8 Reload LOCA and Accidents Transients Reviewed," March 26, 1982
- 24.
Thomas J. Rausch (CECo) to Harold R. Denton (NRC), "Dresden *station Unit 3 -
Proposed Amendrilent to Appendix A Technical Specifications to Support Operatio_n with Fuel Applied by Exxon Nuclear Company, NRC Docket No.
50-249," January 11, 1982
- 25.
Thomas J. Rausch (CECo) to Harold R. Denton (NRC), "Dresden Station Unit 3 Response to Request for Additional Infonnation Concerning the
- 26.
- 27.
28 *
- 29.
- 30.
Cycle 8 Reload with Exxon Fuel NRC, NRC Docket No. 50-249," March 22, 19.82 L. S. Rubenstein (NRC) to Robert L.*Tedesco (NRC), "Review of Exxon Nuclear
- Company's Nuclear Methodology for Boiling Water Reactors - Neutronics Methods for Design and Analysis {Report Number XN-NF-8019(P) Volume 1, Supplement 1 and 2) (TACS 42277)," March 18, 1982 XN-NF-81-21 (P), *Revision 1, "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," January 1982 XN-73,..25, "GAPEXX: A Computer Program for Predict.fog Pellet-to-Cladding Heat Transfer Coefficients," August 13, 1973 XN-74-52, "Oyster Creek 1975 Reload for Cycle 5 Licensing Data Submittal,"
Revision 3, January 22, 1975 T. J. Rausch (CECo) to H. R. Denton, ECCS Analysis Results Presented in
. Exxon Report XN-NF-8l-75(P) dated February 23, 1982
- '11*
f,:1_* _________ :_:~.:__; __ __.___----___ ::..::::_____
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- _ __.._~'...... -* -- ~ --*--.. -'-*--~----------*- _____ :....*-***-***::-*-* ---*--- ---~
1. j
~
- i l l l
. '1.*.
- 31.
R *. P. Denise {NRC) Memorandum for R. J. Mattson, "Summary of Minutes of Meeting on Cladding Rupture Temperature, Cladding Strain, and Assembly Flow Blockage," November 20, 1979
- 32.
NUREG-0630, D. A. Powers and R. 0. Meyer, 11 Cladding Swelling and Rupture Models for LOCA Analysis," April 1980
- 33.
D. G. Eisenhut {NRC) letter to All Operating Light Water Reactors, November 9, 1979
- 34.
H. R. Denton (NRC) memorandum for Commissioners, "Potential Deficiencies in ECCS Evaluation Models," November 26, 1979
.35.
XN-NF-81-75{P}, "Dresden Unit 3 LOCA Analysis Using the ENC EXEM Evaluation Model, MAPLHGR Results," November 1981
- 36.
Exxon Nuclear Company, "Generic Jet-Pump BWR3 LOCJ\\ Analysis Using the ENC EXEM Evaluation Model,
11 XN-NF-81-7l{P}, Supplement l, November 1981
- 37.
NRC Letter: James R. Miller, Div. of Licerising, NRC, to Mr. G. F. Owsley~
Manager, Reload Fuel Licensing, Exxon Nuclear Company, "Acceptance for Referencing of Topical Report XN-NF-80-19(P} Volumes 2, 2A, 2B and 2C
- 38.
- 39.
- 40.
- 41.
- 42.
and Topical Report XN-NF-8l-7l(P},
11 January 27, 1982 Exxon Nuclear Company, "Fuel Rod Thennal-Mechanical Response Evaluation
- Model, 11 XN-NF-81-58(P), August 1981 Exxon Nuclear Company's Plant Transient Code for Evaluation of Abnormal Transients for Jet Pump Boiling Water Reactors, Supplement 2, Generic Application of PTS-JP BWR and COTRANSA to JP BWR Plants, XN-NF~79.:.71(P},
Supplement 2, April 1981 Safety Evaluation Report on. Exxon Nuclear Company's Jet Pump-BWR-Plant Transient Simulator, NRC, March 1981 Exxon Nuclear Company's Plant Transient Simulator Code for the Evaluation of Abnormal Transients for Jet-Pump Bai.ling Water Reactors, XN-NF-79-71(Pr, Supplement 1, October 1980 Exxon Nuclear Plant Transient Model for Jet Pump Boiling Water.Reactors, XN-XF-79~7l{P}, Revision 1, May 1980
- 43.
T. Rausch (CECo} to H. Denton (NRC}, letter transmitting additional ENC Data on Hydraulic Compatibility, March 22, 1~82
- 44.
G. Owsley- (ENC) to c. Berlinqer (NRC), letter transmittinq additional
. an~lyses on MCPR operating limits, April 17, 1982
- 45.
Safety Evaluation Report on GE ODYN Code Methodology, NRC, October 22, 1980
- 46.
- G. Owsley (EN Cl 1 etter to C. Berlfnger (NRC}, 1 etter on Effect of NUREG.. Q6.3Q on Dresden Unit 3 ECCS Analysis, April 14, 1982..
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