ML20236V272
| ML20236V272 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 11/25/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20236V264 | List: |
| References | |
| NUDOCS 8712040200 | |
| Download: ML20236V272 (8) | |
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'o UNITED STATES l
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' ~g NUCLEAR REGULATORY COMMISSION e,
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.E WASHINGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. b4 TO FACILITY OPERATING LICENSE NO. DPR-22 NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-?63
1.0 INTRODUCTION
By letter dated July 27, 1987 (Ref. 1) Northern States Power Company (NSP),
requested changes to the Technical Specifications to allow operation of the Monticello Nuclear Generating Plant (Monticello) using General Electric-(GE) manufactured fuel assemblies and GE analyses and methodologies.
Enclosed were the requested Technical Specification changes and report (Ref. 2) discussing
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the reload and analyses done to support and justify Cycle 13 operation. By letters dated August 28 (Ref. 6) and September 3 and 16,1987, NSP submitted additional information on the proposed Technical Specifications in response to the staff's request for additional information. The August 28 submittal transmitted proprietary infomation on GE8x8EB fuel designs.
In the case of the September 16 submittal, the licensee provided a revised description and safety evaluation supporting changes that were submitted on July 27 and l
September 3, 1987. This supperting infomation does not substantially change the action notice or affect the proposed determination of no significant hazards consideration published in the _ Federal Register on September 23, 1987.
The reload for Cycle 13 is generally a normal reload with no unusual core features and characteristics. Technical Specification changes are few and primarily related to Maximum Average Planar Linear Heat Generation (MAPLHGR) fuel, and Minimum Critical Power Ratio (MCPR) g Rate (LHGR) limit for the new.
limits for the new fuel, Linear Heat Generatin limits for all of the fuel using Cycle 13 core and transient parameters and extended operating regions and condi-tions. The new fuel is the extended burnup type which has been used in several recent BWR reloads (see, for example, Reference 3).
The Cycle 13 reload submittal includes a number of operating flexibility options:
single loop operation (SLO), load line limit analysis (LLLA), extended load line limit analysis (ELLLA), and the Average Power Range Monitor (APRM)/ Rod Block Monitor / Technical Specification improvement program (ARTS). The effects of these i
operating flexibility options have been included in the Cycle 13 reload safety l
analysis.
2.0 EVALUATION 2.1 Reload Description The Monticello Cycle 13 reload will retain 240 P8x8R and 124 BP8x8R GE fuel assemblies from the previous cycle and add 120 new GE8x8EB fuel assemblies.
8712040200 871125' PDR ADOCK 05000263 P
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. The reload safety analysis is based on a previous cycle core nominal average exposure of 21.8 GWd/MTU and Cycle 13 end of cycle core nominal average exposure of 22.3 GWd/MTV. The loading will be a conventional scatter patter.n with low reactivity fuel on the periphery. This loading is acceptable.
2.2 Fuel Design l
The new fuel for Cycle 13 is the GE extended burnup fuel GE8x8EB. The fuel designation is BD319B. This fuel type has been approved in the Safety Eval-uation Report for Amendment 10 to GESTAR-II (see Refs. 4 and 5). The specific description of this fuel has been submitted in Amendment 18 to GESTAR-II but, since this amendment has not yet been accepted, the fuel description has also been presented for Monticello Cycle 13 in Reference 6.
This fuel description is acceptable.
In operation, the GE8x8EB fuel will be assigned a number of axial lattice regions.
Appropriate MAPLHGR limits to a maximum value of 45,000 MWD /STU, which have been determined by approved thermal-mechanical and loss-of-coolant accident (LOCA) analyses, will be applied to each of these regions. There was extensive interaction among the staff, GE, and a number of utilities in deciding on an acceptable format for presentation of this information, suitable for plant use,
and staff requirements for Technical Specifications. Reference 7 provides an example of the Technical Specification for multiple lattice fuel bundles.
