ML20138G100

Safety Evaluation Supporting Amend 107 to License DPR-21

ML20138G100
Person / Time
Site:	Millstone
Issue date:	12/06/1985
From:	Office of Nuclear Reactor Regulation
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ML20138G064	List: ML20138G060 ML20138G082 ML20138G100
References
NUDOCS 8512160237
Download: ML20138G100 (17)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 107 TO PROVISIONAL OPERATING LICENSE NO. DPR-21 NORTHEAST NUCLEAR ENERGY COMPANY MILLSTONE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET N0. 50-245

1.0 INTRODUCTION

By letter dated August 26, 1985 (Ref. 1), the Northeast Nuclear Energy Company (NNECO) made application to amend the Technical Specifications of the Operating License for Millstone Nuclear Power Staiton, Unit 1, in order to operate for Cycle 11. In support of this application, the licensee also provided a Supplemental Reload Licensing Submittal (Ref. 2). The Core Performance Branch has reviewed these submittals. The staff evaluation follows.

2.0 FUEL DESIGN EVALUATION The reload application involves two fuel-design related issues: (1) the replacement of 200 spent fuel assemblies with fresh BP8x8R fuel assemblies, (2) the analysis of safety considerations involved in the determination of

~ Cycle 11 operating limits.

2.1 Fuel Failure and Cycle 11 Core Inventory The Millstone-1 Cycle 11 core contains 580 fuel assemblies of which 200 were changed during the November 1985 Reload 10 outage.

The Cycle 11 core composition is sumarized in Table 1. The 200 replacement assemblies for Cycle 11 operation are fresh fuel assemblies (BP8DRB300),

which are prepressurized 8x8 retrofit barrier fuel assemblies with an average enrichment of 3.00 w/o in U-325. Since (1) the prepressurized 8x8 retrofit barrier fuel has been previously approved (Ref. 3), (2) the average enrichment of the fresh fuel is less than that of the approved maximum enrichment stated in reference 3, and (3) the MAPLHGR limits are established to avoid violation of the peak cladding temperature limit (2200 F) during a loss-of-coolant accident, the staff has concluded that the fuel l

assemblies are acceptable for Millstone-1 Cycle 11 operation. The existing types of fuel assemblies in the core as specified in the following table have been approved previously for application in Millstone-1; the design aspects of these fuel assemblies require no further NRC review.

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Table 1.--MILLSTONE-1 CORE INVENTORY Assembly Designation Cycle Loaded Number P8DRB282 (Irradiated) 9 72 P80RB283H(Irradiated) 9 108 BP8DRB300 (Irradiated) 10 200 BP8DRB300 (New) 11 200 2.2 Cycle 11 Operating Limits The licensee's determination of Cycle 11 operating limits is presented in the reload safety analyses (Ref. 2). In all fuel-design-related areas except those separately identified, the reload report relies on the generic report, General Electric Standard Application for Reactor Fuel (Ref. 3). Reference 3 has been reviewed and approved by the NRC staff. We conclude that additional staff review of those portions of Reference 3 concerning the standard fuel design is unnecessary for the Cycle 11 application.

2.3 MAPLHGR Limits I' The licensee's submittal provided MAPLHGR limits. These limits were generated by methods previously approved (Refs. 3, 4).

We conclude that the MAPLHGRS in the renumbered Figures 3.11.la through 3.11.c continue to be acceptable for cycle 11 operation.

2.4 Fuel Design Conclusions We have concluded that the licensee has adequately described the Millstone-1 Cycle 11 fuel design and its predicted conformance to the applicable regulations and NRC staff positions.

3.0 THERMAL AND HYDRAULIC DESIGN EVALUATION The objective of the review is to confirm that the thermal-hydraulic design of the core has been accomplished using acceptable methods, and provides an acceptable margin of safety from conditions which could lead to fuel damage during normal and anticipated operational transients, and is not susceptible to thermal-hydraulic instability.-

The review includes the following areas: (1) safety limit minimum critical power ratio (MCPR), (2) operating limit MCPR, and (3) thermal-hydraulic stability.

l The licensee has submitted the analysis report for Cycle 11 operation in Reference 2. Discussion of the review concerning the thermal-hydraulic design for Cycle 11 operation follows:

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3.1 Safety Limit MCPR A safety limit MCPR has been imposed to assure that 99.9 percent of the fuel rods in the core are not expected to experience boiling transition during normal and anticipated operational transients. As stated in

. Reference 3, the approved safety limit MCPR is 1.07. The safety limit MCPR of 1.07 is used for the Millstone-1 Cycle 11 operation.

