ML25234A146
| ML25234A146 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 09/30/1976 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| Download: ML25234A146 (1) | |
Text
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1~,,;1l11;1titt11 l~e1ttt1*t related to operation of Supplement No. 4 U.S. Nuclear Regulatory Commission Brunswick Steam Electric Plant, Units 1 and 2 Office of Nuclear Reactor Regulation Docket Nos. 50-325 50-324 Carolina Power and Light Company Supplement No. 4 September 1976
SUPPLEMENT NO. 4 TO THE SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY COMMISSION IN THE MATTER OF CAROLINA POWER & LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PLANT UNITS 1 AND 2 DOCKET NOS. 50--325 AND 50-324 U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D. C.
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TABLE OF CONTENTS PAGE
1.0 INTRODUCTION
1-1 4.0 REACTOR...
4-1
- 4. 1 General.
4.2 Fuel Design..
4.3 Nuclear Design 4.4 General Electric Thermal Analysis Basis and Emergency Core Cooling System Analysis..
4.4. l Introduction.
4.4.2 General Electric Thermal Analysis Basis Evaluation.
4.4.2.1 Fuel Cladding Integrity Safety Limit Minimum Critical 4-1 4-l 4-3 4-4 4-4 4-4 Power Ratio.........
4-5 4.4.2.2 Operating Limit Minimum Critical Power Ratio 4-5 4.4.2.3 Local Event - Rod Withdrawal Error..
4-6 4.4.2.4 Operating Minimum Critical Power Ratio for Less than Rated Power and Fl ow....
4-7 4.4.3 Emergency Core Cooling Systems Appendix K Analysis.
4.4.4 Overpressure Analysis 4.4.5 Conclusions....
4.5 Modifications to Eliminate In-Core Vibration
- 4. 5.1 Introduction....
4.5.2 Brunswick Unit l Modifications.
- 6. 0 ENGINEERED SAFETY FEATURES........
- 6. 2.1 Containment Functional Design 6.2.1.3 Mark I Torus Loads.
7.0 INSTRUMENTATION AND CONTROLS.
7.2 Reactor Trip System...
7.2.5 Anticipated Transients Without Scram.
4-7 4-10 4-10 4-10 4-10 4-ll 6-l 6-l 6-l 7-l 7-l 7-l
TABLE OF CONTENTS (Continued)
PAGE
- 9. 0 AUX I LIA RY SYSTEMS............................... 9-l 9.1 Fuel Storage and Handling.
...................... 9-1 9.1.5 Reactor Building Crane.
- 9. 1.6 Residual Heat Removal Pumps 9-1 9-2 9.5 Other Auxiliary Systems...
...................... 9-1 9.5. l Fire Protection System.
9-3 11.0 RADIOACTIVE WASTE MANAGEMENT.
11-1 16.0 TECHNICAL SPECIFICATIONS...
16-1 APPENDIX A - CHRONOLOGY............................... A-1 APPENDIX B - SAFETY EVALUATION REPORT ON THE MODIFICATION TO ELIMINATE SIGNIFICANT IN-CORE VIBRATIONS IN OPERATING REACTORS WITH 1-INCH BYPASS HOLES IN THE CORE SUPPORT PLATE B-1 APPENDIX C - REFERENCES............................... C-1 ii
1.0 INTRODUCTION
The United States Nuclear Regulatory Corrmission (Corrmission) issued its Safety Evaluation Report of the Brunswick Steam Electric Plant Units land 2 in November 1973.
That Safety Evaluation Report was the first of the Corrmission's reports on the applica-tion by the Carolina Power and Light Company {applicant) for licenses to operate the Brunswick Steam Electric Plant Units l and 2.
Three supplements to that Safety Evalua-tion Report have been issued. Supplement No. 1, issued on January 31, 1974, described the status and resolution of matters reported as unresolved in the Safety Evaluation Report and in the report of the Advisory Committee on Reactor Safeguards.
On December 23, 1974, the Corrmission issued Supplement No. 2 which provided an update of the staff review and evaluation of Amendments 26, 27, 28, and 29 to the Final Safety Analysis Report.
These amendments contained modifications of two safety related systems, the overpressure protection system and the low pressure coolant injection system.
Our evaluation and acceptance of these two modifications is provided in Sup-plement No. 2.
Supplement No. 3 was issued on December 27, 1974, on the same date that the Brunswick Steam Electric Plant Unit 2 was issued operating license DPR-62.
Supple-ment No. 3 provided the evaluation, justification, and restrictions required for the Brunswick Steam Electric Plant Units land 2, to assure conformance with the require-ments of Section 50.46 of 10 CFR Part 50.
Our analysis of the emergency core cooling system performance for 7 x 7 fuel design based on an evaluation model that was wholly in conformance with Section 50.46 and Appendix K of 10 CFR Part 50 was included as an attachment to Amendment No. 5 to the Brunswick Steam Electric Plant Unit 2 operating license DPR-62.
The purpose of Supplement No. 4 is to update our Safety Evaluation Report and its previous three supplements prior to issuance of the operating license for Brunswick Steam Electric Plant Unit l. The scope of this supplement provides our review and evaluation of new information and design modifications to the Brunswick Steam Electric Plant Unit 1, as described in Amendments 30 and 31 to the Final Safety Analysis Report; our review and acceptance of the reactor building 125-ton crane design; and our review and acceptance of standard technical specifications for Brunswick Steam Electric Plant Unit 1.
The principal matters addressed in this Supplement No. 4 to the Safety Evaluation Report are:
(1)
Change of the fuel design from a 7 x 7 to an 8 x 8 fuel bundle array for the initial core loading in the Brunswick Steam Electric Plant Unit 1.
(2)
Adoption of a new method of analysis (GETAB) for defining the fuel damage limits for the 8 x 8 fuel design for Brunswick Steam Electric Plant Unit 1.
(3)
Modification to the core support plate and the lower tie plates of all fuel assembli es to eliminate significant in-core vibrations in Brunswick Steam Electric Plant Unit l.
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(4)
Evaluation of the short term program and the results for the Brunswick Steam Electric Plant Unit l design as affected by a postulated loss-of-coolant accident and the resulting hydrodynamic loads in the suppression pool.
(5)
Evaluation of the emergency core cooling system analysis required by Section 50.46 and Appendix K of 10 CFR Part 50 for the Brunswick Steam Electric Plant Unit l initial 8 x 8 fuel load design.
(6)
Evaluation of the protection provided for the residual heat removal system pumps to prevent pump run out following a postulated loss-of-coolant accident.
(7)
Evaluation and acceptance of the Brunswick Steam Electric Plant Unit l reactor building 125-ton crane design.
(8)
Change to standard technical specifications for the Brunswick Steam Electric Plant Unit l operation.
Subject to the inclusion of conditions related to Item (4) above (discussed in Section 5.0) in the operating license, we conclude that the Brunswick Steam Electric Plant Unit l can be operated without endangering the health and safety of the public.
Appendix A contains an updated chronology of significant milestones and documentation for the Brunswick Steam Electric Plant Unit l since the issuance of Supplement No. 3 to the Safety Evaluation Report.
Appendix B contains the detailed "Safety Evaluation Report on the Reactor Modifications to Eliminate Significant In-Core Vibration in Operating Reactors with 1-Inch Bypass Holes in the Core Support Plate," issued by the Office of Nuclear Reactor Regulation in February 1976.
Appendix C provides a listing of the references used by the staff in their evaluation contained in Section 4.0 of this report.
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4.0 REACTOR 4.1 General 4.2 We reported our evaluation of the fuel design for the initial core loadings for the reactors in Section 4.0 of our Safety Evaluation Report of the Brunswick Steam Electric Plant, Units 1 and 2 d~ted November 1973.
The initial core design reviewed in that evaluation consisted of 560 fuel assemblies.
Each fuel assembly consisted of a fuel bundle and the channel which surrounds it, and each fuel bundle contained 49 fuel rods which are spaced and supported in a square 7 x 7 array by the lower and upper tie plates.
Brunswick Steam Electric Plant Unit 2 is licensed to use fuel assemblies of the 7 x 7 array design.
Amendment 31 to the Final Safety Analysis Report dated November 26, 1975, described a revised fuel assembly design for the initial core loading in the Brunswick Steam Elec-tric Plant, Unit 1.
Amendment 31 and additional information provided by letters from the applicant provided the necessary information used in the staff's evaluation and acceptance of the 8 x 8 fuel design proposed for the Brunswick Steam Electric Plant Unit 1.
Our evaluation of the Brunswick Steam Electric Plant Unit 1 fuel assembly design and its use in the initial core loading is given in the following subsections of this report.
Fuel Design The revised fuel design for Brunswick Unit 1 as described in Amendment 31, November 1975, is the current General Electric 8 x 8 fuel assembly design.
The description of this fuel assembly in Section 3.9 of the Brunswick Final Safety Analysis Report is similar to that given in the General Electric Standard Safety Analysis Report and the General Electric generic reload topical report, "General Electric BWR Generic Reload Application for 8 x 8 fuel, "NEDD 20360, Revision 1, dated November 1974.
Mechanical and operating parameters for the 8 x 8 assemblies are compared to the 7 x 7 assemblies in Table I.
The smaller diameter rods, with lower linear heat generation rate and increased cladding thickness to diameter ratio for the 8 x 8 fuel design com-pared with the 7 x 7 fuel assemblies, result in increased safety margins with respect to maximum design linear power and maximum fuel temperature. In addition, the 8 x 8 fuel incorporates finger springs in the bundles for controlling moderator/coolant bypass flow at the interface of the channel and fuel bundle lower tie plate. This device has been used satisfactorily in General Electric's initial core and reload fuel for all BWR-4/5 plants, and for one BWR-3 p1ant, as evidenced by inspection of more than 900 fuel assemblies in operating plants employing finger springs.
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TABLE I COMPARISON OF PARAMETERS FOR 8 X 8 AND 7 X 7 ROD FUEL ASSEMBLY DESIGN Pellet Outside Diameter (inches)
Rod Outside Diameter (inches)
Rod-to-Rod Pitch (inches)
Water-Fuel Ratio (cold)
U Bundle Weight (pounds)
Cladding Thickness (inches)
Active Fuel Length (inches) 7 X 7 0.477 0.563 0.738 2.53 412.8 0.037 144.
8 X 8 0.416 0.493 0.640 2.60 409.7 0.034 146.
Fuel perfomance calculations that account for the effects of fuel densification have been performed with our approved version of the General Electric analytical model, GEGAP III A Model for Prediction of Pellet/Clad Thermal Conductance in BWR Fuel Rods," NEDD 20181 dated December 3, 1973, and a Commission letter to I. S. Mitchell "Modified GE Model for Fuel Densification" dated March 22, 1974.
The effects of fuel densification on the fuel rod will increase the stored energy, increase the linear thermal output, and increase the probability of local power spike from axial gaps.
The primary effects of densification on the fuel rod mechanical design are manifested in calculations on fuel-cladding gap conductance and cladding collapse time.
The approved analytical model incorporates time-dependent fuel densification, time-dependent gap closure, and cladding creepdown for the calculation of gap conductance.
Cladding collapse has not been observed in boiling water reactor fuel rods and is calculated to occur at core residence times in excess of five years.
Four different uranium-235 enrichments are used in the fuel assemblies to reduce the local power peaking factor.
In addition, gadolinia-urania pellets are used in four of the highest enrichment rods.
