ML20087D840
| ML20087D840 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 03/09/1984 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20087D794 | List: |
| References | |
| NUDOCS 8403140188 | |
| Download: ML20087D840 (12) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION WASHINGTON D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION '
SUPPORTING Af'ENN:ENT N0. 74 TO FACILITY OPEPATING LICENSE NO. DPR-25 COI:MONWEALTH EDISON COMPANY DRESDEN NUCLEAR POWER STATION, UN!T'NO.3 DOCKET N0.
50-249
1.0 INTRODUCTION
By letters dated July 18,1983 (Ref.1) and August 25, 1983 as supplemented by letters dated November 3, 10, and 30, 1983, two letters dated December 13, 1983, and a letter dated December 16, 1983 Commonwealth Edison Company (CECO)
(the licensee) proposed to amend Appendix A of Facility Operating License No. DPR-25.
The requested amendment furnished information tc support authorization for Dresden 3 to install, in place of eight standard control blades, eight lead control blades designed and built by ASEA-Atom and to support Cycle 9 operation of Dresden 3 with reload fuel supplied by and the associated analyses performed by Exxon Nuclear Company.
The A5EA-Atom blades are being tested as part of a demonstration program sponsored by EPRI aimed at qualifying a new blade' design which would provide a greater exposure lifetine than the current design.
In support of their proposal, the licensee has submitted a Technical Report TR-BR 82-98, Revision 1 (Ref. 2) for review.
The Dresden 3 Cycle 9 (D3C9) reload will consist of 408 fuel bundles fabricated by Exxon Nuclear Company (ENC).
These 8x8 bundles are comprised of 63 active fuel rods and one inert water rod. During Cycle 9 operation the ENC fuel will reside with the 316 General Electric (GE) fuel assemblies presently in the core.
In support of the D3C9 reload Commonwealth Edison Company (CECO) submitted topical reports which described the steady-state reload analysis, XN-NF-83-47, the plant transient analysis, XN-NF-83-58, and-theloss-of-coolantaccident(LOCA) analysis,XN-NF-81-75, Supplement 1.
Notices of Consideration of Issuance of Amendment to License and Proposed No Significant Hazards Consideration Determination and Opportunity for Hearing related to the requested actions in the July 18 and August 25, 1983 letters were published in the Federal Register on November 22, 1983 (48 FR 52807 and 52808).
No request for hearing was received.
A verbal comment from Mr. R. Minue of the Illinois Department of Nuclear Safety was received on December 5, 1983.
His concern was related to the indications of cracking in ASEA-Atom control blades at high burnup as discussed in the licensee's November 10, 1983 supplemental letter.
The blade cracking issue is addressed in Section 2.5 of this safety evaluation.
The supplementary letters furnished clarifying information needed by the staff but made no changes in the content of the amendments and were, therefore, encompassed within the prenotices published November 22, 1983.
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2.0 EVALUATION OF THE USE OF THE ASEA-AT0li CONTROL BLADES i
j 2.1 GENERAL DESCRIPTION OF BLADE DESIGN AND PRESENT OPERATING SE0VENCE The ASEA-AT0ft (A-A) blades to be installed in Dresden Unit 3-have been designed to be mechanically conpatible with existing blades.
The blade 7
profile is quite close to the standard blade end the velocity limiter and drive coupling portions are identical. The blades may be manipu-lated with the same handling tools as used on the standard blade.
The blade weight is slightly less than the standard blade.
The absorber naterial is vibratory-compacted B C but the blade design permits sig-g nificantly more boron to be placed in the blade.
Dresden Unit 3 is currently operating with the Exxon single sequence control strategy.
This means that the same control rods remain in the core throughout the Cycle (as opposed to periodic sequence changes in previous cycles).
The A-A rods will be among those remaining in the core in order to raximize their exposure.
P.2 11ATERIALS C0ftPATIBILITY There will be two types of A-A blades used in D3C9.
Four of the eight blades will have only 8 C as an absorber material, and four will have 4
both B c and hafnium retal as absorber materials.
The hafnium will t
comprise only the top six inches of the absorber section of these four blades. This design provision has been nade to allow additional blade lifetime and reduce internal pressure in the blades. The use of hafnium in control blades has previously been approved for GE test blades in Peach Bottom, and is an alternative for the silver-indium-cadmium (Ag-In-Cd) used in Westinghouse reactors.