3.2 Operating Limit MCPR The most limiting events have been analyzed by the licensee to detemine which event could potentially induce the largest reduction in the initial critical power ratio (WCPR). The WCPR values given in Section 10 of Reference 2 are plant specific values calculated by the approved methods including ODYN methods. The calculated WCPR values are adjusted to reflect the calculation uncertainties by employing the same conversion methods used during Cycle 10 (Ref. 8). The operating limit MCPR values are determined by adding the adjusted WCPRs to the safety limit MCPR. Section 12 of Reference 2 presents both the Cycle-11 MCPR values of the pressurization and non-pressurization transients. The maximum cycle MCPR values in Section 12 are specified as the operating limit MCPRs and incorporated into the Technical Specifications. The value of operating limit MCPR resulting from the limiting transient, the generator load rejection without bypass transient, is 1.48 for Cycle 11, which is the same value as that for Cycle 10. Since (1) the approved method was used to determine the operating limit MCPRs to avoid violation of the safety limit MCPR in the event of any anticipated transient, and (2) the Cycle 11 operating limit MCPRs are the same as that for Cycle 10, which was previously approved, we conclude that these limits are acceptable. This conclusion applies to both options (A and B) of Technical Specifications Table 3.11.1. The difference between 1.43 for cycle 10 and 1.42 for cycle 11 option B is attributed to the method of rounding off the numerical value.

3.3 Thermal-Hydraulic Stability The Millstone-1 Cycle 11 core meets the thermal-hydraulic stability criteria stated in Reference 5 since it is loaded with an approved fuel design and is a BWR 3. Therefore, we conclude that the thermal-nydraulic stability aspects are acceptable for Cycle 11 operation.

3.4 Technical Specifications for the Operating Limit MCPRt Table 3.11.1 in Section 3/4.11-10 of the proposed Technical Specifications (Ref.1) has included the operating limit MCPRs for operation of Cycle 11.

Based on our review, we find that the proposed operating MCPR limits have been established using approved thermal-hydraulic methods to avoid violation of the safety limit MCPR during normal operation and anticipated l operational occurrences. We, therefore, conclude that the Technical l Specification MCPR limits as indicated in Table 3.11.1 of Section 3/4.11-10 of the proposed Technical Specifications continue to be acceptable.

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- 4-3.5 Thermal-Hydraulic Design Conclusions 1 i

1 We have concluded that the licensee has adequately described the Millstone-1  :

j Cycle-11 thermal-hydraulic design and its predicted confonnance to the  !

j applicable regulations. Based on the above, we have determined that the thermal-hydraulic characteristics of the reload core for Millstone-1

Cycle-11 are acceptable.

4.0 NUCLEAR DESIGN The Millstone-1 Cycle 11 reload will consist of 580 fuel bundles as shown in Table 1. The reload fuel is similar in physical design to the initial core

! load fuel, but it has a maximum average enrichment of 3.00 w/o in U-235.

All fuel bundles consist of 62 fuel rods and 2 water rods. The active fuel length is 150 inches.

} The shutdown margin of the new core meets the Technical Specification requirement that the core be at least 0.33% WK subcritical in the most reactive condition when the highest worth control rod is fully withdrawn and all other rods are fully inserted. For Millstone-1 Cycle 11, GE i calculated that the K under cold conditions with the strongest rod out is equal to 0.979 resBNing in a shutdown margin of 2.1% WK.

The standby liquid control system is capable of bringing the reactor from

' full power to a cold shutdown condition assuming none of the withdrawn control rods are inserted. The 600 ppm boron concentration will bring the 1

reactor subcritical to k,ff = 0.954 at 20 C xenon free conditions (Ref. 2).