Gadolinia is a burnable poison and supplements the control rods in flattening the power distribution of the core.
The use and perfonnance of gadolinia-bearing fuel was evaluated by us for the initial core of Quad Cities Units 1 and 2 and found acceptable, With respect to the refueling accident described in Section 14.4.4 of the Final Safety Analysis Report, the applicant has not revised the analysis and results to reflect the change from the 7 X 7* core loading.
The total activity from an 8 X 8 assembly (63 rods) is given as equal to the 7 X 7 assembly (49 rods). Accordingly, under the worst postulated conditions, i.e., equivalent fission gas release from the fuel, the total activity release from a fuel handling accident with an 8 X 8 assembly would be equivalent to the 7 x 7 assembly.
For the 7 X 7 assemblies, the fission gas release 4-2
4.3 has been calculated to be 20.8 percent of the 10 CFR Part 100 limits.
Based on the conservative assumptions for fission gas release and the lowe~ final temperatures in the new fuel design, we have concluded generically that the consequences of the refuel-ing accident with the 8 X 8 assemblies would not exceed that with the 7 X 7 assemblies.
To eliminate significant vibration of instrument and source tubes and the resultant wear on channel box corners, Brunswick Unit 1 will incorporate the plant modification described in Section 4.5 of this report.
The effect of this plant modification on the mechanical design is negligible and is discussed for generic application in the safety evaluation on this matter by the NRC staff. This safety evaluation is provided in Appendix B to this report.
Fuel assemblies employing the 8 X 8 fuel design are currently in operation in Nine Mile Point Unit 1, Pilgrim Unit 1, Monticello, Dresden Units 2 and 3, Quad Cities Units and 2, and Vermont Yankee.
Post-irradiation examination of 8 X 8 assemblies at Monticello after one complete cycle indicated satisfactory performance as reported in "8 X 8 Fuel Surveillance Program at Monticello, End of Cycle 3, First Post-Irradiation Measurements," NEDM-20867 dated April 1975.
On the bases of the Brunswick Unit 1 8 X 8 fuel design and the confirmatory results from irradiated assemblies, we conclude that the 8 X 8 fuel design for Brunswick Unit 1 is acceptable based on:
(1)
The fuel rod mechanical design provides acceptable safety margins for normal operation.
(2)
The effects of fuel densification have been acceptably accounted for in the fuel.
(3)
Fuel rod and channel box integrity will be acceptably maintained during transients and accidents.
Nuclear Design The initial core loading for Brunswick Unit l will consist of 8 X 8 fuel assemblies rather than the previously reviewed 7 X 7 fuel assemblies.
The 8 X 8 fuel assembly is a square array of rods, 63 of which are fuel rods and one which is a water filled rod.
The water-to-fuel volume ratio (cold) for the 8 X 8 fuel assembly is 2.60, compared to 2.53 for the 7 X 7 fuel assembly.
Because of the similar water-to-fuel volume ratio, and similar enrichments of the 8 X 8 and 7 X 7 fuel assemblies, their neutronic prop-erties are not expected to differ markedly.
The change in fuel design results in a reduction of maximum design rod power (kilowatts per foot) of approximately 28 percent, from 18.5 to 13.4 kilowatts per foot, due pri-marily to an increase of approximately 30 percent in linear feet of fuel per fuel assembly.
Both the fuel Doppler coefficient of reactivity and the moderator void coefficient of reactivity have very nearly the same values for the 8 X 8 fuel assemblies as have been previously reported for the 7 X 7 fuel assemblies.
4-3
4.4 4.4.l 4.4.2 The calculational techniques used in the design of the Brunswick Unit 1 8 x 8 fuel assembly and reactor core are the same as those that have been used for the 7 x 7 fuel assembly and reactor core.
The eleminate significant vibration of instrument and source tubes and the resultant wear on channel box corners, Brunswick Unit l will incorporate the plant modification of plugging the one-inch bypass flow holes in the lower core support plant modification two holes in the lower tie plate of each fuel assembly to provide an equivalent alternate flow path.
Our evluation of this plant modification's effect on the nuclear design indicates there is no significant change in the nuclear characteristic of the reactor.
We have reviewed the information provided by the applicant in Amendment 31 to the Final Safety Analysis Report and based on the neutronic similarity of the 8 x 8 fuel assembly design to the 7 x 7 assembly design, and on the use of the sam calculational techniques for both assembly designs and reactor cores, we conclude that the nuclear design of the Brunswick Unit l reac_tor core with 8 x 8 fuel assemblies is acceptable.
We also conclude that the plant modification to eliminate significant instrument and source tube vibration has a negligible effect on the nuclear design and is acceptable in this report.
General Electric Thermal Analysis Basis and Emergency Core Cooling System Analysis Introduction The applicant has submitted the analyses supporting the proposed General Electric Thermal Analysis Basis technical specifications and the loss-of-coolant accident analysis in conformance to Appendix K of 10 CFR Part 50.
The analyses are based on the initial core loading of the Brunswick Unit l reactor with 8 x 8 fuel and a modified core to eleminate in-core vibration.
The applicant submitted information consisting of Amendment 31, dated November 26, 1975,.l/ and of supporting letters dated January 2, 1976,Y March 19, 1976,l! and May 7, 1976.lQ/
We have reviewed the submitted information and report our safety evaluation herein.
The applicant has indicated that the previous analysesW are applicable to the modified core.
The staff has evaluated the safety analysis for the modified core on a generic basis and finds the previous analyses acceptable for use with the modified core.Y The references noted in these sections of the report are provided in Appendix C.
General Electric Thermal Analysis Basis Evaluation To apply General Electric Thermal Analysis Basis to the technical specifications involves l) establishing the fuel damage safety limit, 2) establishing limiting con-ditions of operation such that the safety limit is not exceeded for normal operation and anticipated transients, and 3) establishing limiting conditions for operation such that the initial conditions assumed in accident analyses are satisified.
We have evaluated and report herein the Brunswick Steam Electric Plant Unit l developed thermal 4-4
margins based on the NED0-10958 report 3 and plant specific input information provided by the applicant.
As described below, we conclude that the calculated consequences of the anticipated abnormal transients do not violate the thermal and plastic strain limits of the fuel or the pressure limits of the reactor coolant boundary.
4.4.2.1 Fuel Cladding Integrity Safety Limit Minimum Critical Power Ratio The fuel cladding integirty safety limit minimum critical power ratio is based on the General Electric Thermal Analysis Basis statistical analysis which assures that more than 99.9 percent of the fuel rods in the core are expected to avoid boiling transition.
The uncertainties in the core and system operating parameters and the General Electric Critical Quality Xe - Boiling Length correlation of the licensee submitta1,ZI combined with the relative bundle power distribution in the core form the basis for the General Electric Thermal Analysis Basis statistical determination of the safety limit minimum critical power ratio. These uncertainties are the same as or conservative with respect to those reported NED0-1905slf and NED0-20340.Y The reactor core selected for the General Electric Thermal Analysis Basis statistical analyses is a typical core consisting of 764 fuel assemblies. This typical core is of the same reactor class as the Brunswick Unit 1 core, consisting of 560 fuel assemblies but is larger. The bundle power distribution used for the General Electric Thermal Analysis Basis application has more high power bundles than the distribution expected during opera-tion of the Brunswick Unit l reactor. This results in a conservative value of the minimum critical power ratio which meets the 99.9 percent criterion.
We conclude that the proposed fuel integrity safety limit, minimum critical power ratio of 1.05, is acceptable for the Brunswick Unit l initial fuel cycle.
4.4.2.2 Operating Limit Minimum Critical Power Ratio Various transient events will reduce the required operating limit minimum critical power ratio.
To assure that the fuel cladding integrity safety limit minimum critical power ratio of 1.05 is not violated during anticipated abnormal operational transients, the most limiting transients have been analyzed to determine which one results in the largest reduction in critical power ratio (delta minimum critical power ratio). The applicant has submitted the results of those core wide transients analyses which show a significant decrease in minimum critical power ratio.
The types of transients evaluated were losses of flow, pressure and power increases, and coolant temperature decreases.
The most limiting transients in the stated categories were two-pump trip, a turbine trip without turbine bypass, and a loss of feedwater heating.
Of these, the most limiting transient was a turbine trip without turbine bypass resulting in a delta minimum critical power ratio of 0.23.
Addition of the delta minimum critical power ratio to the safety limit minimum critical power rate gives the minimum operating limit minimum critical power ratio required to avoid violation of the safety limit, should this limiting transient occur.
4-5
4.4.2.4 Operating Minimum Critical Power Ratio Limit for Less than Rated Power and Flow 4.4.3 For the limiting transient of recirculation pump speed control failure at lower than rated power and flow condition, the applicant will conform to technical specification limiting conditions for operation. This requires the applicant to maintain the required operating minimum critical power ratio greater than 1.28 times the Kf factor for core flows less than rated.
The Kf factor curves were generically derived and assure that the most limiting transient, a speed control increase, occurring at less than rated flow will not exceed the safety limit minimum critical power ratio of 1.05.
Emergency Core Cooling System Appendix K Analysis On January 4, 1974, the Atomic Energy Commission published its decision in the rule-making proceeding (Docket No. RM-50-1) concerning acceptance criteria for emergency core cooling systems for light water cooled nuclear power reactors.
This decision included an amendment to 10 CFR Part 50 which incorporated the ruling.
The ruling specified that boiling and pressurized light water nuclear power reactors fueled with uranium oxide pellets within cylindrical Zircaloy cladding licensed after December 28, 1974 shall be provided with an emergency core cooling system which shall be designed such that its calculated cooling performance following a postulated loss-of-coolant accident conforms to the criteria set forth in subparagraph (b) of Section 50.46, "Acceptance Criteria for ECCS for Light Water Cooled Nuclear Power Reactors," 10 CFR Part 50.
On November 26, 1975, the applicant submitted an evaluation of emergency core cooling system performance for the design basis pipe break for Brunswick Unit l along with an amendment requesting changes to the technical specifications for Brunswick Unit l to implement the results of the evaluation.Y The applicant incorporated further information relating to the details of the emergency core cooling system evaluation by referencing an appropriate lead plant analysis,£/ to show compliance to Section 50.46 and Appendix K to 10 CFR Part 50.
The background of the staff review of the General Electric emergency core cooling system models and their application to Brunswick is described in the staff Safety Evaluation Report for these facilities dated December 27, 1974 (the December 27, 1974 Safety Evaluation Report issued in connection with the Order).
The bases for acceptance of the principal portions of the evaluation model are set forth in the staff's Status Report of October 1974 which are referenced in the December 27, 1974 Safety Evaluation Report.
The December 27, 1974 Safety Evaluation Report also describes the various changes required in the earlier General Electric evaluation model.
Together the December 27, 1974 Safety Evaluation Report and the Status Report and its supplement 4-7
describe an acceptable emergency core cooling system evaluation model and the basis for the staff's acceptance of the model.
The Brunswick Unit 1 evaluation which is covered by this report properly conforms to the accepted model.
With respect to reflood and refill computations, the Brunswick Unit 1 analysis was based on a modified version of the SAFE computer code, with explicit considerations of the staff recommended limitations. These are described on pages 7 and 8 of the December 27, 1974 Safety Evaluation Report.
The Brunswick Unit 1 evaluation did not attempt to include any further credit for other potential changes which the December 27, 1974, Safety Evaluation Report indicated were under consideration by General Electric at the time.