The staff is unaware of any na-terials problems associated with the use of hafnium, and finds this aspect of the design acceptable.
The absorber in the A-A blade design is contained in horizontally drilled absorber holes in low-carbon stainless steel sheets. The staff's review of the mechanical design of the blades included a request for additional information (Ref. 3) related to the potential for blocking of the individual slits which interconnect these holes to equalize internal gas pressure in each blade wing. The applicant's response (Ref 4) provides adequate assurance that there is no potential nechanism for blocking gas communi-cation between the 8 C holes.
4 In addition, the staff evaluated additional information furnished by Conmon-wealth Edison (Ref. 4) on the conservatism of a 10 percent helium release rate (fron B C) on blade temperature calculations, maxinum internal gas pressure, 4
nechanical strength and strain design reouirements, use. of gridrads, and the-seisnic design. Conmonwealth Edison (Ref. 4) provided justification that each of these concerns has been addressed satisfactorily in the design of the control blades.
1.
2.3 NUCLEAR DESIGN CHARACTERISTICS The nuclear design characteristics of the improved A-A control blades has been performed by A-A with the PHOENIX lattice and depletion code. While this code has not been reviewed by the st9ff, a-sufficient description of it has been included (Ref 2) to penait.the conclusion that it is acceptable for use in performing the comparsions between the neutronic characteristics of the standard and A-A blades that are presented.
The code has been used to compare reactivity worths at cold xenon-free conditions and hot voided and unvoided conditions as a function of fuel burnup.
In addition power distribution effects and absorber depletion effects have been studied.
The conclusions of the analyses are discussed below.
The presence of a larger baron inventory in the rods inplies a greater reactivity worth.
The calculations by A-A have shown that the worth of-the all B C rods is 6 to 9 percent greater than that of the standard 4
rods. A control blade containing all hefnium would have about the same worth as the standard blade.
An important effect of the increased rod worth is to increase the shutdown maroin.
However, the increase in shutdown nargin will be small for Dresden 3 since there are only eight of the stronger control rods and they will be placed in low worth regions of the core. Another effect of the increased boron content is a steeper flux gradient in assemblies surrounding an inserted control blade.
The maximum difference is in the wide-wide corner and is about 5 percent. The difference at the LPRM location is only about 0.5 percent.
These differences are accounted for in the reload analyses.
The increased blade worth may cause the consequences of a rod withdrawal or rod drop event to be more severe.
The effect of the presence of the A-A rods in Dresden 3 will be addressed for each reload containing them.-
1 The increased boron loading of the blades also provides a longer exposure lifetine. A-A calculations show that the improved blade will have a 60 percent greater life if end-of-life is defined as.a 10 percent reduction in blade worth.
If the lifetire is determined on the basis of equal end-of-life worths, the improved rod would have more than twice the lifetine of the standard rod.
2.4 CONTROL R0D MANEUVERING The A-A control rods are essentially identical in exterior envelope to the standard rods. The all B C rods are about 12 pounds lighter than g
current rods and the rods with hafnium tips are about 7 pounds lighter.
Thus, the insertion speed should be greater for these rods.
- However, the presence of fr.iction pads rather than rollers and an.open central structure (increasing flow resistence) tends to offset the smaller weight.
It is concluded that the insertion (scram) speed will not be significantly affected by the improved rods.
The scran speed will be neasured as part of the startup testing program.
2.5 BLADE SURVEILLANCE PROGRAM By letter dated November 10, 1983 (Ref. 5) the licensee informed the staff that evidence of cracking with sone loss of B C had occurred in 4
similar rods being used in a Swedish reactor.
Based on the proposed positioning of the eight lead A-A rods in the Dresden 3 core, the burn-up of the rods in the Swedish reactor, at the time the cracking was discovered, was greater than that which will occur during two 18 month cycles in Dresden 3.
However, the lead rod burnup will be greater after three 13 month cycles than the burnup of the rods in the Swedish reactor.
Despite this, the staff has concluded that, because there are differences between the two sets of rods, concerns relating to their use are alleviated.
In addition, the licensee has proposed an extensive monitoring program while they are being used at Dresden 3 so that indications of inferior performance will be detected promptly. These factors are significant i) enough for the staff to conclude that the Swedish problems would not be expected at Dresden 3.