4.1 Nuclear Design Conclusions i

Based on our review of the licensee submittal (Ref.1) and the plant specific 1

analysis (Ref. 2), we have determined that the nuclear characteristics and the expected performance of the reload core for Millstone-1 Cycle 11 are acceptable.

5.0 Technical Specifications l

. In addition to changes in MAPLHGR and MCPR limits which were discussed previously the licensee has proposed changes (Ref. 7) to Section 5.5 of the. Technical Specifications concerning fuel storage. The proposed change would introduce a maximum kB (1.35) criterion for fuel placed in the spent fuel pool. The new criterion replaces the existing criterion that the U-235 loading be less than or equal to 15.2 gm/cm which was derived without taking credit for Gadolinia or reactivity. depletion due to burnup.

The staff has reviewed the proposed change to Section 5.5 and finds it

{ acceptable for the following reasons:

(1) The staff has previously reviewed and approved the change from the gm/cm to kB criterion for other BWR sites.

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(2) The applicant's submittal (Ref. 6) shows that its fuel racks can be acceptably loaded with fuel having a maximum kB value of 1.4 which is greater than its proposed value of 1.35.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment involves a change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance requirements. 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 Comission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public coment on such finding. Accordingly, this amendment meets the eligibility criteria for categorial 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.

7.0 CONCLUSION

S We have reviewed the reload safety analysis for Cycle 11 operation of Millstone-1 including the necessary changes to the Millstone-1 Technical Specifications, and we conclude that the application is acceptable. We find that the reactor may be reloaded and operated for Cycle 11 without undue risk to the public health and safety. Because margins of safety are e intain d and no change in operating modes is proposed, there is no increase in the probability or consequences of accidents or malfunctions of equip...ent important to safety.

The staff has concluded, based on ,ie considerations discussed above, that: (1) there is reasonable asswance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

8.0 REFERENCES

1 1.

Letter, 1985.

J. F. Opeka (NNECO) to J. L. Zwolinski (NRC), dated August 26, '

2. Plotycia, G. D., " Supplemental Reload t.icensing Submittal for Millstone-1 Unit 1, Reload 10 "GE Report-23A4696, August 1985.

3.

" General Electric Standard Application for Reactor Fuel' GE Report NEDE-240ll-P-A-6, April 1983.

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4. D. G. Eisenhut (NRC) letter to E. D. Fuller (GE), June 30, 1977.

5. Letter, C. O. Thomas, (NRC) to H. C. Pfefferlein (GE), " Acceptance for Referencing of Licensing Topical Report NEDE-24011, Rev. 6, Amendment 8, ' Thermal Hydraulic Stability Amendment to GESTAR II'",

April 24, 1985.

6. Letter, D. C. Surtzer (NNECO) to G. Jean (NRC), dated July 15, 1976, Docket No. 50-245.

7. Telecopy, Ed Bireley to Jim Shea, " Millstone Unit No.1 Reload 10 License Amendment Request," dated October 15, 1985.

8. Amendment No. 98 Letter, D. M. Crutchfield to W. G. Counsil, dated June 14, 1984 Dated: December 6, 1985 Principal Contributor:

George A. Schwenk

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6. Oxygen Concentration The relatively small containment volume inherent in the GE-BWR pressure suppression containment and the large, amount of zirconium in the core are such that the occurrence of a very limited (a percent of so) reaction of the zirconium and steam during a loss of coolant accident would lead to the liberation of sufficient hydrogen to result in a flammable concentration in the containment. Subsequent ignition of the .

hydrogen if it is present in sufficient quantities to result in excessively rapid recombination, could result in a loss of containment intergrity.

The 4% oxygen concentration minimizes the possibility of hydrogen combustion following a loss of coolant i

accident. Significant quantities of hydrogen could be generated if the core cooling systems did not sufficiently cool the core.