During the course of our review, we concluded that additional individual break sizes should be analyzed to substantiate the break spectrum curves submitted in connection with the evaluation provided in August 1974.
We also requested that other break locations be studied to substantiate that the limiting break location was the recirculation line.
The additional analyses supported the earlier submittal which concluded that the worst break was the complete severance of the recirculation line. These additional calcula-tions provided further details with regard to the limiting location and size of break as well as worst single failure for the Brunswick Unit 1 design.
The limiting break is the complete severence of the recirculation discharge line assuming a failure of the low pressure coolant injection valve.
During our review, two motor-operated valves not required for safety, Gl6-F015 and G16-F018 located in the drywell, were identified as presenting the possibility of being flooded following a postulated loss-of-coolant accident.
The valves are equipment drain isolation valves and have no post-loss-of-coolant or containment isolation function.
They are powered, however, from Motor Control Center IXL which is safety related.
Responding to our concern regarding the potential development of electrical shorts as a result of flooding, the applicant proposed to implement a scheme that will remove electric power from the motors operating these two valves on closure of containment isolation valves Gl6-F019 and Gl6-020.
We find the implementation of this solution acceptable in that it provides assurance, in addition to that offered by existing Class IE breakers, that any shorts in the motors of the subject non-safety valves will not unduly jeopardize the function of any safety related systems or components.
The results for this Appendix K calculation show a peak cladding temperature of 1925 degrees Fahrenheit, a peak oxidation of less than one percent, and a maximum core average hydrogen generation of 0. 15, for the worst small break assuming failure of the low pressure coolant injection valve.
For the worst small break area of 0.05 square feet assuming failure of the high pressure coolant injection system, a peak cladding 4-8
temperature of 1320 degrees Fahrenheit and a peak oxidation of less than one percent was calculated.
We have reviewed the evaluation of the emergency core cooling system performance submitted by Carolina Power and Light Company for Brunswick Unit l and conclude that the evaluation has been performed wholly in conformance with the require-ments of 10 CFR Part 50, Section 50.46(a). Therefore, operation of the reactor would meet the requirements of 10 CFR Part 50, Section 50.46 provided that operation is limited to the maximum planar linear heat generation rates of Figure 6.8-17 in Amendment 31 to the Brunswick Unit l submittal dated November 1975, and to a minimum critical power ratio greater than 1. 18.
However, certain changes must be made to the proposed technical specifications to conform with the evaluation of the emergency core cooling system performance.
The emergency core cooling system performance analysis assumed that reactor operation will be limited to a minimum critical power ratio of 1.18.
However, a more limiting technical specification limits operation of the reactor to a minimum critical power ratio of 1.28 for 8 X 8 fuel based on consideration of a turbine trip transient with failure of turbine bypass valves.
A statement will be added to the technical specifications bases for the limiting condition of operation indicating the minimum critical power ratio used in the emergency core cooling system performance evaluation.
The technical specifications will require as a reportable event, operation in excess of the limiting maximum average planar/linear heat generation rate values, even if corrective action was taken upon discovery.
We believe that such events should be reported in conformity with the technical specifications.
An evaluation was not provided for emergency core cooling system performance during reactor operation with one recirculation loop out of service. Therefore, continuous reactor operation under such conditions will not be authorized.
The reactor may, however, operate for periods up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with one recirculation loop out of service.
This short time period permits corrective action to be taken and reduces the number of shutdowns which is consistent with other technical specifications of other licensed facilities. During this period, the reactor will be operated within the restrictions of the thermal analysis and will be protected from fuel damage resulting from anticipat-ed transients.
The applicant did not include the equalizer line area in the postulated loss-of-coolant accident analysis; therefore, the technical specifications will require that the equal-izer line remain closed at all times during reactor operation.
The postulated loss-of-coolant accident analysis assumed that all automatic depressuri-zation system valves operated for small break area with high pressure coolant injection failure. Since the applicant did not provide a postulated loss-of-coolant accident analysis with one alternate depressurization system valve out of service for small line breaks, the technical specifications will be modified so as not to allow continuous operation with any automatic depressurization valve out of service; as with other emergency core cooling system equipment, one valve may be out of service for seven days.
4-9
4.4.4 4.4.5 4.5 4.5. l Overpressure Analysis The applicant submitted an overpressure analysis in order to demonstrate that an adequate margin exists below the ASME code allowable vessel pressure of 110 percent of vessel design pressure.
The transient analyzed was the closure of all main steam isolation valves using the indirect high neutron flux scram.
The analysis was performed at 100 percent power with the end of cycle scram reactivity insertion rate curve, scram initiated by high neutron flux, void reactivity applicable to this initial core, no credit for relief function of safety/relief valves, and one valve fails to operate.
The peak pressure at the vessel bottom was calculated to be 1925 pounds per square inch gage yielding an 80 pound per square inch margin below the Code allowable pressure, which is acceptable to the staff.
Conclusions We conclude that the submitted safety analyses of abnormal operational transients for Brunswick Unit l are acceptable.
The minimum critical power ratio established for Brunswick Unit l required to avoid violation of the safety limit minimum critical power ratio should the most limiting transient occur, is acceptable.
The applicant submitted the emergency core cooling system loss-of-coolant analysis which we conclude is in conformance to the requirements of Appendix K to 10 CFR Part 50 and therefore conclude that. the emergency core cooling system is acceptable.
Modifications to Eliminate In-Core Instrument Vibration Introduction In late 1974, a change in the characteristics of the readings from certain of the in-core instruments was observed in a foreign boiling water reactor. Subsequent examina-tion of the fuel bundle channel boxes in this foreign reactor revealed significant wear on the corners of channel boxes adjacent to instrument and source tubes.
This wear had led to cracking and holes in the channel boxes adjacent to the instrument that had displayed the anomalous readings.
During the next year and a half an investigation was made by the General Electric Company and the Commission staff to determine the cause and extent of the above instrument tube vibration and to *establish an acceptable modifi-cation to eliminate the instrument tube vibration.
In February 1976, the Commission, issued the generic "Safety Evaluation Report on the Reactor Modification to Eliminate Significant In-Core Vibration in Operating Reactors With 1-Inch Bypass Holes in the Core Support Plate," which is provided as Appendix B to this supplement.
This report describes the complete history of the in-core instrument tube vibration, the acceptable modification to eliminate the vibration, and analyzed any effects which the modification might have on the core performance during normal transient and accident conditions.
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4.5.2 Brunswick Unit 1 Modifications On April 13, 1976, the applicant submitted a General Electric proprietary report, "Brunswick Steam Electric Plant Unit 1 Safety Analysis Report for Plant Modifications to Eliminate Significant In-Core Vibration," NEDC-21215 on April 30, 1975.
We have reviewed this report and find that our generic safety evaluation of this report dated February 1976 is applicable to the Brunswick Steam Electric Plant Unit 1.
The modification which was made to eliminate the in-core vibration for the Brunswick Steam Electric Plant Unit 1 was to weld plugs in the one-inch holes in the core support plate and drill two holes in the lower tie plate of each of the fuel assemblies.
We have reviewed this modification and conclude that it is acceptable to eliminate signifi-cant in-core vibration.
We required that the Carolina Power and Light Company implement a surveillance program for the Brunswick Unit 1 to provide verification as to the adequacy of the modification to eliminate significant in-core vibration. This surveillance program will require the applicant to:
(1)
Perform a detailed visual inspection of statistically significant number of channel boxes for the first two refueling cycles after the initial fuel loading.
(2) Monitor unfiltered traveling in-core probe traces as required by the technical specifications and report any anomalous behavior to the Con-mission.
(3)
Install accelerometers on a number of in-core instrument/source tubes and monitor the accelerometer signal at least monthly and report any anomalous behavior to the Commission.
In a letter dated July 22, 1976, the applicant corrrnitted to implement the above surveil-lance program.
We conclude that the reactor modifications made on the Brunswick Steam Electric Plant Unit l will eliminate the in-core vibrations and are acceptable.
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- 6. 0 ENGINEERED SAFETY FEA-i"URES 6.2 Containment Systems 6.2.l Containment Functional Design 6.2.l.3 Mark I Torus Loads During a postulated loss-of-coolant accident and events such as safety-relief valve actuation, the torus forming the suppression pool and other structures of the contain-ment system experience dynamic phenomena and loads, which were not adequately considered in the original Mark I containment design.
In order to evaluate the effects of these loads on the containment system, utility companies which own boiling water reactors using the Mark I containment design formed an owners group to study and evaluate the effects as a joint effort. This study is to be accomplished in two stages, a short term program and a long term program.
The objective of the short term program was to assess the integrity of the Mark I containment design based on the latest information available on pool dynamic loads, and to establish that there is no loss of the contain-ment function.
The long term program has the same objective but it consists of a combination of tests, analyses and development of design criteria, and of potential modifications of the containment structures to meet the design criteria. These studies have been mainly performed by the General Electric Company and Bechtel Corporation.
The Brunswick Steam Electric Plant Unit l has unique features for the Mark I contain-ment structures design, that is, the torus in Unit l is constructed as a reinforced concrete shell with a 3/8 inch thick steel liner.
As a result of the short term pro-gram review, only the anchor bars, which are embedded in the concrete shell and con-nected to the ring header support columns, were identified as a problem area in main-taining containment function during a postulated loss-of-coolant accident.
In response to our concerns, the General Electric Company and Bechtel Corporation submitted to the staff on May 3, 1976 a reanalysis of the Brunswick Unit l anchor bar loads taking into account the effects of the 3/8 inch thick steel liner and the Nelson studs which are embedded in the concrete shell and welded to the steel liner. These features were not considered in the earlier calculations.
We have reviewed this reanalysis and concur with the conclusion that there is sufficient margin of safety to preclude failure of the anchor bars. Therefore, we conclude that from the results of the short term program, no modification to the Brunswick Unit l torus structure is required to protect against loss of containment function as a result of a postulated loss-of-coolant accident.
At the conclusion of the long term program, we will require, if necessary, those modifications to the torus structure to assure mair,taining adequate margins of safety based on the loads as determined by the long term program.
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7.2 7.2.5 7.0 INSTRUMENTATION AND CONTROL Reactor Trip System Anticipated Transients Without Scram On December 9, 1975, the staff issued a "Status Report on Anticipated Transients Without Scram for General Electric Reactors." This report provides the staff positions regarding this matter.. On July 7, 1976, a letter was transmitted to the Carolina Power and Light Company indicating that our review of the anticipated transients without scram had been completed and that additional analyses and justification of the analysis model are needed and, that changes in typical boiling water plant designs are indicated.
We have requested that the Carolina Power and Light Company provide by March 30, 1977, the information requested in our letter.
We expect that when the additional analyses have been provided, plant modifications necessary for the Brunswick Unit l have been identified, and the schedule to implement these modifications are provided, the anticipated transients without scram concern will have been resolved.
With regard to the effect of this matter on the review and licensing process, at this time we see no reason to change the conclusion as stated in WASH-1270, that limitations on operation on this account are not necessary or appropriate.
This conclusion is based on our determination that the likelihood of an anticipated transient without scram event is very low considering the number of plants now in operational status, or expected to come into operation before our requirements can be fully implemented.