Reference 16 provides the NSP version of the Technical Specification. The Technical Specification agreed to by the staff, GE and certain utilities presents the least and most limiting lattice MAPLHGR as a function of burnup. However, the process computer used by the licensee contains, and acts on, full details of the MAPLHGR information. When hand calculations of MAPLHGR are required (process computer is inoperative), the most limiting MAPLHGR values as a function of burnup are used as limits for all the lattices of that bundle type. The proposed Technical Specification is acceptable although NSP does not irclude the least limiting lattice MAPLHGR as a function of burnup)in its version of the Technical Specification. A proprietary report (Ref. 6, reviewed by the staff and available to the NSP staff, provides complete details of the lattice "
definitions and MAPLHGR limits.
The proposed LHGR limit for the GE8x8EB fuel is 14.4 kW/ft rather than the 13.4 kW/ft for other GE fuel. This LHGR has been reviewed and accepted for this l
I fuel in the GE extended burnup fuel review and meets the criteria for fuel material design set forth in SRP 4.2 (Ref. 4).
(See the Reference 9 referrals to References 18 and 19. These references are responses to questions and presen-tations relating to the GE8x8EB fuel which provide information on the 14.4 kW/ftLHGR). This LHGR is acceptable for the GE8x8EB fuel in Monticello Cycle 13.
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Reference 2 states that not all the fuel channels to be used in Cycle 13 were supplied by GE but that GE, at the direction of the licensee, assumed that the l
performance characteristics of the non-GE fuel channels are identical to the characteristics of the channels supplied by GE. The staff has previously approved l
the use of non-GE fuel channels for Cycle 11 and these channels have been used I
at Monticello with no adverse effects. The staff concludes, therefore, that l
the use of non-GE fuel channels is acceptable.
. 2.3 Nuclear Desig,r The nuclear design for Monticello Cycle 13 has been performed.by GE with the approved methodology described in GESTAR-II (Ref. 5). The results of these i
analyses are given in the GE reload report (Ref. 2) in standard GESTAR-Il fomat. The results are within the range of those usually encountered for BWR reloads.
In particular, the shutdown margin is 1.6% and 1.0% delta K at beginning-of-cycle (B0C)andattheexposureofminimumshutdownmargTbf
)
i respectively, thus fully meeting the Technical Specification required amount of 0.25% delta K The Standby Liquid Control System (SLCS) also meets shutdownrequireNbtswithashutdownmarginof4.3%deltaK Since these andotherMonticelloCycle13nucleardesignparametershave@enobtained i
with previously approved methods and fall within expected ranges, the nuclear design is acceptable, i
2.4 Thermal-Hydraulic Desien The thermal-hydraulic design for Monticello Cycle 13 has been performed by GE with the approved methodology described in GESTAR-II (Ref. 5) and the result _s are given in the GE reload report (Ref. 2). The parameters used for the analyses are those approved in Reference 5 for the Monticello class BWR/3 unless otherwise indicated in Reference 2.
The GEMINI system of methods (approved in Ref. 9) was used for the relevant transient analyses, i
The Operating Limits MCPR (0LMCPR) values are determined by the limiting j
transients, which are usually the local Rod Withdrawal Error (RWE) and the core-wide transients Feedwater Controller Failure (FWCF), Loss of Feedwater Heating (LFWH) and turbine trip without bypass (TTWOBP). The analyses of the FWCF and TTWOBP events for Monticello Cycle 13 used the standard, approved (Ref. 5) ODYN Options A and B approaches for pressurization transients.
However, the RWE is limiting with a minimum CPR of 1.30 (Ref. 10) for Rod Block Monitor (RBM) upscale setpoints of 120, 115 and 110 for the low, intermediate and high trip setpoints, respectively. The Monticello Cycle 13, Technical Specifications will not require OLMCPR's, as a function of average scram time, for operation in both standard and extended operating regions.
The Technical Specification OLMCPR is 1.30 with other changes made to remove the dependency on scram speed. Approved methods (Ref. 5) were used to analyze these events and the analyses and results are acceptable and fall within expected ranges.