First, the stainless steel in the rods to be used in Dresden 3 has been fabricated with tighter chenistry control than that used in the blades used in the Swedish reactor.
Second, nondestructive examination of the Dresden 3 A-A rods will be conducted following each usage cycle. Tests will be perforned to check dinensional stability, corrosion effects and the integrity of the B C containment. A high re-g solution TV canera will be used for visual inspection, a guaging fixture for dimensional stability and a neutron transmission measurement.for demonstrating B C presence. After the third 18 months cycle, an extensive 4
examination of one or more rods will be made after their renoval from the Core.
Based on the above and upon the fact that four of the rods use hafnium instead of B C in the top six inches naking them less susceptible to IGSCC from B C swelling, the staff has concluded that there is nct a cracking-rel ted safety concern from use of A-A rods in Dresden 3.
2.6
SUMMARY
On the basis of its review the staff has concluded that the use of the A-A inproved control blades in Dresden 3 is acceptable. This conclusion is based on the following considerations:
1.
The inproved blades are nechanically and hydraulically compatible with the present control blades.
2.
Only eight of the rods will be installed in the reactor.
3.
The nuclear characteristics of the blades have been determined by acceptable methods.
4.
The presence of the blades will be,taken into account in the design end analysis of core reloads.
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5.
Sufficient experience has been had with the rod design in other (Swedish and Tinnish) R;!Rs to pernit the conclusion that they will operate without significant deterioration.
6.
A satisfactory surveillance program has been established-to monitor
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the blade performance.
3.0 EVALUATION OF THE DRESDEN 3 CYCLE 9 RELOAD SUBMITTALS
3.1 BACKGROUND
The 03C9 core will consist of 184 fresh ENC XN-2 8x8 fuel assemblies, 224
'nce-irradiated ENC XN-1 8x8 fuel and 316 GE 8x8 fuel assemblies. The o
ENC XN-2 8x8 fuel design is described in the approved generic report on the jet-pump (JP) BilR fuel design (XN-NF-81-21). This design is acceptable for use in the D3C9 reload with the exception of a previously applied burnup restriction on MAPLHGR limit (see section 3.2 of this report) and several conditions of approval on the generic fuel design report.
These conditions are:
(1) The licensee must confirm that the design power profile shown in Fig. 5.10 of Ah-NF-81-21 bounds the power limits for the application in question.
(2) Unless RODEX2 (XN-NF-81-58) is approved without modification, the licensee must confirm or redo the following analyses, which were reviewed on the basis of R0DEX2 results:
design strain, external corrosion, rod pressure, overheating of fuel pellets, and pellet cladding interaction.
(3) Until such time that the Exxon revised cladding swelling and rupture models (XN-NF-82-07) are approved and incorporated in the ENC ECCS evaluetion model, a supplemental calculation using the NUREG-0630 cladding models must be provided on a plant-specific basis each tine a new ECCS analysis is performed.
(4) The licensee must make sure that the fuel perfornance code that is used to initialize Chapter 15 accident analyses has current NRC approval.
The staff has evaluated these four conditions during the course of our review, and its conclusions are described in _the following paragraphs.
3.1.1 Power Historv The licensee stated in the D3C9 reload submittal (XN-NF-83-47)
(Ref. 6) that the 03C9 expected power history is bounded by the design profile in. Fig. 5.10 of XN-NF-81-21 (Ref. 7). The staff has reviewed the references relating to the power history and concludes that the Cycle 9 power history is within the design limit and condition 1 is satisfied.
. s 3.1.2 R0DEX2 -- Strain, Corrosion, Rod Pressure, Overheating of Fuei Pellets, and Pellet - Clad Interaction (PCI) Analyses The analyses of. strain, corrosion, rod pressure, overheating of fuel pellets, and PCI were described in the approved JP-BUR fuel design.
The staff has completed the review of the RODEX2 code'used in this analysis and approved it with some modifications for licensing applications.
Using the approved version limits on these physical parameters would not be exceeded throughout the entire lifetine. Since these analyses bound the Cycle 9 applications, the staff concludes. that these analyses are acceptable for Cycle 9.