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- 9. 1
- 9. 1.5 9.0 AUXILIARY SYSTEMS Fuel Storage and Handling Reactor Building Crane Design During our review of the Brunswick Steam Electric Plant Final Safety Analysis Report, it was determined that the reactor building 125-ton crane for this facility were not designed to meet the single failure criterion as detailed in Branch Technical Position Auxiliary and Power Conversion Branch 9-1.
Based on this, a fuel handling accident involving dropping a fully loaded spent fuel cask was postulated at all points along the cask travel path, and the consequences analyzed.
Results of this analysis showed that the cask drop could not be tolerated and still maintain capability to safely shut down the reactor. During the operating license review of Unit 2, the question of the fuel handling accident was not resolved, and a restriction on fuel handling was imposed pending resolution.
Subsequently the applicant made modifications to the reactor building crane in an effort to conform to the guidelines of the above cited branch technical position.
A description of the modified crane and a point by point comparison to the branch position was submitted by letter dated June 18, 1976.
We have reviewed the appli-cant's data and concl~de there is adequate redundancy in the crane control system to preclude the crane traveling outside of the prescribed path for maximum safety during fuel cask handling. Also, the crane has redundancy in the areas of brakes, gear trains, load attaching points, and cask lifting devices, as well as crane control components and systems which are designed fail safe.
Based on our review, we conclude that the integrated design of crane and controls meets the intent of the branch technical position single failure criteria, except in the specific areas of crane reevtng system, protection against two blocking, and control braking.
The crane reeving system does not meet the recommended criteria for wire rope safety factors and fleet angles. The purpose of these criteria is to ensure a design which minimizes wire rope stress and thereby provide maximum assurance of crane safety under all operating and maintenance conditions.
To compensate in these design areas, the applicant, by letter dated July, 2&, 1976, committed to upgrade the wire rope safety factors by either replacing the present wire rope with one having the highest breaking strength available in that size rope, or by limiting the maximum fuel transfer cask load to be handled by the crane to approximately 60 percent of its 125-ton design rated capacity.
In addition, the applicant committed to a specific program of wire rope inspection and replacement, the purpose of which is to ensure that the reeving system will be maintained as close as practicable to original design safety factors at 9-1
- 9. l.6 all times.
Upgrading wire rope safety factors, in conjunction with the inspection/
replacement program will satisfy the staff's concerns, and we. conclude the reeving system as proposed is acceptable.
The applicant has also committed to install an "eddy current" type of control brake in the main hoist system prior to handling spent fuel casks.
The installation of this device satisfies the staff's concerns, and we conclude the crane control braking is acceptable.
The crane control system does not provide adequate protection against two blocking in the event of a fused contactor in the main hoist control circuitry.
However, the applicant has agreed to provide and install a mechanically operated power limit switch in the main hoist motor power circuit on the load side of any hoist motor power circuit controls. This power limit switch will interrupt power to the main hoist motor and cause the holding brakes to set prior to two blocking in the event of a fused contactor.
We conclude the proposed addition will provide adequate protection against two blocking, and the control system is acceptable.
In addition to the above modifications, the applicant has also committed to a program of recertification of the crane to show that the crane has been restored to original condition following use in the plant construction phase, and is capable of handling the design rated loads in normal plant operation.
Based on our evaluation of the data provided, and the commitments made by the appli-cant in the areas of crane reeving, control braking, and two blocking protection, we conclude that the overhead handling system for Brunswick Unit l as proposed is acceptable.
Residual Heat Removal Pumps The low pressure coolant injection system for the Brunswick Steam Electric Plant Unit l was modified to eliminate the loop selection logic system.
A description and our evaluation of this modification is provided in Supplement No. 2 to the Safety Evaluation Report of the Brunswick Steam Electric Plant Units l and 2 issued on December 23, 1974.
Recently our review of this modified low pressure coolant injection system has raised a concern that following a hypothetical loss-of-coolant accident from a recirculation pipe break, this accident could result in a low resistant flow path and a run out condition could occur for the residual heat removal pumps.
If this run out condition were to occur these pumps could be damaged and unavailable for their long term cooling function following the hypothetical accident.
On July l, 1976, Carolina Power and Light Company submitted a letter report providing test data and information which indicate that a pump run out condition will not occur for two residual heat removal pumps discharging through a broken loop.
We are reviewing this report generically and may require additional information from the applicant.
If the results of our evaluation of the residual heat removal pump run out protection 9-2
9.5 9.5.1 indicates the long term cooling capacity and function of these pumps is not adequate, we will require a modification, such as a smaller orifice, for Brunswick Unit l and other similar plants to provide the necessary protection against a pump run out condition.
Other Auxiliary Systems Fire Protection Systems We concluded in the Safety Evaluation Report that the fire protection systems for the Brunswick Steam Electric Plant Unit 1 conform with the requirements of Criterion 3 of the General Design Criteria and, therefore, are acceptable.
On May 3, 1976 the Commission sent a letter to the applicant indicating that to the extent reasonable and practicable, the guidelines provided in the revised Standard Review Plan dated May 1, 1976, Section 9.5.1, "Fire Protection" will be used to reevaluate the fire protection provisions for the Brunswick Unit 1.
This generic investigation presently being conducted by the Commission may impose additional requirements to further improve the fire protection system.
After completing our generic investigation, we will require that the appropriate modifications, if any, to the fire protection systems be implemented.
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11.0 RADIOACTIVE WASTE MANAGEMENT On May 5, 1975, the Corrrnission published in the Federal Register (40 FR 19439) amendments to Sections 50.34a and 50.36a to 10 CFR Part 50 and adopted Appendix I to 10 CFR Part 50.
Appendix I sets forth numerical guidance on design objectives and limiting conditions for operation of light water cooled nuclear power reactors to meet the "as low as practicable" criterion. This term was amended (40 FR 58847) to read:
"as low as is reasonably achievable."
The amended rule became effective on June 4, 1975 and provides in Section V.B. of Appendix I that for each light-water-cooled nuclear power reactor constructed pursuant to a permit for which application was filed prior to January 2, 1971, shall, within a period of twelve months from June 4, 1975, file with the Commission such information as is necessary to evaluate the means employed for keeping levels of radioactivity in effluents to unrestricted areas as low as is reasonably achievable, and to file with the Commission plans and proposed technical specifications developed for the purpose of keeping releases of radioactive materials to unrestricted areas during normal reactor operations, including expected operational occurrences, as low as is reasonably achievable.
Since the application for a license to construct and operate the Brunswick facility was filed with the Commission on July 31, 1968, it falls within the requirements of the amended rule.
The radioactive waste management systems for Brunswick Steam Electric Plant Unit l were evaluated according to the requirements of Section 50.34a and 50.34b of 10 CFR Part 50 published in the Federal Register (35 FR 18385) on December 3, 1970.
By letter dated June 4, 1976, the applicant has provided additional information pursuant to Appendix I to 10 CFR Part 50.
Until such time as we have completed our detailed evaluation of the information for Brunswick Unit l to determine conformance with the requirements of Appendix I to 10 CFR Part 50, we are imposing requirements in Appendi x B technical specifications of the operating license that implement the dose design objectives contained in the Concluding Statement of the Regulatory Staff (Rule Making RM 50-2), dated February 20, 1974.
These numerical values are somewhat more restrictive than those contained in Appendix I, and will assure that releases of radioactive material in effluents will be as low as is reasonably achievable in conformance with Section 50.34a of 10 CFR Part 50.
After we complete our evaluation of the information provided by the applicants, we intend to revise the technical specifications to reflect the requirements of Appendix I to 10 CFR Part 50.
We conclude that using t he Appendix B technical specifications, which are more restrictive than the requirements of Appendix I to 10 CFR Part 50, is acceptable until our evaluation to determine conformance with Appendix I is completed.
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16.0 TECHNICAL SPECIFICATIONS During our review of the Brunswick Steam Electric Plant Units 1 and 2 the Carolina Power and Light Company agreed to use the Standard Technical Specifications for General Electric Boiling Water Reactors which have been developed over the past several years by the staff. The Brunswick Units l and 2 will be the first boiling water reactors to have these standard technical specifications as Appendix A-Prime to their operating licenses.
The applicant has indicated that they will need additional time to modify the operating procedures, which are based on the current technical specifications used for the operation of Unit 2, to be compatible with the standard technical specifications. The applicant indicated that it will complete the necessary procedure changes such that both Units l and 2 can have the Appendix A-Prime Standard Technical Specification effective on or before the date of Unit 2's initial approach to criticality following Unit 2's refueling outage.
During the interim period from the issuance of the Unit l operating license until about April 1977, the technical specifications, Appendix A to the operating license for Unit l will be the same as those technical specifications currently included in the Unit 2 operating license DPR-62 with the necessary changes in the safety limits, limiting safety system settings and limiting condition of operation to account for the design difference between Units l and 2 and dual plant operation.
We have reviewed both the current Appendix A technical specifications in effect for Unit 2 through Amendment No. 20 and the standard technical specifications for Unit l and conclude that either will be acceptable and provide operational limits which will assure the operation of Brunswick Unit l under either of these technical specifica-tions without undue risk to the health and safety of the public.
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APPENDIX A CHRONOLOGY OF REGULATORY REVIEW OF CAROLINA POWER AND LIGHT COMPANY, BRUNSWICK STEAM ELECTRIC PLANT UNITS l AND 2 December 9, 197 4 December 13, 1974 December 13, 1974 December 16, 1974 December 19, 1974 December 20, 1974 December 23, 1974 December 26, 1974 SINCE DECEMBER 2, 1974 Carolina Power and Light Company transmits report entitled "Diesel Generator Load Tests Results."
Carolina Power and Light Company submits response to request to provide information regarding the opera-tion of the recirculation line discharge isolation valves.
Carolina Power and Light Company submits revised figures D-4A, D-4B, D-5A and D-5B for review.
The original figures were forwarded in the Brunswick Report Loss-of-Coolant Accident Analysis Conformance with 10 CFR Part 50, Appendix K.
Carolina Power and Light Company submits a report entitled Reactor Containment Building Integrated Leak Rate Test."
Summary of Meeting Regarding Status of Outstanding Items from the March 1974 Site Visit and Review of Low Pressure Coolant Injection System Modifica-tions issued.
Carolina Power and Light Company submits revised pages to the Brunswick Steam Electric Plant, Units l & 2 and requests that these pages be withheld from public disclosure as proprietary information.
Issuance of Supplement No. 2 to the Brunswick Safety Evaluation.
Initial Decision issued by Atomic Safety and Licens-ing Board for issuance of 100 percent Operating License for Brunswick Unit 2.
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December 27, 1974 December 27, 1974 December 27, 1974 December 30, 1974 January 6, 1975 January 14, 1975 January 15, 1975 January 16, 1975 January 17, 1975 January 23, 1975 January 24, 1975 Facility Operating License DPR-62 issued for opera-tion (100 percent) of the Brunswick Steam Electric Plant, Unit 2.
Directorate of Licensing letter transmitting DPR-62, Operating License for 100 percent power, Federal Register Notice, List of Outstanding Items and 2 copies of Indemnity Agreement No. B-71.
Directorate of Licensing letter regarding the emer-gency core cooling system.
Directorate of Licensing letter transmitting Amend-ment No. 2 to the Brunswick Safety Evaluation.
Assignment of Members of Atomic Safety & Licensing Appeal Board issued.