The results of thermal-hydraulic analyses for two recirculation loop operation show that the maximum core stability decay ratio was 0.63 for Monticello Cycle 11 and Cycle 10. Since Monticello is a BWR/3 with a conventional fuel design, the staff concluded in its Safety Evaluation (SE) that no additional stability analysis was required for Cycle 12. NSP states that the new GE8x8EB fuel has only a small impact on stability performance. The staff agrees with this assessment because of the similarities in the nuclear parameters of the new fuel (e.g., gap conductance, void coefficient) as compared to the previously used fuel. Therefore, Monticello Cycle 13 is typical of previous reload cores which have acceptable stability margin. The staff concludes that Monticello Cycle 13 is acceptable for two recirculation loop operation since it conforms to the staff position of Generic Letter 66-02 (Ref. II) for BWR/3's.
. Thermal-hydraulic stability for single loop operation for Monticello has been addressed and found acceptable in a staff SE for a previous amendment (Ref.12). Monticello has Technical Specifications which set.forth the limiting conditions of operation and surveillance requirements in confomance with the guidance proposed by GE in Service Information Letter (SIL) No. 380, Revision 1 (Ref. 13). Therefore, no additional analyses are required for establishing themal-hydreulic stability for single loop operation.
2.5 Transient and Accident Analyses The transient and accident analysis methodologies used for Monticello Cycle 13 are described in the NRC approved GESTAR-II (Ref. 5). The GEMINI system of methods (Ref. 9) option was used for the transient analyses. The limiting MCPR events for Monticello Cycle 13 are indicated in Section 2.4 above. The core wide transient analysis methodologies and results are acceptable and fall within expected ranges.
The rod withdrawal error is analyzed in the Average Power Range Monitor, Pod Block Monitor and Technical Specification Improvements (ARTS) program topical report (Ref. 14), which has been approved by the staff. A recently apprcved.
submittal supports the upscale setpoint changes for a RWE MCPR limit of 1.30 (Ref. 15). The mislocated assembly event was not analyzed for Monticello Cycle 13 since the event is less limiting than for an initial core. This is acceptable since this position was approved by the staff in Reference 5.
The I
disorientation event was analyzed with standard methods for the Monticello Cycle 13 D lattice (non-symmetric water gaps) fuel, giving a non-limiting i
l value of MCPR. The local transient event analyses are thus acceptable.
Thelimitingpressurizationevent,themainsteamisolationvalve(MSIV) closure with flux scram, analyzed with standard GESTAR-II methods, gave results for peak steam dome and vessel pressures well under the limits l
required by ASME Code Section III for upset conditions (i.e., 110% of l
design pressure - 1375 psi). These are acceptable methodologies and results,.
The licensee's submittal indicates that LOCA analyses, using approved methodologies (SAFE /REFLOOD/ CHASTE) and parameters, were performed using FAPLHGR values for the new reload fuel bundles (GE8xBEB). These results were within the limits of 10 CFR 50.46 and are, therefore, acceptable.
Since banked position withdrawal sequence rod patterns are used for Monticello, a cycle specific control rod drop accident analysis is not required. The basis for this position and NRC approval is presented in Amendment 9 to Reference 5.
l 2.6 Technical Spec 4*ications i
The Technical Specification (TS) changes for Monticello Cycle 13 are to provide for:
(a) The14.4kW/ftLHGRlimitforthenew(GE8x8EB) fuel.
The change is to TS 3.11.B. The time to initiate corrective action in the ACTION statement has been changed to correspond to Standard Technical l
Specifications (STS). These changes are acceptable.
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The reload safety analysis is based on a previous cycle core nominal average exposure of 21.8 GWd/MTU and Cycle 13 end of cycle core nominal average exposure of 22.3 GWd/MTV. The loading will be a conventional scatter patter.n with low reectivity fuel on the periphery. This loading is acceptable.
2.2 Fuel Design
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The new fuel for Cycle 13 is the GE extended burnup fuel GE8x8EB. The fuel designation is BD3198. This fuel type has been approved in the Safety Eval-uation Report for Amendment 10 to GESTAR-II (see Refs. 4 and 5). The specific description of this fuel has been submitted in Amendment 18 to GESTAR-II but, since this amendment has not yet been accepted, the fuel description has also j
i been presented for Monticello Cycle 13 in Reference 6.