3.1.3 Cladding Swelling and Ruoture The cladding swelling and rupture models in XN-NF-82-07 (Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model) have been approved for use in the ENC ECCS evaluation model and have been incorporated in the approved ENC EXEM/ BUR ECCS model. Since ENC used that approved swelling and rupture model for cladding in ECCS analysis, Condition 3 has been satisfied.
3.1.4 LOCA Initial Conditions ENC used the recently approved steady-state code, RODEX2 (XN-NF-81-58), to calculate Cycle 9 LOCA initial conditions including stored energy and rod pressure for the. ENC EXEft/BWR evaluation model. Thus Condition 4 is satisfied by the use of the approved code RODEX2.
3.2 MAPLHGR LIMIT The MAPLHGR limit for XN-1 fuel during Cycle 8 operation was approved for burnups only up to 10,000 mwd /MTU due to the use of then unapproved code RODEX2.
Subsequently the licensee requested that the 10,000 mwd /tiTU limit be extended to 15,000 mud /MTU, which was also approved (Ref. 8). Since the staff has approved the RODEX2 code, the lic.ensee has confirmed that the MAPLHGR limit remains the same with the use of the approved RODEX2 code (Ref. 9). The staff finds this acceptable.
The MAPLHGR limit for XN-2 fuel in Cycle 9 is the same as the one for XN-1 fuel because of identical fuel design.
The staff concludes that the tiAPLHGR limit is acceptable for XN-2 fuel assenblies in Cycle 9.
3.3
SUMMARY
The NRC staff has reviewed the Dresden 3 fuel design and analyses for the Cycle 9 reload, and concludes that they are acceptable for Cycle 9 operation.
j 4.0 NUCLEAR DESIGN The nuclear design of the Cycle 9 reload has 'been performed in accordance with the procedures described in XN-NF-80-19. The procedures have been previcusly used and approved for this purpose (see, for example, Dresden 3 Cycle 8 reload) and their use for Cycle 9 is acceptable.
The results of the design analyses I
t.
are given in Section 4.0 of Xf!-f F-83-47 (Ref. 6) including Table 4.1 and in Table 3.2 of XN-NF-83-58 (Ref. 6). These results are within the range nornally expected for BUR reloads and are acceptable.
i The use of eight A-A control blades for Cycle 9 has been approved as discussed.
in Section 2 of this Safety Evaluation.
4.1 TRA!!SIENT AND ACCIDENT ANALYSIS The control rod withdrawal error, the fuel loading error and the rod drop accident were evaluated for Cycle 9.
The use of the Single Sequence Control stretegy (in which rods inserted during power operation have low worth) assures that the control rod withdrawal error will not be limiting. Using a Rod Block Monitor setting of 110 percent of full power results in a ACPR of only 0.16.
The maxinum change in CPR due to a fuel loading error is 0.19 and this event is not liniting either.
The control rod drop accident evaluation yields a value of 85 calories per gren for the maxinum deposited fuel enthalpy. This is well below the staff's acceptance criterion of 280 calories per gram.
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The effect of the presence of the eight A-A control blades on rod withdrawal
- and rod drop events has been considered by the licensee.
The A-A blades will be located in low reactivity positions within the core and thus will not be the limiting rods for the rod withdrawal event.
The startup withdrawal sequences were examined and the maximun potential ejected red worth for the A-A blades was likewise found to be below that for the standard blades.
THe resultant peak enthalpy was also lower for these blades.
The staff concludes that the presence of the A-A blades has been adequately evaluated.
5.0 THERl!AL-HYDRAULIC DESIGN
5.1 BACKGROUND
The review of the thermal hydraulic aspect of D3C9 consisted of the following:
(a) the operating safety limit minimun critical power ratio (0Lf1CPR),
(b). thermal-hydraulic stability, l
(c) the Technical Specification changes.
The objective of the review was to confirm that the thermal-hydraulic design of the reload core was accomplished using acceptable analytical methods, to confirm that an acceptable margin of safety from conditions which would lead to fuel damage during normal operation and anticipated operational. occurrences (A00s) is provided, and to confirm that the Cycle 9 core is not susceptible to thernal-hydraulic instability.
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5.2 MINIt'Uti AND OPERATING LIMIT CPR The methodology for deternining uncertainties and their application in deternining the MCPR linit is contained in XN-NF-80-19 Volune 1 (Ref.10) and XN-NF-512 (Ref.11) and XN-NF-524 (Ref.12).