The Appeal Board is made up of:
R. S. Salzman, Chairman, Michael C. Farrar, Member and Dr. Lawrence R. Quarles, Member.
Carolina Power and Light Company transmits a Final Report on the 24" Residual Heat Removal Outboard Isolation Valve (2Ell-F0-15A) Motor Operator Failure at Brunswick Steam Electric Plant.
Nuclear Regulatory Commission letter advising that the revised pages to the Brunswick Industrial Security Plan have been withheld from public dis-closure as proprietary information.
Nuclear Regulatory Commission letter advising that the proposed revised sheets to the Industrial Security Plan have been approved for inclusion* into the Brunswick Industrial Security Plan.
Carolina Power and Light Company transmits an additional copy of the revised pages to the Indus-trial Security Plan.
Division of Reactor Licensing letter transmitting "Review and Evaluation of GETAB (General Electric Thermal Analysis Basis) for BWRs."
Carolina Power and Light Company requests that the time to respond to the Quality Assurance Program letter of December 3, 1974 be extended to March 1, 1975.
A-2
January 29, 1975 January 31, 1975 February 3, 1975 February 6, 1975 February 7, 1975 February 12, 1975 February 12, 1975 February 14, 1975 February 18, 1976 February 19, 1975 February 26, 1975 February 26, 1975 February 27, 1975 Carolina Power and Light Company submits report Failure of Relay Module in Reactor Manual Control System.
Carolina Power and Light Company submits reevalua-tion of emergency core cooling system performance.
Decision by the Atomic Safety and Licensing Appeals Board affirming the Licensing Board's Initial Decision issued.
Carolina Power and Light Company letter regarding the plant stack exhaust flow measuring capability.
Carolina Power and Light Company and Nuclear Regula-tory Commission representatives meet in Bethesda, Maryland to discuss proposed electrical power dis-tribution for Brunswick Unit 2.
Nuclear Regulatory Commission letter transmitting a copy of D. M. Stewart's Petition for Show Cause and a related Federal Register Notice.
Summary of Meeting Regarding Brunswick Unit 2 Electrical Power Distribution issued.
Carolina Power and Light Company letter transmitting Improvements to Startup Test Program.
Nuclear Regulatory Commisssion letter regarding BWR plants with Mark I containments issued.
Carolina Power and Light Company letter clarifying the electrical distribution to emergency buses.
Carolina Power and Light Company letter transmitting Addendum I to Reactor Containment Building Integrat-ed Leak Rate Test.
Carolina Power and Light Company transmits letter and requests that it be withheld from public disclo-sure as proprietary information.
Carolina Power and Light Company letter regarding the Quality Assurance Program.
A-3
February 27, 1975 March 3, 1975 March 11, 1975 March 19, 1975 March 2D, 1975 March 20, 1975 March 21, 1975 March 25, 1975 March 26, 1975 March 27, 1975 April 3, 1975 Carolina Power and Light Company transmits an Annual Statistical Report of Radiation Exposures.
Meeting Summary for February 24, 1975 - Emergency Diesel-Generator Tests issued.
Carolina Power and Light Company submits request for license amendment - revision of technical specifications.
Representatives from Nuclear Regulatory Commission, Carolina Power and Light Company, and Show Cause Petitioners, David Stewart, et al. meet in Bethesda, Maryland to review the data upon which the request for Show Cause Order was based.
Directorate of Reactor Licensing letter advising that Carolina Power and Light Company can proceed with initial criticality.
Carolina Power and Light Company letter advising of plant organization change.
Directorate of Reactor Licensing confirmation of the letter giving oral authorization to technical specifications change in Appendix B, Section 2.5.2g.
Directorate of Reactor Licensing letter advising that the letter submitted by Carolina Power and Light Company on February 26, 1975 will be placed in the public domain within 30 days unless addition-al justification or a withdrawal request is supplied.
Carolina Power and Light Company letter concerning Decontamination Room -Ventilation System Monitors.
Meeting Summary - Information Meeting with Carolina Power and Light Company, Nuclear Regulatory Commis-sion staff, and Petitioners D. Stewart, Et Al. Re-garding Request to Nuclear Regulatory Commission for Issuance of a Show Cause Order Regarding Possi-ble Dilatancy Near Brunswick Site issued.
Carolina Power and Light Company letter requesting License Amendment - Revisions to Technical Specifications.
A-4
April 9, 1975 April 10, 1976 April 17, 197 5 April 18, 1975 April 18, 1975 April 23, 1975 April 24, 1975 April 25, 1975 April 26, 1975 May 2, 1975 May 2, 1975 May 2, 1975 May 5, 1975 Directorate of Reactor Licensing letter confirming the authorization given by ~he Office of Inspection and Enforcement, Regional II to proceed with startup program including those beyond the initial opening of the main steam line isolation valves.
Order to Show Cause (David M. Stewart, et al. issued by Directorate of Reactor Licensing.
Directorate of Reactor Licensing letter regarding BWR Mark III testing.
Carolina Power and Light Company transmits Supplemen-tary Report on Diesel Generator Special Load Tests.
Carolina Power and Light Company transmits an Improved Startup Test Program.
Carolina Power and Light Company submits gravity survey and seismic data.
Directorate of Reactor Licensing letter advising that the Startup Test Program proposed by Carolina Power and Light Company is acceptable.
Carolina Power and Light Company letter transmitting Request for License Amendment - Revisions to Technical Specifications.
Nuclear Regulatory Commission letter concerning channel box wear.
Carolina Power and Light Company and Nuclear Regula-tory Commission representatives meet to discuss the requirements necessary to satisfy the Show Cause Order issued April 14, 1975.
Issuance of Amendment No. 1, Change No. l to DPR-62 (Unit 2).
Carolina Power and Light Company submits gravity survey and seismic data.
Summary of Meeting to Discuss Requirements of the Order to Show Cause, Issued April 14, 1975.
A-5
May 7, 1975 May 8, 1975 May 8, 1976 May 9, 1975 May 9, 1975 May 9, 1975 May 9, 1975 May 23, 1975 May 29, 1975 June 4, 1975 June 6, 1975 June 12, 1975 Carolina Power and Light Company advises that the traveling in-core probe and local power range monitoring requested by Directorate of Reactor Licensing staff has not as yet been conducted.
Carolina Power and Light Company submits Additional Containment Design Information on DPR-62.
Carolina Power and Light Company advises that commer-cial operation of Brunswick, Unit l is now scheduled for March 1977 instead of March 1976.
Order Granting Extension of Time in Which to Respond to Order to Show Cause issued by Nuclear Regulatory Commission.
Monthly Operations Report for April submitted by Carolina Power and Light Company.
Carolina Power and Light Company transmits Appendix K calculations and revised Technical Specifications.
Carolina Power and Light Company transmits Propri-etary 10 CFR 50 Appendix K calculations.
Directorate of Reactor Licensing letter transmitting Amendment No. 2, Change No. 2 to DPR-62.
Carolina Power and Light Company letter regarding the Suppression Chamber Grout Deficiency on Unit l.
Carolina Power and Light Company letter concerning the recent acquisition of geophysical data which is pertinent to the subsurface geology of the Brunswick Plant Site.
Directorate of Reactor Licensing letter advising that a proposed issuance of amendment to DPR-62 was noticed in the Federal Register. This proposed amendment concerns 10 CFR Part 50 Appendix K calcu-lations and General Electric Thennal Analysis Basis -
GETAB.
Carolina Power and Light Company submits Monthly Operations Report for May 1975.
A-6
June 13, 1975 June 13, 1975 June 18, 1975 June 20, 1975 June 25, 1975 June 27, 1975 June 30, 1975 July 1, 1975 July 3, 1975 July 9, 1975 Directorate of Reactor Licensing letter transmitting Proposed Issuance of Amendment to Operating License, DPR-62 (Unit 2). This amendment defines new tempera-ture limits for the suppression pool water to provide additional assurance of maintaining primary contain-ment integrity and function in the event of extended relief valve operation.
Carolina Power and Light Company submits Request for License Amendment - Revision of Technical Specifica-tions. This revision deletes the requirement to maintain both recirculation pumps speed within cer-tain limits.
Carolina Power and Light Company submits letter regarding Feedwater Sparger Design modification.
Nuclear Regulatory Commission and Carolina Power and Light Company representatives meet with Petitioner T. Erwin to discuss the program proposed to determine if dilatancy is or is not occurring at the Brunswick site.
Summary of Meeting Regarding Fault Information and Porposed Seismic Monitoring Program issued.
Carolina Power and Light Company letter advising that they consent to the issuance of the "Order Modifying License and Revoking Order to Show Cause" with the changes noted.
Carolina Power and Light Company submits Startup Test and Fuel Preconditioning Program.
Carolina Power and Light Company transmits the June 27, 1975 Revision of the "Brunswick Steam Electric Plant Program for Seismic Monitoring. "
Directorate of Reactor Licensing issues letter regarding the emergency core cooling system final acceptance criteria analysis.
Order Modifying License and Revoking Order to Show Cause issued by Directorate of Reactor Licensing.
A-7
July ll, 1975 July ll, 1975 July 21, 1975 July 22, 1975 July 28, 1975 July 29, 1975 July 31, 1975 August 5, 1975 August 5, 1975 August 5, 1975 August 6, 1975 Carolina Power and Light Company submits Additional Information for 10 CFR 50 Appendix K Calculations.
Non-Proprietary.
Carolina Power and Light Company transmits Proprie-tary 10 CFR 50 Appendix K Calculations.
Directorate of Reactor Licensing letter advising that the modification described in Carolina Power and Light Company letter of June 18, 1975 is acceptable.
Carolina Power and Light Company submits letter discussing Thermal Hydraulic Analysis.
Carolina Power and Light Company submits Additional Information for 10 CFR 50 Appendix K Calculations.
Carolina Power and Light Company submits Amendment No. 30 to the Final Safety Analysis Report.
This amendment consists of documentation of the low pressure coolant injection modification, plant organization changes, the Startup Test Program, and continuing quality assurance requirements.
Carolina Power and Light Company letter transmitting Additional Containment Design Information.
Directorate of Reactor Licensing letter requesting a review to determine if containment leaking testing is in full compliance with Appendix J.
Carolina Power and Light Company transmits Startup Test and Fuel Preconditioning Program.
Carolina Power and Light Company transmits Request for License Amendment - Revision of Technical Specifications.
Nuclear Regulatory Commission and Carolina Power and Light Company representatives meet to discuss and resolve those exceptions which Carolina Power and Light Company indicated in its letter, dated 2/27/75 and which were unacceptable to the staff.
A-8
August 8, 1975 August 11, 1975 August 11, 1975 August 12, 1975 August 13, 1975 August 13, *1975 August 14, 1975 August 18, 1975 August 19, 1975 Carolina Power and Light Company submits page changes to the Brunswick Steam Electric Plant Industrial Security Plan.
The changes have resulted from in-creases in the security guard force and redefinition of their posts and duties.
Directorate of Reactor Licensing letter issuing Amendment No. 1 to CPPR-68.
This amendment was effective July 9, 1975 the issuance date of the "Order Modifying License and Revoking Order to Show Cause."
Letter to Carolina Power and Light Company trans-mitting Amendment No. 4.
Change No. 4 to License DPR-62.
The technical specification change defines a new temperature limit for the suppression pool water to provide additional assurance of maintaining primary containment integrity and function in the event of extended relief valve operation. This amendment also adds a safety condition to the license pursuant to Order Modifying License and Revoking Order to Show Cause issued on July 9, 1975.