This fuel description
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is acceptable, j
l In operation, the GE8x8EB fuel wili be assigned a number of axial lattice regions.
1 Appropriate MAPLH5R limits to a maximum value of 45,000 MWD /STU, which have l
been determined by approved thermal-mechanical and loss-of-coolant accident 1
(LOCA) analyses, will be applied to each of these regions. There was extensive interaction among the staff, GE, and a number of utilities in deciding on an I
acceptable format for presentation of this information, suitable for plant use.
and staff requirements for Technical Specifications. Reference 7 provides an i
example of the Technical Specification for multiple lattice fuel bundles.
Reference 16 provides the NSP version of the Technical Specification. The Technical Specification agreed to by the staff, GE and certain utilities presents the least and most limiting lattice MAPLHGR as a function of burnup. However, the process computer used by the licensee contains, and acts on, full details of the MAPLHGR information. When hand calculations of MAPLHGR are required (process computer is inoperative), the most limiting MAPLHGR values as a function of burnup are used as limits for all the lattices of that bundle type. The proposed Technical Specification is acceptable although NSP does not include the least limiting lattice MAPLHGR as a function of burnup)in its version of the Technical Specification. A proprietary report (Ref. 6, reviewed by the staff and available to the NSP staff, provides complete details of the lattice ~
definitions and MAPLHGR limits.
The proposed LHGR limit for the GE8x8EB fuel is 14.4 kW/ft rather than the 13.4 kW/ft for other GE fuel. This LHGR has been reviewed and accepted for this fuel in the GE extended burnup fuel review and meets the criteria for fuel l
material design set forth in SRP 4.2 (Ref. 4).
(See the Reference 9 referrals l
to References 18 and 19. These references are responses to questions and presen-tations relating to the GE8x8EB fuel which provide information on the 14.4 l
kW/ftLHGR). This LHGR is acceptable for the GE8x8EB fuel in Monticello Cycle 13.
l Reference 2 states that not all the fuel channels to be used in Cycle 13 were supplied by GE but that GE, at the direction of the licensee, assumed that the performance characteristics of the non-GE fuel channels are identical to the characteristics of the channels supplied by GE. The staff has previously approved the use of non-GE fuel channels for Cycle 11 and these channels have been used at Monticello with no adverse effects. The staff concludes, therefore, that the use of non-GE fuel channels is acceptable.
. 2.3 Nuclear Desian The nuclear design for Monticello Cycle 13 has been performed.by GE with the approved methodology described in GESTAR-II (Ref. 5). The results of these analyses are given in the GE reload report (Ref. 2) in standard GESTAR-II fomat. The results are within the range of those usually encountered for BWR reloads.
In particular, the shutdown margin is 1.6% and 1.0% delta K at beginning-of-cycle (B0C)andattheexposureofminimumshutdownmargT[i,f respectively, thus fully meeting the Technical Specification required amount of 0.257 delta K The Standby Liquid Control System (SLCS) also meets shutdown require $ts with a shutdown margin of 4.3% delta K Since these and other Monticello Cycle 13 nuclear design parameters have$e.n obtained with previously approved methods and fall within expected ranges, the nuclear design is acceptable.
2.4 Thermal-Hydraulic Desian The thermal-hydraulic design for Monticello Cycle 13 has been performed by GE with the approved methodology described in GESTAR-II (Ref. 5) and the results are given in the GE reload report (Ref. 2). The parameters used for the analyses are those approved in Reference 5 for the Monticello class BWR/3 unless otherwise indicated in Reference 2.
The GEMINI system of methods (approved in Ref. 9) was used for the relevant transient analyses.