XN-NF-524 Volume 1 has been reviewed and approved by the staff-(Ref. 13)
The staff has completed the generic review of XN-NF-512 (Ref.14) and XN-NF-524 (Ref.15) and has concluded that the nethodology for applying the XN-3 nean and standard deviation to arrive at a 1.05 for ENC fuel and a 1.06 for GE 8x8R fuel is acceptable.
Various operational ~ transients could reduce the MCPR below the intended safety limit. The most limiting transients have been analyzed by the licensee to determine which event could potentially induce the laroest reduction in the initial critical power ratio ( ACPR).
Table 2.1 of XN-NF-83-58 contains the results of these analyses.
The transient which resulted in the largest ACPR was the load rejection without bypass.
The ACPR for the load rejection without bypass was calculated using the statistical methodology described in XN-NF-81-22 (Ref.16), which has been reviewed and approved by the staff (Ref. 17).
Based on this analysis the applicant has proposed a ACPR of 0.25 at a 95% probability level.
The addition of this ACPR to the safety limit MCPR results in an operating limit MCPR (0LMCPR) of 1.30 for the ENC and GE 8x8 fuel designs, and 1.31 for the GE 8x8R design (Ref. 6).
Until the staff completes its generic review of XN-NF-79-71 (Ref.18) the staff will require that code uncertainties be accounted for using the methods discussed in the safety evaluation report on the GE ODYN code (Ref. 19) as described for implementation in the staff safety evaluation for Dresden 3 Cycle 8 (Ref. 20).
Such a procedure for Dresden 3 requires tha't an ENC code uncertainty value of 0.022 aCPR/ICPR be applied deterministically to aCPR calculations. When this a CPR is added to the MCPRs the resultant OLMPCRs are 1.33 for ENC and GE 8x8 fuel designs and 1.34 for GE 8x8R fuel (Ref. 6).
The staff concludes that such an increase in ACPR acceptably bounds the code uncertainties and that the limits so derived will assure that the safety limit MCPR is not violated in the event of any anticipated transients.
5.3 THERMAL-HYDRAULIC STABILITY The thernal-hydraulic stability of the Cycle 9 core was analyzed using the methods described in XH-NF-80-19, Volune 1, Supplement 2.
The calculated.
decay ratio at the natural circulation - 100% rod line intersection (which is the least stable physically attainable point of operation) is 0.33.
The calculated decay ratio for Cycle 8 was 0.45..The smaller decay ratio reported for Cycle 9 operation is attributed to the use of higher inlet orifice loss cnefficients (which are more representative of the hydraulic characteristics of the ENC fuel assen51y) in the Cycle 9 core stability analysis.
Based on the
fact that,iet pump BWRs are not pernitted to operate in the natural circulation node and the fact that the decay ratio shows a large margin of stability, the staff concludes that the stability analysis of the Cycle 9 core is acceptable.
5.4' TECHNICAL SPECIFICATIONS The licensee has submitted proposed Technical Specifications for D3C9 operation (Ref. 6).
Section 3.5.K specifies the operating limit MCPRs, which are 1.33 for the ENC and GE 8x8 fuel designs and 1.34 for GE 8x8P, fuel when scram times are less than or equal to 2.58 seconds.
When the neasured scram time becomes greater than 2.58 seconds the OLMCPR must be increased using the equation specified in Appendix A to XN-NF-83-47.
Both the OLMCPR limit and the equation for adjusting the OLMCPR are currently in the Dresden 3 Technical Specifications.
The only change to Technical Specification 3.4.K is to revise Figure 3.5-2 to incorporate the ENC curves for determining the OLMCPR for core flows less than rated flow. The revised Figure 3.5-2 is determined using the ENC methods documented in XN-NF-81-84, which is still under review by the staff. However, the review has progressed to the point where the staff concluded that the ENC methodology and the cal-culated results are acceptable for the D3C9 reload.
The staff,therefore, has concluded that the revised Figure 3.5-2 is acceptable.
5.5 FINDINGS The staff has reviewed the thernal-hydraulic design for the D3C9 reload core and has found that the results of analyses (XN-NF-83-47) support the proposed operating. limit MCPRs, which avoid violation of the safety limit MCPR for design transients.