Directorate of Reactor Licensing letter concerning deletion of technical specification 3.6.F.1.
Authorization given for deletion but amendment to technical specification will be issued at a later date.
Summary of Meeting with Carolina Power and Light Company on Quality Assurance Program issued.
Carolina Power and Light Company letter requesting License Amendments to DPR-62 and CPPR-68.
Memo transferring responsibility of Brunswick Unit 2 from Light Water Reactor Branch to Operating Reactor Branch with an effective date of September 2, 1975.
Directorate of Reactor Licensing issues letter concerning a Surveillance Program Using Traveling In-core Probe Subsystem.
Directorate of Reactor Licensing issues Amendment No. 4, Change No. 4 to operating License DPR-62.
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August 25, 1975 August 28, 1975 September 8, 1975 September 11, 1975 September 18, 1975 October 3, 1975 October l O, 1975 October 17, 1975 November 6, 1975 November 25, 1975 Carolina Power and Light Company submits additional geophysical information.
Directorate of Reactor Licensing issues letter transmitting Amendment No. 5, Change No. 5 to DPR-62, Technical Specifications, Negative Declara-tion and Environmental Impact Appraisal, Federal Register Notice of Issuance of Amendment & Safety Evaluation for GETAB & ECCS Appendix K Analysis.
Carolina Power and Light Company letter requesting an exemption to the Appendix J requirement to test main steam isolation valves at accident pressure.
Carolina Power and Light Company letter concerning Quality Assurance Program.
Carolina Power and Light Company provides informa-tion on water chemistry monitoring programs at the Brunswick Site.
Carolina Power and Light Company letter concerning Additional Containment Design Information - Mark I Containment Short-Term Program Final Report.
Directorate of Reactor Licensing letter advising that the exceptions identified in Carolina Power and Light Company letters of 2/27/75, 8/26/75 and 11/11/75 are acceptable alternatives to WASH documents guidance.
Carolina Power and Light Company letter re Addi-tional Containment Design Information - Mark Containment Evaluation - Long Term Program.
Carolina Power and Light Company letter concerning Seismic Monitoring Reporting Requirements.
Carolina Power and Light Company submits report "Brunswick Steam Electric Plant Unit 2 Safety Analysis Report for Plant Modifications to Elimi-nate Significant In-Core Vibrations.
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November 26, 1975 November 26, 1975 December 5, 1975 December 18, 1975 December 30, 1975 January 2, 1976 January 2, 1976 January 16, 1976 February 6, 1976 February 20, 1976 March 5, 1976 Carolina Power and Light Company letter transmit-ting Amendment No. 31 to the FSAR.
This amendment consists of revised pages concerning fuel design, stability analysis, ECCS analysis, and plant transient analysis for 8 x 8 fuel.
Carolina Power and Light Company submits Proprie-tary 8 x 8 Fuel Design and Emergency Core Cooling System Analysis Information.
This transmittal is part of Amendment No. 31 to the FSAR.
Letter from Carolina Power and Light Company transmitting revised proposal for requalification of licensed reactor operators and senior reactor operators.
Letter from Carolina Power and Light Company trans-mitting information relative to conceptual design modifications for Mark I containment.
Letter to Carolina Power and Light Company request-ing identification of local public document room coordinator.
Letter from Carolina Power and Light Company con-cerning additional containment design information.
Letter from Carolina Power and Light Company trans-mitting revised pages for Technical Specifications.
Letter from Carolina Power and Light Company ad-vising that Mr. James Rutherford has been appointed to coordinate periodic assessment of the status of local public document room for Brunswick.
Letter from Carolina Power and Light Company trans-mitting Seismic Monitoring Program Interim Quar-terly Report, October-December 1975.
Letter to Carolina Power and Light Company request-ing Appendix I information.
Letter from Carolina Power and Light Company con-cerning technical specifications.
A-11
March 24, 1976 March 25, 1976 March 30, 1976 April l, 1976 April 13, 1976 April 21, 1976 April 28, 1976 April 29, 1976 May 4, 1976 May 7, 1976 May 7, 1976 Letter to Carolina Power and Light Company concern-ing Regional Meeting (Region II) to discuss Appendix I requirements.
Letter to Carolina Power and Light Company concern-ing standard technical specifications.
Meeting with Carolina Power and Light Company to discuss latest analytical results performed for torus header and torus header support columns loads during loss of coolant.
Carolina Power and Light Company letter concerning rescheduling fuel loading for Unit l for September l, 1976.
Meeting with Carolina Power and Light Company to perform initial detailed review of standard technical specifications.
Summary of Meeting held with Carolina Power and Light Company to Review Standard Technical Specifi-cation for Unit l issued.
Carolina Power and Light Company transmits Seismic Monitoring Program Interim Quarterly Report -
January - March 1976.
Division of Project Management letter advising that information transmitted for withholding from public disclosure in accordance with Appendix K to 10 CFR 50 (8 x 8 fuel), and 7 x 7 fuel should not be proprietary.
Carolina Power and Light Company letter transmit-ting an updated construction schedule for the Brunswick l Unit.
Nuclear Regulatory Commission letter regarding Appendix I and enclosing a draft copy of Boiling Water technical specifications.
Carolina Power and Light Company responds to Divi-sion of Project Management's request for additional infonnation concerning Amendment 31.
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May 13, 1976 May 18, 1976 May 18, 1976 May 24, 1976 May 24, 1976 May 25, 1976 May 28, 1976 May 28, 1976 June l, 1976 June 4, 1976 Division of Project Management letter advising that Preliminary Safety Analysis Report, Final Safety Analysis Report and Amendments and Environmental Reports, Amendments and Supplements thereto must be filed directly to the Federal, State and local officials by the applicants.
Carolina Power and Light Company letter concerning Additional Information Regarding Mark I Contain-ment Evaluation.
Division of Project Management letter withholding NEDC-21215 from public disclosure as proprietary information.
Division of Project Management letter advising that the applicant's comments on the Proof and Re-view copy of Brunswick Unit l Standard Technical Specifications should be submitted by June 18, 1976.
Carolina Power and Light Company letter concerning Additional Information on Mark I Containment Evaluation.
Division of Project Management letter requesting additional information concerning Bypass Flow Hole Plugs, NED0-21215.
Carolina Power and Light Company transmits Applica-tion for Extension of Construction Permit -
Brunswick Unit l requesting extention from June 30, 1976 to December 31, 1976.
Carolina Power and Light Company letter advising that the material referenced in the Division of Project Management letter of April 29, 1976 can be placed in the public document rooms.
Carolina Power and Light Company letter regarding unexpectedly high flow resistance (head loss) in the discharge canal.
Carolina Power and Light Company issued a letter transmitting Information for 10 CFR 50 Appendix I Evaluation.
A-13
June 14, 1976 June 16, 1976 June 18, 1976 June 18, 1976 July l, 1976 July 7, 1976 July 12, 1976 July 14, 1976 July 21, 1976 July 22, 1976 Division of Project Management letter advising that Carolina and Light Company's request for extension of construction completion date for Brunswick, Unit l is under review.
Memo to Records Facilities Branch and public docu-ment rooms placing a table and figures in the public document rooms for disclosure to the public. This data was previously considered proprietary.
Carolina Power and Light Company issued a letter describing plugs used in the Plant Modification to Eliminate In-Core Vibrations.
Carolina Power and Light Company letter describing how the Units l and 2 reactor building 125 ton cranes meet the Branch Technical Position Auxiliary Power Conversion System Branch 9.1, "Overhead Handling System For Nuclear Power Plants."
Carolina Power and Light Company response to resid-ual heat removal pump run out concern.
Division of Project Management letter requesting infonnation regarding ATWS.
Carolina Power & Light Company's letter furnishing infonnation for Appendix I evaluation.
Division of Project Management letter enclosing staff's "Safety Evaluation Report on the Reactor Modifications to Eliminate Significant In-Core Vibration in Operating Reactors with 1-Inch Bypass Holes in the Core Support Plate."
Carolina Power & Light Company's letter regarding additional infonnation concerning Mark I containment evaluation.
Carolina Power & Light Company's letter furnishing infonnation to surveillance program to ensure plant modifications.
A-14
July 26, 1976 July 26, 1976 Carolina Power & Light Company's letter regarding reactor coolant specific activity technical specifications.
Carolina Power & Light Company's letter furnishing information pertaining to reactor building spent fuel handling cranes.
A-15
APPENDIX B Safety Evaluation Report On the Reactor Modification to Eliminate Significant In-Core Vibration In Operating Reactors with 1-Inch Bypass Holes in the Core Support Plate By Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission February 1976 B-1
-i-Table of Contents 1.0 Introduction 1
2.0 Fuel Channel and Reactor Internals Inspection 4
- 2. 1 Inspection and Wear Criteria....................... 4 2.2 In-Core Instrument Noise 8
Evaluation of Reactor Changes 12 Mechanical Effects 12 Nuclear Performance................................ 15 4.0 Thermal Hydraulic Effects.............................. 17
- 4. 1 Normal Operation.................................. 17 4.2 Transient.......................................... 17 Accident Analysis................................. 19 4.4 Stability Analysis................................ 22 5.0 Demonstration Tests....................................
5.1 Mechanical......... ".............................. 24 5.2 Thermal and Hydraulic............................. 27 6.0 Post Reactor Modification Surveillance................. 28
- 6. 1 TIPs.............................................. 28 6.2 Accelerometers..................................... 29 Internals
- * * * * * * * * * * *
- e * * * * * * * * * * * * * * * * * * * * * * * * * *
- 31 Conclusions ******** 0 *********************************** 32 8.0 References.............................................
B-2 1.0 Introduction In late 1974, a foreign BWR observed a ch2nge in the characteris-tics of the readings fro certain of the in-core instruments.
Sub-sequent examination of the fuel bundle channel boxes in the foreign reactor revealed significant wear on the corners of channel boxes adjacent to instrument and source tubes.
This wear had led to crack-ing and holes in the channel boxes adjacent to the instrument that had displayed the anomalous readings.
The General Electric Company notified the NRC immediately of a possibly similar problem in domestic BWR/4 plants.
Subsequently, the NRC ordered all the utilities with a similar reactor to inspect for this characteristic noise* and to notify the NRC if the noise level exceeded the prede-termined acceptable level.
The channel degradation was caused by vibration of instrument and source tubes excited by high velocity coolant flow from the 1-inch diameter bypass holes in the core support plate.
The presence of cracks or holes in a channel box is of concern since it would allow part of the cooling water that normally flows through the fuel bundle to flow out of the cracks or holes and by-pass the fuel rods.
Such a change in flow pattern would decrease the safety margins for the thermal performance of the fuel.
These reduced margins could lead to overheating and damage to the fuel in the event of some anticipated operating transients or some postulated accidents.
Significant wear and cracking of the channel boxes would also affect their mechanical strength for transients and accidents.
- Noise is defined as the ratio of fluctuations in the signal in the frequency range of interest (generally 1-4 Hz), divided by the mean value of the signal.
B-3 If large cracks occur in channel boxes, there could be a potential for direct ir!ipacting of the tubes on fuel rads or interference ~,ith control rod movement.
The NRC ordered those plants with a high Traversing In-Core Probe (TIP) noise level to lower coolant flow and power to minir:tize the damage to the channel.
On July 18, 1975, the staff issued a safety evaluation report( 1)* stating that *no further damage to the channel boxes is expected when the flow rate is reduced.