The Operating Limits MCPR (0LMCPR) values are determined by the limiting i
transients, which are usually the local Rod Withdrawal Error (RWE) and the core-wide transients Feedwater Controller Failure (FWCF), Loss o' Feedwater Heating (LFWH) and turbine trip without bypass (TTWOBP). The analyses of the FWCF and TTWOBP events for Monticello Cycle 13 used the standard, approved (Ref. 5) ODYN Options A and B approaches for pressurization transients.
However, the RWE is limiting with a minimum CPR of 1.30 (Ref. 10) for Rod Block Monitor (RBM) upscale setpoints of 120, 115 and 110 for the low, intermediate and high trip setpoints, respectively. The Monticello Cycle 13, l
Technical Specifications will not require OLMCPR's, as a function of average scram time, for operation in both standard and extended operating regirns.
The Technical Specification OLMCPR is 1.30 with other changes made to remove the dependency on scram speed. Approved methods (Ref. 5) were used to analyze these events and the analyses and results are acceptable and fall within expected ranges.
The results of thermal-hydraulic analyses for two recirculation loop operation i
show that the maximum core stability decay ratio was 0.63 for Monticello l
Cycle 11 and Cycle 10. Since Monticello is a BWR/3 with a conventional fuel design, the staff concluded in its Safety Evaluation (SE) that no additional stability analysis was required for Cycle 12.
NSP states that the new GE8x8EB fuel has only a small impact on stability performance. The staff agrees with this assessment because of the similarities in the nuclear parameters of the new fuel (e.g., gap conductance, void coefficient) as compared to l
the previously used fuel. Therefore, Monticello Cycle 13 is typical of previous reload cores which have acceptable stability margin. The staff j
concludes that Monticello Cycle 13 is acceptable for two recirculation loop
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operation since it conforms to the staff position of Generic Letter 86-02 (Ref. 11) for BWR/3's.
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i Thermal-hydre.ulic stability for single loop operation for Monticello has been addressed and found acceptable in a staff SE for a previous amendment (Ref.12). Monticello has Technical Specifications which set.forth the limiting conditions of operation and surveillance requirements in conformance with the guidance proposed by GE in Service Information Letter (SIL) No. 380, i
Pevision 1 (Ref. 13). Therefore, no additional analyses are required for establishing thermal-hydreulic stability for single loop operation.
2.5 Transient and Accident Analyses j
i The transient and accident analysis methodologies used for Monticello Cycle l
13 are described in the NRC approved GESTAR-II (Ref. 5). The GEMINI system 1
of methods (Ref. 9) option was used for the transient analyses. The limiting MCPR events for Monticello Cycle 13 are indicated in Section 2.4 above. The core wide transient analysis methodologies and results are acceptable and fall within expected ranges.
l The rod withdrawal error is analyzed in the Average Power Range Monitor, Rod l
Block Monitor and Technical Specification Improvements (ARTS) program topical report (Ref. 14), which has been approved by the staff. A recently approved.
I submittal supports the upscale setpoint changes for a RWE MCPR limit of 1.30 l
(Ref. 15). The mislocated assembly event was not analyzed for Monticello Cycle 13 since the event is less limiting than for an initial core. This is acceptable since this position was approved by the staff in Reference 5.
The disorientation event was analyzed with standard methods for the Monticello Cycle 13 D lattice (non-symmetric water gaps) fuel, giving a non-limiting value of MCPR. The local transient event analyses are thus acceptable.
The limiting pressurization event, the main steam isolation valve (MSIV) closure with flux scram, analyzed with standard GESTAR-II methods, gave results for peak steam dome and vessel pressures well under the limits i
required by ASME Code Section III for upset conditions (i.e., 110% of design pressure - 1375 psi). These are acceptable methodologies and results,.
The licensee's submittal indicates that LOCA analyses, using approved methodologies (SAFE /REFLOOD/ CHASTE) and parameters, were performed using MAPLHGR values for the new reload fuel bundles (GE8x8EB). These results were i
within the limits of 10 CFR 50.46 and are, therefore, acceptable.
1 Since banked position withdrawal sequence rod patterns are used for Monticello, a cycle specific control rod drop accident analysis is not required. The basis for this position and NRC approval is presented in Amendment 9 to Reference 5.