The occurrence of primary system leakage following a major refueling outage or other scheduled shutdown is

' more probable than the occurrence of the loss of coolant accident upon which the specified oxygen concentra-tion limit is based. . Permitting access to the drywell for leak inspections during a startup is judged prudent in terms of the added plant safety without significantly reducing the margin of safety. Thus to preclude the possibility of starting the reactor and operating for extended periods of time with significant l leaks in the primary system, leak inspections are scheduled during startup periods, when the primary system is at or near rated operating temperature and pressure. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period to provide inerting is ju.;ed to be sufficient to perform the leak inspection *

-establish the required oxygen concentration. The primary containment is normally slightly pressurized during periods of reactor operation assuring no air in-leakege through the primary containment. Ilowever, at least once a week, the oxygen concentration will be determined as added assurance.

l e j B. Standby Gas Treatment Systes-The standby gas treatment system is designed to filter and exhaust the reactor building atmosphere to t

the stack during secondary containment isolation conditions. Both standby gas treatment system fans i are designed to automatically start upon containment isolation and to maintain the reactor building pressure to the design negative pressure so that all leakage should be in-leakage. Each of the two

, fans has 100 percent capacity.

! High efficiency particulate absolute (NEPA) filters are installed before and after the charcoal absorbers to minimize potential release of particulates to the environment and to prevent clogging

! of the iodine absorbers. The charcoal absorbers are installed to reduce the potential release of

radiciodine to the environment. The in place test results should indicate a system leak tightness of less than 1 percent bypass leakage for*the charcoal absorbers and a llEPA efficiency of at least 99 percent removal of DOP particulates. The laboratory carbon sample test results should indicate a I

Ame,ndment JDA, 192, 107 .B 3/4 7-5

radiocctiva methyl iodida removal efficiency of at least 95 parcent for expacted accidznt cenditions.

i If the efficiencies of tbn llEPA filters and charcoal absorbers are as specified, the resulting doses will be less than the 10 CFR 100 guidelines for the accidents analyzed. Operation of the fans significantly different from the design flow will change the removal ef ficiency of the HEPA filters and charcoal absorbers.

Only one of the two standby gas treatment systems is needed to clean up the reactor building f tmo-sphere upon containment isolation. If one system is found to be inoperable, there is'no isenediate threat to the contaisument system performance and reactor operation or refueling operation may continue while repairs are being made. During refueling two off-site power sources (345KV or 27KV) and one emergency power source would provide an adequate and reliable source of power and allow i

! completion of annual diesel or gas turbine preventative maintenance.

C. Secondary Containment The result secondary containment from a serious is designed to minimize any ground level release of radioactive materials which might accident.

when the drywell is sealed and inThe reactor building provides secondary containment during reactor operation, service; the reactor building provides primary containment when the reactor is shutdown and the drywell is open, as during refueling. Because the secondary containment is an integral part of the required.

complete containment system, secondary containment is required at all times that primary containment is D. Primary Contairusent Isolation Valves Double isolation valves are provided on lines penetrating the primary containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression.

system. Automatic initiation is required to minimize the potential leakage paths from the contain-meat in the event of a loss of coolant accident.

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Amendment No. 8,19J,107 8 3/4 7-6 l .

5.0 DESIOi FEKIURES 5.11 Site The Unit I reactor building is located on the site at Millstone Point in Waterford, Connecticut.%e nearest site boundary on land is 1620 feet northeast of the reactor building, which is the minimtsu distance to the .

boundary of the exclusion area as described in 10 CFR 100.3(a). No part of the site which is closer to the-reactor building than 1620 feet shall be sold or leased except to (1) %e Connecticut Light and Power Cmpany, Western Massachusetts Electric Capany or Northeast Nuclear Energy Ccunpany or their corporate affiliates for use in conjunction with normal utility operations and (ii) to the two leasees under the leases referred to in the following paragraph.

A United States Navy research Laboratory and a desalination pilot operation of the Maximtsn Evaporator Division of the Cuno Engineering Corporation may be permitted to operate within the exclusion area under leases which make activities and persons on the leased premises subject to health and safety requirements of the owner of the site.

5.2 Reactor A. %e core shall consist of 580 fuel assenblies.

B. The reactor core shall contain 145 cruciform-shaped control rods. We control material shall be hafnitsu and/or boren carbide powder (B C) ccmpacted to approximately 70% of theoretical density.

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5.3 Reactor Vessel

%e reactor vessel shall be as described in Table IV-1 of the PSAR. %e applicable design codes shall be as described in Table IV-1 of the FSAR.

4 5.4 Containment A. The principal design paraneters and applicable design codes for the primary contairunent shall be as given in Table V-1 of the PSAR.