Also, the staff concluded that when the reactors are operated at the reduced power level described in the GE submittal( 2) the reactors will not present an undue risk to the health and safety of the public, even with degraded channel boxes.
Some utilities, e.g., operators of the Duane Arnold and the Ver.nont Yankee BWR/4's, decided to shut down the reactors and plug the bypass holes in the lower core plate.
The NRC approved such an action( 3) and stated that plugging only could result in an allowable power penalty for some reactors.
Concurrent with this action, GE has developed a permanent reactor modification to eliminate significant in-core vibration.
The permanent modification consists of both drilling two holes in each fuel bundle lower tie plate to provide an alternate bypass flow path and at the same time plugging the 1-inch bypass holes.
The GE development of this permanent modification for the channel box wear problem has been completed and reported to the staff( 5).
The staff has completed
- References are numbered and listed in Section 8.
B-4 its generic revie~ of the pe::-manent modification only for reactors em-ploying fuel bundles with the holes drilled in all l owe~ tie plates in conjunction with plugging of all the 1-inch byp~ss holes (e.g., Browns Ferry 3).
The review is summarized in this safety evaluation report.
Concurrently the staff has reviewed the effects of drilling holes in ths lower tie plates for some but not all of the fuel bundles within the core (e.g., Browns Ferry 1 and 2).
Since the number of bundles with holes drilled in the lower tie plate directly affects the bypass-region-to-bundle flow rates and the reflood rate for ECCS, the safety analysis for those reactors not having holes drilled in all fuel bundles must be reviewed on an individual basis.
Thus, the complete evaluation for operating limits on any reactor having drilled holes in only some of the fuel bundles is excluded from the scope of this summary.
However, the mechanical and hydraulic con-siderations of operating with only some of the fuel bundles having drilled holes were considered.
B-5 2.0 Fuel Channel and Reactor Internal Insoection 2.1 Insoections and Wea~ Criteria As a routine part of planned reactor shutdowns, the chan-nel boxes and instr~~ent and source tubes are visually inspected for corner wear.
Cracks or holes in the channel boxes are readily apparent in the spent fuel pool without optical aids.
The results on each channel are compared with predetermined acceptance criteria for reuse.
The bases for establishing acceptable wear limits as well as the inspection plan are discussed in the GE report NEDC-20994< 4>
The radial depth of the wear on the channel box corners was estimated from an inspection procedure used at several B\\-IR/4 reactor sites.
The inspection station was located at the fuel storage pool using a fuel preparation machine, a borescope and a visual standard.
The channel wear was observable visually by the contrast between the Zircaloy-4 metal and the zirconium oxide adhering to the unworn por-tion of the channel box.
Cracks and penetrations were observable oy their lack of light reflection.
The widths of the wear marks were measured by direct comparison with the known dimensions on the visual standard.
The depth of wear was infe.rred from a simple Pythagorean derivation for the radial overlap of two eccentric circles (Figure 2-2, reference 4).
This inference assumes no horizontal wiping of the tube on the channel.
The depth from uniquely wiping wear 1s only 42i of that inferred by this technique.
Thus, the technique used to estimate corner wear was conservative.
8-6 General Elect r ic has perfor~ed visua l i nspections specifically for channel box wear at 18 reactors (9 with bypass flow holes in the lower core plate and 9 without bypass holes).
The results of all the reported inspections have been reviewed in detail by the staff.
More than 1600 channel boxes were examined during these inspections at those plants with bypass flow holes.
Only some in-core tubes are adjacent to bypass holes.
No unusual wear was observed at any chan-nel box corner not adjacent to in-core instrument or source tube.
The reject rate for channels adjacent to bypass holes is about two times higher than the reject rate for channels adjacent positions with no bypass holes.
Thus, the staff has concluded that the joint presence of both in-core instrument and source tubes and bypass flow holes was necessary to cause significant channel box corner wear.
The results of the more detailed inspections at nine other reactors having no bypass holes in the core plate have also been reviewed.
The inspections were focused upon more than 100 channels adjacent to in-core instrument and source tubes.
The results show that reactors without bypass holes.in the lower core support plate have exhibited no signifi-cant channel box corner wear.
B-7 General Electric recommends two types of channel inspections*:
diagnostic and general.
The procedure i s tc diagnose the extent of wear by saapling selected channels and by performing a general in-spection for all the channels adjacent to an in-core instrument tube only when the diagnosis yields evidence of significant wear.
When the channel wear problem was first identified GE, re-investi-gated their channel box design margins.
They found that when a chan-nel box corner was worn less than.01 to.02 inches (the nominal wall thickness is 0.08 inches) the original design limits were not violated.
This conclusion was based upon a stress analysis of the channel boxes considering all modes of loading conditions such as steady state, fatigue, steam line break and seismic.
GE identified fatigue as the limiting design loading.
The fatigue loadings result from pressure variations from normal operations (e.g., startups and shutdowns, daily and weekly load reductions, and rod worth tests) as well as the various abnormal transients (e.g., pump trip, turbine trip, generator load rejection, etc.). The information supplied< 4> was not sufficiently comprehensive to perform an exhaustive review of the channel integrity.
However, the staff performed several bounding calculations for maximum allowable wear and found that GE wear limits are acceptable.
8-8 There are four typ~s of instrument and source tube3 in a BWR.
They are Local Power Range Monitor (LPRM), Source, Intermediate Range Monitor (IRM), and Source Range Monitor (SRM).
When there is excessive vibration, these stainless steel tubes impact or rub against the Zircaloy channel box corners.
GE has inspected over half of the total number of in-core instrument tubes at two dif-ferent BWR/4 reactors.
Two LPRM tubes were replaced because they exceeded GE's wear limits. It should be noted though that those two tubes were located where channels experienced through-wall wear and some pieces of the channel were torn off.
The GE criterion for allowable wear on the instrument tube is approximately 20% of the nominal thickness and could mean that the tube resistance to collapse was reduced to half its original resis-tance.
The staff's calculation indicated that.01 inches of wear does not constitute a significant reduction from the original safety margin.
We therefore conclude that the allowable wear for the SRH and IRM tube should not exceed 0.01 inches and the criterion be applied in all future plant inspections.
Furthermore, we require that all.the LPRM tubes adjacent to a one-inch hole in the lower co1:P..o:1mport olate be inspected prior to restart when the diagnostic inspection indicates th~t there is significant wear on the channels in a BWR/4.
B-9 2.2 In-Core Instrument Noise When the core flow exceeds about 40 percent of rated flow for BWR/4's with bypass flow holes, the signal from the fission detectors of the LPRM subsystem and the TIP subsystem exhibit a characteris-tic noise associated with vibrating LPRM instrument tubes.
This characteristic noise in the TIP traces and LPRM time traces has a frequency range of about 1 to 4 Hz.
However, other low frequency noise is also observed in these signals and is similar to that ob-served in BWR/3's.
The neutronic signals generated by the fixed LPRM detectors and the moveable (or parked) TIP detectors and as recorded by plant or special recording instrumentation can be correlated with the im-pacting of channel box corners and instrument tubes in a number of ways.
A direct approach consists of estimating the 1 to 4 Hz noise content in a TIP trace or an LPRM time trace.
Another approach consists of using noise analysis techniques and computing either the power spectral density (PSD) as a function of frequency for a detector or the cross power spectral density (CPSD) as a function of frequency for any two detectors.
The acoustic* noise caused by impacting in-strument or source tubes on channel boxes can also be measured with accelerometers attached to instrument/source tube components that are external to the reactor pressure vessel.
Other approaches which use piezoelectric affects (TIP detector as a sensor) may also be used as an indicator of vibration.
- The signals recorded with the accelerometers are termed "acoustic noise" in this report for the sake of brevity and convenience.
B-10 All of the various methods of relating observations on this impacting and vibration of instr-t.:ment/source tubes indicate the same trends.
BWRs with plugged bypass flow holes in the lower core support plate indicate little neutronic or acoustic noise characteristic of the vibrating or impacting of instrument tubes on channel box corners.
BWR/4s with bypass flow holes open but with core flows restricted to 40 percent or less of rated flow also in-dicate similar results.
But BWR/4s with bypass flow holes open and operating in the range of 40 to 100 percent of rated flow exhibit neutronic/acoustic noise varying froc slight to considerable for the affected instrument/source tubes.
The measured channel box corner wear for several BWR/4's has been shown to correlate with neutronic noise, either directly esti-mated or computed PSDs or CPSDs.
However, the correlations are not strong.
All that can be said is that the greater the neutronic noise with a frequency content of 1 to 4 Hz at a given location the greater the expectation of channel box corner wear.
Establishing a reliable correlation is difficult due to the complexity of the phenomena (e.g., number and placement of bypass flow holes around an instru-ment source tube, the motion of the affected tube and fuel channels, the control rod position and previous history, the in-channel void content, the bypass region void content, core wide flux gradients, microphonic noise of the detectors, variations in core flow 1 and the quality of the plant measuring systems).
Quantitative aspects B-11 of the effect 0f position and 'loids on the detector signal have been studied by our consultants at the Brookhaven National Lab-oratory (5).
The calculations performed by our consultants generally support the previously stated observations concerning neutronic noise caused by vibrating instrument tubes.
Although the effect of instrument tube movement and channel box corner wear on neutronic noise is generally understood, it is currently not possible to predict the occurrence of holes, splits, or cracks in channel boxesf We believe that the general complexity of the associated phenomena, the range of reactor opera-ting states and the lack of sophistication of plant instrumentation precludes exact predictions of the occurrence of holes, splits, or cracks in channel boxes.
However, we conclude that trends in measurements over a period of time, with reactor operation at substantial core flow rates permits an assessment of the po-tential for substantial channel box damage.
Therefore, based on our own analysis of the channel box corner wear data and neutronic noise, the study performed by our consul-tants, and a review of the information from domestic BWR/4s con-cerning channel wear and noise, we conclude that:
(1)
BWRs with plugged or no bypass flow holes in the lower core suHport plate do not have any significant B-12 neutronic or acoustic noise of the type associated with the channel wear problem, (2)
BWR/4s with bypass flow holes do no.t have any significant neutronic or acoustic noise, of the type associated with the channel wear problem, if the core flow is restricted to abou~ 40 percent of rated flow or less, (3) the measured neutronic and acoustic noise, for BWR4s with bypass flow holes open, increase as a function of increased core flow, (4) neither neutronic or acoustic methods are presently capable of indicating the occurrence of holes, splits, or cracks in a channel box, and (5) noise measurements need to be evaluated over a period of time to monitor any changes or abnormalities as an indication of potential for channel box wear.
B-13 3.0 Evaluation of Reactor Changes 3.1 Mech2nical Effe c t s General Electric has proposed to reduce the vibration of instrument and source tubes by eliminating adverse crossflow because of the 1-inch bypass holes in the lower core support plate adjacent to these tubes.
The design change proposed to eliminate adverse coolant crossflow at in-core tube elevations is to both drill two holes in each fuel bundle lower tie plate and to plug the bypass holes in the lower core support plate.
The two drilled holes are always located at the narrow-narrow interchannel gap and not at the wide-wide gap where the flow might impinge on the control blades.
With all the bundles drilled there are approximately ten times as many holes as there were in the core support plate, and the total flow area is slightly less.
The holes in the fuel bundle lower tie plate are slanted to direct coolant flow down toward the core support plate prior to mixing into the total bypass flow which is upward.