2.6 Technical Specifications The Technical Specification (TS) changes for Monticello Cycle 13 are to provide for:
(a) The 14.4 kW/ft LHGR limit for the new (GE8x8EB) fuel.
The change is to TS 3.11.B.
The time to initiate corrective action in the ACTION statement has been changed to correspond to Standard Technical Specifications (STS). These changes are acceptable.
)
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- (b) MAPLHGR limits for the fuel..
The changes, which were revised, in.part, in References 8 dnd 16, are to TS 3.11.A. Table 3.11.1, and MAPLHGR's for fuel that is no longer used have been deleted. MAPLHGR's for the new fuel and future fuel have been included in Tabic 3.11.1.
The time to initiate corrective action in the ACTION statement has been changed to correspond to STS.
l These changes are acceptable.,
(c) The new MCPR limits for Cycle 13.
Since the proposed TS MCPR limit of 1.30 is higher than the Option A and B limits, all references to MCPR varying as a function of scram time are deleted. The changes are to Table of Contents Item 3.11.2, TS 3.3.C.3, Bases 3.3 and 4.3, TS 3.11.C, Table 3.11.2, and Bases 3.11.C.
Bases 3.11.C will now state that the LOCA analyses assumes a MCPR of 1.24, thus correcting an error. The time to initiate corrective action in the ACTION statement has been changed to correspond to STS. All of these changes are acceptable.
(d) Single loop operation surveillance power / flow curve.
Figure 3.5.1 has been redrawn to more clearly define for the operators the permissible operating regimes. This change is acceptable.
(e) Reactor Design Features.
TS 5.2.B has been rewritten in a more general manner so that control rods whose design has been reviewed and approved by the NRC may be used by Monticello. This change is acceptable.
I Each of the above changes has been previously discussed and approved in this review except for items (d) and (e), which are acceptable for the reasons I
stated above.
3.0 ENVIRONMENTAL CONSIDERATION
l This amendment involves a change to a requirement with respect to the in-sta11ation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation j
exposure. The Commission has previously issued a proposed finding that this i
amendment involves no significant hazards consideration and there.has been i
no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental
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l assessment need be prepared in connection with the issuance of this amendment.
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. L (b) MAPLHGR limits for the fuel.
The changes, which were revised, in part, in References 8 dnd 16, are to TS 3.11.A. Table 3.11.1, and MAPLiGR's for fuel that is no longer used have been deleted. MAPLHGR's for the new fuel and future fuel have been included in Tabic 3.11.1.
The time to initiate corrective action in the ACTION statement has been changed to correspond to STS.
These changes are acceptable.
(c) The new MCPR limits-for Cycle 13.
Since the proposed TS MCPR limit of 1.30 is higher' than' the Option A and B limits, all references to MCPR varying as a function of scram time are-deleted. The changes are to Table of Contents Item 3.11.2, TS 3.3.C.3, Bases 3.3 and 4.3, TS 3.11.C, Table 3.11.2, and Bases 3.11.C.
Bases 3.11.C will now state that the LOCA analyses assumes a MCPR of.1,.24, thus.
correcting an error. The time to initiate corrective action in the ACTION.
statement has been changed to correspond to STS. All.of these changes-are acceptable.
I (d) Single loop operation surveillance power / flow curve.
Figure 3.5.1 has been redrawn to more clearly define for the operators the permissible operating regimes. This change is acceptable.
(e) Reactor Design Features.
TS 5.2.B has been rewritten in a more general manner so that control rods whose design has been reviewed and approved by the NRC may be used by Monticello. This change is acceptable.-
Each of the above changes has been previously discussed and approved in this review except for items (d) and (e), which are acceptable for the reasons stated above.
3.0 ENVIRONMENTAL CONSIDERn..qN, This amendment involves a change to a requirement with respect to the in-sta11ation or use of a facility component located within the' restricted area as defined in 10 CFR Part 20. The staff has determined that-the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released' offsite and that there is no significant increase in individual or cumulative occupational radiation exposure..The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has.been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set'forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact-statement or environmental assessment need be prepared in connection with the' issuance of this amendment.