B.  % e secondary containment shall be as described in Section V-3 of the PSAR and the applicable codes shall

, be as described in section XII of the FSAR.

Amendment No. I6, 76, 9J,10/ 5-1

C.

designed in accordance with standards set forth in Section V-2 of the FSA .

5.5 Fuel Storage A. The new storage facility shall be such that the K is less than 0.95. 8ff dry is less than 0.90 and flooded B. The K value*f[ of the spent fuel storage pool shall be less than or equal to 0.90, This K bundles issatisfied less than if the maximum exposure - dependent K of the individual fuel 1.35. *ff 5.6 Seismic Design earthquake ground motion with an acceleration of 17% of gravity.The reactor b to determine reactor building.the earthquake acceleration applicable to the various elevations in theDynamic analysis was us Amendment No. 39, 107 5-2

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 107 TO PROVISIONAL OPERATING LICENSE NO. DPR-21 NORTHEAST NUCLEAR ENERGY COMPANY MILLSTONE NUCLEAR POWER STATION, UNIT NO. I DOCKET NO. 50-245

1.0 INTRODUCTION

By letter dated August 26, 1985 (Ref. 1), the Northeast Nuclear Energy Company (NNECO) made application to amend the Technical Specifications of the Operating License for Millstone Nuclear Power Staiton, Unit 1, in order to operate for Cycle 11. In support of this application, the licensee also provided a Supplemental Reload Licensing Submittal (Ref. 2). The Core Performance Branch has reviewed these submittals. The staff evaluation follows.

2.0 FUEL DESIGN EVALUATION The reload application involves two fuel-design related issues: (1) the replacement of 200 spent fuel assemblies with fresh BP8x8R fuel assemblies, (2) the analysis of safety considerations involved in the determination of

~ Cycle 11 operating limits.

2.1 Fuel Failure and Cycle 11 Core Inventory The Millstone-1 Cycle 11 core contains 580 fuel assemblies of which 200 were changed during the November 1985 Reload 10 outage.

The Cycle 11 core composition is sumarized in Table 1. The 200 replacement assemblies for Cycle 11 operation are fresh fuel assemblies (BP8DRB300),

which are prepressurized 8x8 retrofit barrier fuel assemblies with an average enrichment of 3.00 w/o in U-325. Since (1) the prepressurized 8x8 retrofit barrier fuel has been previously approved (Ref. 3), (2) the average enrichment of the fresh fuel is less than that of the approved maximum enrichment stated in reference 3, and (3) the MAPLHGR limits are established to avoid violation of the peak cladding temperature limit (2200 F) during a loss-of-coolant accident, the staff has concluded that the fuel assemblies are acceptable for Millstone-1 Cycle 11 operation. The existing l

types of fuel assemblies in the core as specified in the following table have been approved previously for application in Millstone-1; the design aspects of these fuel assemblies require no further NRC review.

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Table 1.--MILLSTONE-1 CORE INVENTORY Assembly Designation Cycle Loaded Number P8DR8282 (Irradiated) 9 72 P80RB283H(Irradiated) 9 108 BP8DRB300(Irradiated) 10 200 BP80RB300(New) 11 200 2.2 Cycle 11 Operating Limits The licensee's determination of Cycle 11 operating limits is presented in the reload safety analyses (Ref. 2). In all fuel-design-related areas except those separately identified, the reload report relies on the generic report, General Electric Standard Application for Reactor Fuel (Ref. 3). Reference 3 has been reviewed and approved by the NRC staff. We conclude that additional staff review of those portions of Reference 3 concerning the standard fuel design is unnecessary for the Cycle 11 application.

2.3 MAPLHGR Limits The licensee's submittal provided MAPLHGR limits. These limits were generated by methods previously approved (Refs. 3, 4).

We conclude that the MAPLHGRS in the renumbered Figures 3.11.la through 3.11.c continue to be acceptable for cycle 11 operation.

2.4 Fuel Design Conclusions We have concluded that the licensee has adequately described the Millstone-1 Cycle 11 fuel design and its predicted conformance to the applicable regulations and NRC staff positions.