The staff, therefore, concludes that this core reload will not adversely affect the capability to operate Dresden 3 safely during Cycle 9 operation and the proposed Technical Specification 3.5.K and the revised Figure 3.5-2 of the Technical Specifications discussed above are acceptable.
6.0 CONCLUDING
SUMMARY
The staff has completed its review of D3C9 submittals including XN-NF-81-75 4
Supplement I, XN-NF-83-47, XN-NF-83-58 and information relating to the use of eight ASEA-Aton control blades and found that they arh acceptable.
The staff thus. concludes that Cycle 9 operation for Dresden 3 with the eight ASEA-Aton control blades and with 184 fresh ~ ENC XN-2 fuel assemblies is acceptable.
7.0 ENVIRONMENTAL OUALIFICATION The staff has determined that the amendnent does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant envircnnental impact.
Having made this determination, the staff further concludes that the amendrent 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 cnd environmental impact appraisal, need not be.
prepared in connection with the issuance of this amendment.
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8.0 CONCLUSION
i The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable-assurance that the health and safety of the will not be endangered by operation in the proposed nanner, and (2) public
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such 1
activities will be conducted in compliance with the Commission's regulations and the issuance of this anendment will not be inimical to the common defense and security or to the health and safety of the public.
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9.0 ACKNOWLEDGEftENT 1
The following staff members have contributed to this evaluation:
S. Wu j.
S. Sun j
II. Brooks C. Graves j -'
fl. Dunenfeld R. Gilbert Date: March 9, 1984 1
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REFERENCES i
1.
Letter to H. R. Denton, NRC, fron B. Rybak, Cecmonwealth Edison, dated July 18, 1933.
2.
Technical Report TR-BR 82-98, Revision 1, " Performance Verification
~,
of an Inproved BUR Control Blade Design" dated May 5, 1983.
3.
Letter to D. L. Farrar, Commonwealth Edison, from D. Crutchfield dated Novenber 1, 1983 4
Letter to H. R. Denton, NRC, from B. Rybak, Conmonwealth Edison, dated November 30, 1983.
Letter to H. R. Denton, NRC, from B. Rybak, Commonwealth Edison, dated 5
November ID, 1983, with attachnent.
6.
Letter to H. R. Denton, NRC, with attachnents, including XN-NF-83-47, XN-NF-83-58 and XN-NF-81-75, Supplement 1, from B. Rybak, Commonwealth Edison, dated April 25, 1983.
7.
XN-NF-81-27 " Generic Design Report - Mechanical Design for Exxon Nuclear Jet Pump Bl!R Fuel Assemblies," October,1981.
8.
Letter to D. Farrar, Commonwealth Edison from D. Crutchfield dated July 7, 1983.
9.
Letter to H. R. Denton from B. Rybak, Commonwealth Edison, dated l
December 13, 1983.
- 10. XN-NF-80-19 (p), Volume 1 and Supplements 1 and 2, May 1980 Exxon Nuclear Methodology for Boiling Water Reactor Neutronics tiethods for Design and Analysis.
- 11. XN-NF-512(P), Revision 1, The XN-3 Critical Power Correlation, March 1982.
- 12. XN-NF-524(?), Exxon Nuclear Critical Power Methodology for Boiling Water Reactors, November 1979.
- 13. Letter. to G. Owsley, Exxon Nuclear Company from J. Miller dated April 7,1982.
- 14. Letter to G. Owsley, Exxon Nuclear Company from H. Bernard dated July 22, 1982.
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- 15. Letter to J. C. Chandler, Exxon Nuclear Company from C. Thomas dated f
October 31,'1983.
- 16. XN-NF-81-22(P) Generic Statistical Uncertainty Analysis Methodology, l
September 1981.
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- 17. Letter to J. C. Chandler, Exxon Nuclear Company fron C. Thomas dated October 28, 1983.
-18. XN-NF-79-71(P), Revision 2, Exxon i;uclear Plant Transient I:ethodology l
l-for Boiling Uater Reactors, November 1981.
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- 19. Letter to Dr. G. G. Sherwood, General Electric Corpany from R. L.'Tedsco dated February 4, 1981.
- 20. Letter to L. Del George, Connonwealth Edison Company, from J. Hegner
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dated April 29, 1982.
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