4.0 CONCLUSION
S I
I The staff has reviewed the reports submitted for the Cycle 13 operation of Monticello with extended operating regions.
Based on this review, it is concluded that appropriate material was submitted and that the fuel design, nuclear design, thermal-hydraulic design and transient and accident analyses are acceptable. The Technical Specification changes submitted for this reload suitably reflect the necessary modifications for operation in this cycie.
l 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 manner, and (2) public such activities will be conducted in ecmpliance with the Commission's regulations, and the issuance of the amendment will not be inimical to the common defense l
and security or to the health and safety of the public.
Principal Contributor: Dan Fieno l
Dated: November 25, 1987 I
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I REFERENCES 1.
Letter and enclosure from David Musolf (NSP) to NRC, dated. July 27, 1987. Application requesting changes to the Monticello Technical Specifications for Cycle 13 operation.
2.
GE Report 23A5827, Revision 0, dated June 1987, " Supplemental Reload-Licensing Submittal for Monticello Nuclear Generating Station, Cycle 13."
3.
Letter (and enclosure) from R. Clark (NRC) to E. Bauer (PEC), June 1987
-(Cycle 8 core reload for Peach Bottom Unit 2).
4.
Letter (and attechn.ent) frorr, C. Thomas (NRC) to J. Charnley'(GE) dated i
May 28, 1985, " Acceptance for Referencing of Licensing Topical Report
.i NEDE-24011-P-A-6, Amendment 10."
5.
GESTAR-II, NEDE-24011, Revision l8, " General Electric Standard Application.
a for Reactor Fuel."
d 6.
Letter and enclosure from David.Musolf-(NSP) to NRC, dated August 28, 1987. The enclosure dated August 1987 is NEDE-24050-2, Supplement 2,
" Supplement 2 to Loss-of-Coolant Accident Analysis for Monticello Nuclear Generating Plant."
.(
I 7.
Letter from J. Charnley (GE) to M. W. Hodges (NRC) dated March 4, 1987, l
" Recommended MAPLHGR Technical Specifications for Multiple Lattice Fuel Designs."
i l
8.
Letter and enclosure from David Musolf (NSP) to-NRC, dated September 3 1987. The enclosure presents a revised wording for the MAPLHGR Technical Specification.
9.
Letter (and attachment) from G. Lainas (NRC) to J. Charnley (GE) dated
- March 22, 1986, " Acceptance for Referencing of-Licensing Topical' Report, i
NEDE-24011-P-A, 'GE Generic Licensing Reload Report,' Supplement to Amendment 11."
j
- 10. Letter and enclosure from David Musolf (NSP) to NRC,' dated February 4 1987. The enclosure provides RBM setpoints for a CPR of 1.30 for the rod withdrawal error event.
- 11. Generic letter No. 86-02, " Technical Resolution of Generic Issue B-19-Thermal-Hydraulic Stability," January 23, 1986.
- 12. Letter from John Zwolinski (NRC) to David Musolf (NSP) dated October 22, 1986. The letter transmitted the staff SE on. Amendment 47 for Single Loop Operation at Monticello.
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.13.
General Electric Service Infomation Letter No. 380, Revision 1.-
February 10, 1984.
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.s.
- 14. " General Electric BWR Licensing Report: Average Power Range Monitor, Rod i
BlockMonitorandTechnicalSpecificationImprovement(ARTS)Programfor l
Monticello Nuclear Generating Plant," NEDC-30492-P, April 1984.
1
- 15. Letter from Dino Scaletti (NRC) to David Musolf (NSP) dated August 26, 1987. The letter transmitted the staff SE on Amendment 49 for Rod Block j
' Monitor setpoint changes.
J
- 16. Letter and enclosure from David Musolf (NSP) to NRC,' dated' September 16, 1987. The enclosure presents revised wording for the MAPLHGR Technical Specification, the LHGR Technical Specification revision to accommodate the new reload fuel, and other changes.
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