3.0 THERMAL AND HYDRAULIC DESIGN EVALUATION The objective of the review is to confirm that the thermal-hydraulic design of the core has been accomplished using acceptable methods, and provides an acceptable margin of safety from conditions which could lead to fuel damage during normal and anticipated operational transients, and is not susceptible to thermel-hydraulic instability.

The review includes the following areas: (1) safety limit minimum critical pcwer ratio (MCPR), (2) operating limit MCPR, and (3) thermal-hydraulic stability.

The licensee has submitted the analysis report for Cycle 11 operation in Reference 2. Discussion of the review concerning the thermal-hydraulic design for Cycle 11 operation follows:

3.1 Safety Limit MCPR A safety limit MCPR has been imposed to assure that 99.9 percent of the fuel rods in the core are not expected to experience boiling transition during normal and anticipated operational transients. As stated in Reference 3, the approved safety limit MCPR is 1.07. The safety limit MCPR of 1.07 is used for the Millstone-1 Cycle 11 operation.

3.2 Operating Limit MCPR The most limiting events have been analyzed by the licensee to determine which event could potentially induce the largest reduction in the initial critical pcwer ratio (WCPR). The WCPR values given in Section 10 of Reference 2 are plant specific values calculated by the approved methods including ODYN methods. The calculated WCPR values are adjusted to reflect the calculation uncertainties by employing the same conversion methods used during Cycle 10 (Ref. 8). The operating limit MCPR values are detennined by adding the adjusted WCPRs to the safety limit MCPR. Section 12 of Reference 2 presents both the Cycle-11 MCPR values of the pressurization and non-pressurization transients. The maximum cycle MCPR values in Section 12 are specified as the operating limit MCPRs and incorporated into the Technical Specifications. The value of operating limit MCPR resulting from the limiting transient, the generator load rejection without bypass transient, is 1.48 for Cycle 11, which is the same value as that for Cycle 10. Since (1) the approved method was used to determine the operating limit MCPRs to avoid violation of the safety limit MCPR in the event of any

^

anticipated transient, and (2) the Cycle 11 operating limit MCPRs are the same as that for Cycle 10, which was previously approved, we conclude that these limits are acceptable. This conclusion applies to both options (A and B) of Technical Specifications Table 3.11.1. The difference between 1.43 for cycle 10 and 1.42 for cycle 11 option B is attributed to the method of rounding off the numerical value.

3.3 Thermal-Hydraulic Stability l

i The Millstone-1 Cycle 11 core meets the thermal-hydraulic stability criteria stated in Reference 5 since it is loaded with an approved fuel design and is a BWR 3. Therefore, we conclude that the thermal-hydraulic stability aspects are acceptable for Cycle 11 operation.

3.4 Technical Specifications for the Operating Limit MCPRs Table 3.11.1 in Section 3/4.11-10 of the proposed Technical Specifications (Ref.1) has included the operating limit MCPRs for operation of Cycle 11.

Based on our review, we find that the proposed operatir.g MCPR limits have i

been established using approved thermal-hydraulic methods to avoid violation of the safety limit MCPR during normal operation and anticipated operational occurrences. We, therefore, conclude that the Technical Specification MCPR limits as indicated in Table 3.11.1 of Section 3/4.11-10 of the proposed Technical Specifications continue to be acceptable.

l 3.5 Thermal-Hydraulic Design Conclusions i

We have concluded that the licensee has adequately described the Millstone-1 Cycle-11 thermal-hydraulic design and its predicted confonnance to the applicable regulations. Based on the above, we have detennined that the thermal-hydraulic characteristics of the reload core for Millstone-1 Cycle-11 are acceptable.

4.0 NUCLEAR DESIGN The Millstone-1 Cycle 11 reload will consist of 580 fuel bundles as shown in Table 1. The reload fuel is similar in physical design to the initial core load fuel, but it has a maximum average enrichment of 3.00 w/o in U-235.

All fuel bundles consist of 62 fuel rods and 2 water rods. The active fuel length is 150 inches.

The shutdown margin of the new core meets the Technical Specification requirement that the core be at least 0.33% WK subcritical in the most reactive condition when the highest worth control rod is fully withdrawn and all other rods are fully inserted. For Millstone-1 Cycle 11, GE calculated that the K under cold conditions with the strongest rod out is equal to 0.979 res6Ning in a shutdown margin of 2.1% WK.

The standby liquid control system is capable of bringing the reactor from full power to a cold shutdown condition assuming none of the withdrawn control rods are inserted. The 600 ppm boron concentration will bring the reactor subcritical to k,ff = 0.954 at 20*C xenon free conditions (Ref. 2).

4.1 Nuclear Design Conclusions Based on our review of the licensee submittal (Ref. 1) and the plant specific analysis (Ref. 2), we have determined that the nuclear characteristics and the expected performance of the reload core for Millstone-1 Cycle 11 are acceptable.

5.0 Technical Specifications In addition to changes in MAPLHGR and MCPR limits which were discussed previously the licensee has propnsed changes (Ref. 7) to Section 5.5 of the Technical Specifications concerning fuel storage. The proposed change would introduce a maximum kB (1.35) criterion for fuel placed in the spent fuel pool. The new criterion replaces the existing criterion that the U-235 loading be less than or equal to 15.2 gm/cm which was derived without

, taking credit for Gadolinia or reactivity depletion due to burnup.

The staff has reviewed the proposed change to Section 5.5 and finds it acceptable for the following reasons:

(1) The staff has previously reviewed and approved the change from the gm/cm to kB criterion for other BWR sites.

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(2) The applicant's submittal (Ref. 6) shows that its fuel racks can be acceptably loaded with fuel having a maximum kB value of 1.4 which is greater than its proposed value of 1.35.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment involves a change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance requirements. 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 Comission has previously issued a proposed j

' finding that this amendment involves no significant hazards consideration and there has been no public coment on such finding. Accordingly, this amendment meets the eligibility criteria for categorial exclusion set 3"

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.

7.0 CONCLUSION

S i

We have reviewed the reload safety analysis for Cycle 11 operation of Millstone-1 including the necessary changes to the Millstone-1 Technical Specifications, and we conclude that the application is acceptable. We find that the reactor may be reloaded and operated for Cycle 11 without undue risk to the public health and safety. Because margins of safety are maintained and no change in operating modes is proposed, there is no increase in the probability or consequences of accidents or malfunctions of equipment important to safety.

The staff has concluded, based on the considerations discussed above, i that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed

' manner, and (2) such activities will be conducted in compliance with

' the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

8.0 REFERENCES

1.

Letter, J. F. Opeka (NNECO) to J. A. Zwolinski (NRC), dated August 26,

1985.

2. Plotycia, G. D., " Supplemental Reload Licensing Submittal for Millstone-1 Unit 1, Reload 10 "GE Report-23A4696, August 1985.

3. " General Electric Standard Application for Reactor Fuel, GE Report NEDE-24011-P-A-6, April 1983.

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4. D. G. Eisenhut (NRC) letter to E. D. Fuller (GE), June 30, 1977. -

5. Letter, C. O. Thomas, (NRC) to H. C. Pfefferlein (GE), " Acceptance for Referencing of Licensing Topical Report NEDE-24011, Rev. 6 Amendment 8, ' Thermal Hydraulic Stability Amendment to GESTAR II'",

April 24, 1985.

6. Letter, D. C. Surtzer (NNECO) to G. Jean (NRC), dated July 15, 1976, Docket No. 50-245. .

7. Telecopy, Ed Bireley to Jim Shea, " Millstone Unit No.-l' Reload 10 License Amendment Request," dated October 15, 1985. '

8. Amendment No. 98 Letter, D. M. Crutchfield to W. G. Counsil, dated June 14, 1984

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Dated: December 6,1985 Principal Contributor:

George A. Schwenk i

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December 6,1985 Mr. John F. Opeka . -

A copy of our related Safety Evaluation is enclosed. The notice of issuance will appear in the Commission's biweekly Federal Register Notice.

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Sincerely, Original Signed By Chris Grimes Christopher I. Grimes, Director

. Integrated Safety Assessment Project Directorate Division of PWR Licensing - B

Enclosures:

1. Amendment No.107 to License No. DPR-21 2; Safety Evaluation cc w/ enclosures:

See next page DISTRIBUTION EDoctet F11eJ: % .3 '

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