This results in a more uniform flow throughout the*core at elevations adjacent to the in-core tubes.
The uniformity of flow was demonstrated at the GE cold flow test facility by measuring axial velocity distributions.
Drilling only some of the fuel bundles is expected to provide a partial benefit of reduced adverse crossflow at elevations ad-jacent to in-co~e tubes.
Thus, no adverse effect on channel box wear is expected when operating with only some of the bundles having holes drilled in their lower tie plates.
B-14 The lower tie plate serves to support the weight of the fuel bundle and rests on a fuel support casting (see Figure 5-3, refer-ence 4).
Both components are stainless steel.
The thickness of the tie plate wall is approximately 1/2 inch at the holes.
A stress analysis (including the stress concentration factor for the holes) indicated that the stress levels are an order of magnitude below the allowable stress when all the expected loads are considered for normal, abnormal and postulated accident conditions.
GE also investigated implications of a misoriented bundle where the flow would be directed toward the control blade.
Simu-lated tests in the cold flow facility at San Jose showed no abnormal control rod vibration.
GE further examined the effect of this design change on other internal components (e.g., core support plate, guide tubes, shroud support) and found no significant effect.
Plugging the bypass holes is also a part of the reactor modi-fication.
The staff' s safety evaluation of such plugs was performed and found acceptable when the Duane Arnold plant requested such an action without drilling holes in the ti~ plates(3).
The conclu-sions of that evaluation are supported by the service experience or plugs at the Vermont Yankee and the Pilgrim 1 reactors where plugs were installed to eliminate control curtain vibration.
Post-service examination of an extracted plug exhibited neither degradation no~
wear of the plug after one fuel cycle.
The possibility of plug vibration from the flow through the two drilled tie plate holes was investigated by GE at the same cold test facility with full size plugs and tie plates.
No unacceptable plug vibrational re-sponse was found as measured by accelerometers.
B-15 Long-ten:;i fatigue, creep and relaxation of parts of the plug however, should be monitored by reasonable sampling inspection at each outage of the l~ad plants including some rion-destructive and (4) destructive tests.
GE proposed an extensive plug surveillance program which the staff considers mandatory (see section 6).
While developing and demonstrating the plant modification to eliminate wear caused by in-core tube vibration, GE has also developed a method of machining the lower tie plates.
The i!nple-mentation will be performed in two steps: drilling and deburring of the fuel bundle lower tie plate.
These operations on exposed fuel will be performed in the fuel storage pool under about 25 feet of water.
The implementation procedure employs pneumatic drills and clamping devices.
Care has been taken in the design of the equip-ment to preclude misorientation of the fuel bundle.
The verifi-cation that all debrts can be removed was demonstrated in a full-scale underwater test facility.
We observed the underwater machining pro-cedure.
The rigors of the underwater machining procedure will necessitate close adherence by the personnel doing the machining to the specific Quality Assurance requirements.
General Electric has established several levels of contingency plans for possible difficulties during implementation.
The plans begin with simple procedures and progress to the replacement of the entire fuel bundle.
All contingency plans will be demonstrated before their implementation.
B-16 3.2 Nuclear Performance Plugging the bypass flow holes in the lower core supporG plate and providing an alternate flow path through holes drilled in the lower tie plate of each fuel bundle has* essentially no effeqt on the nuclear characteristics of a BWR/4 as compared to the same BWR/4 with bypass flow holes open.
A consideration of the important nuclear parameters for a modified BWR/4 and a BWR/4 with bypass holes open shows that:
(1) the in-channel void fraction for steady-state operation will be the same, (2) the voiding in the bypass flow region will be the same (negligible),
(3) the void coefficient of reactivity will be the same, (4) the Doppler coefficient of reactivity will be the same (5) the full power scram reactivity function will be the same, and (6) the TIP uncertainty will be the same.
Thus, we conclude that (1) the total control system worth, temperature, and void dependent behavior of a modified BWR/4 will not differ significantly from that previously reported for the same BWR/4 with bypass flow holes open, B-17 and therefore, (2) the p~eviously deter.nined and reported steady-state operating characteristics and abnormal operating transient results, for a BWR/4 with bypass flow holes open, will also apply for the same modified BWR/4.
Our assessment of the nuclear performance is based on all fuel bundles being drilled with two holes for the alternate bypass flow holes. If only some of the fuel bundles are drilled, then we will require that either
- 1) a plant specific evaluation be submitted for a partially modified reactor, or
- 2) the plant nuclear parameters, characteristics, and performance be based on the more conservative plugged-only core configuration (e.g., reference 3).
B-18
4.0 Thermal :Iydraulic Effects 4.1 Nor:!!al Ooeration GE has evaluated the thermal-hydraulic effects of the proposed core flow path modifications dur.ing normal operation.
The evaluated core flow path modifications consist of drilling new holes in the fuel bundle lower tie plates and plugging the bypass flow holes in the core support plate.
For this analysis, a model. of the reactor core was assumed which included a hydraulic description of orifices, the lower tie plates, fuel rods, fuel rod spacers, upper tie plates, fuel channels, and the core bypass.
One criterion used by GE for normal operation is that the core bypass flow shall range between 10 to 12% of the total core flow.
The hole size selected by GE to be drilled in the fuel bundle lower tie plates was based on tests performed in the ATLAS test facility( 4) which simulated the inlet geometry and bypass region for one fuel bundle under typical BWR operating conditions.
Based on the test results, the size determined by GE satisfies the normal operating conditions for bypass flow even though the total flow area of the new holes in the fuel bundle lower tie plates is slightly less than the previous bypass flow area in the core support plate.
4.2 Transient Abnormal operational transients required to be analyzed are the results of single equipment failures or operator errors that are expected to occur at low frequencies during any normal or planned 8-19 mode of operation.
For a typical BWR/4, the worst total core dynamic events with scram are turbine trip without bypass to the condenser and loss of feedwater heater.
The worst local event is the rod with-drawal error.
The most severe core dynamic event result~ng in nuclear system pressure rise (when direct scram signals are not used in the analysis) is closure of all main steam isolation valves with reactor scram occurring in the analysis on the high neutron flux scram signal (the "safety valve sizing" transient).
The above limiting transients have been previously analyzed for typical BWR/ lf reactor/7
~ 8) with core support plate bypass holes not plugged and no holes drilled in the lower tie plate.
The holes to be drilled in the lower tie plate have been sized so that core average voids remain the same, and it has been shown by a parametric study presented in Table 4-3 of reference 4 that there are no significant differences in active coolant flow percent, bypass flow percent, average exit bypass voids, and core differ-ential pressureo For all BWR/4 plants in which the lower tie plate holes are sized so that there are no significant changes in these parameters, we conclude that previous analyses made assuming holes not plugged in the core support plate and no holes drilled in the lower tie plates will be applicable and acceptable for the new core configuration having plugged holes in the core support plate and holes drilled in the fuel bundle lower tie plate.
B-20 4.3 Accident Analysis In evaluating the effects of the alternate fl ow path on emergency core cooling system performance, it has been proposed that each applicant not perform new "Appendix K" calculations.
Instead, a generic, parametric study of the coolant flow rates through the alternate flow path has been made (NEDE-21156)( 6).
The parametric study has demonstrated that during the reflooding stage of a LOCA, the new flow paths (including the two new holes in each lower tie plate and the path around the "finger springs" between the fuel channel box and the lower tie plate), will supply flow to the lower plenum equal to or greater than the flow through the previously available flow paths (which included the old 1-inch holes in the core support plate but not the "finger spring" path).
GE has obtained test data which demonstrates enhanced flow around the "finger springs" which, for ECCS calculations, would result in an overall increase in flow from the bypass region into the lower plenum and thus would result in an earlier reflood and lower peak-clad-temperature.
Howeverf GE has chosen to utilize the new data only to justify the conservatism of continuing to apply present ECCS models to plants with the new flow paths.
We concur that this is conservative and acceptable for plants where it can be shown that the total flow through all paths from the bypass region to the lower plenum is at least equal to such flow before the reactor B modification.
Flows after the modification can take credit for the finger 3pring path only on those bundles havi~g finger springs.
Two additional concerns were initially expressed by the NRC staff.
These have now been satisfactorily resolved as follows:
GE has evaluated the potential for counter current flow limitation (CCFL) through the new holes and concludes that no flow limitation will occur because steam flowing from the lower core plenum into the core region will not flow through the new holes.
They have stated that steam flow through this path is precluded if a pressure dif-ference corresponding to a few inches of water is maintained in the bypass region.
To justify this position, GE has calculated the pressure drop required so that water will flow from the bypass region into the core.
The result of this calculation indicates that an accumulation of about two inches of water in the bypass region suffices to divert all steam through the active core there-by precluding the possibility of CCFL in the alternate flow path.
In addition to the CCFL consideration, GE has evaluated the potential for entrainment of the core reflood water.
For this cal-culation, _the lower plenum steaming rate was calculated assuming the Wilson bubble rise correlation~9)
Thus, the steam velocity passing by the holes in the lower tie plates was calculated to determine the possibility of entrainment of core reflood water B-22 resulting from the drag force on the drops.
To facilitate this drag calculation, a further calculation was performed to deter~ine the water droplet size distribution of the fluid flowing out of the holes in the lower tie plates into the core.
The drag calcu-lations indicated that the amount of water entrained is negligible.
B-23 4.4 Stability Analysis The GE stability analysis depends upon the type of reactor modification.
The reactor modification within the scope of this evaluation is limited to the complete modification (both plugging of the bypass flow holes in the lower core support plate and drilling the alternative flow path holes in all fuel bundle lower tie plates).
The stability analysis presented( 4) consisted of a description of the criteria employed and the results of a channel hydrodynamic stability evaluation and a core reactivity stability evaluation.
The criteria presented are in terms of the ultimate performance limit criteria, which is a stability boundary, and the operational dynamic characteristics, which are margins to the stated boundary.
These criteria are identical to those used in similar analysis performed for the Hatch and Duane Arnold plants.
The results of the stability analysis are generically presented.
as typical for a BWR/4 core using the modified design.
The results of the analyses are presented as curves of decay ratio versus power for core reactivity stability and channel hydrodynamic stability.
In each case, such curves are presented for the existing design, which has bypass flow holes in the core plate, and curves for plugged bypass holes in the core support plate and the alternate flow path through the fuel bundle lower tie plates.
In all cases, over the power range investigated {52 percent to 100 percent), the design with the alternate flow path has equivalent or greater stability margins than the existing design.
B-24 In past analyses, the channel hydrodynamic stability has been evaluated at the most limiting condition that occurs at the end of the core life, with power peaked to the bottom of the core because the control rods are fully withdrawn.
Similarly, the core stability is also evaluated at this most limiting c*ondition.
Because of the decrease in the delayed neutron fraction, the magnitude of the void reactivity increases.
The most sensitive reactor operating condition is that corresponding to both natural circulation flow and to a power level corresponding to the rod block set point.
In conclusi,pn, it can be stated that in terms of decay ratio, the channel and core hydrodynamic stability margins are nearly the same (slightly enhanced) for the modified BWR/4 as compared to the original BWR/4o The stability analysis encompassed a power range from 52 per-cent of rated flow to 100 percent of rated flow.
These conclusions are based upon the assumptions that:
(1)
Decay ratios are an adequate measure of stability for the power range encompassed.
(2)
The codes and models used in the analysis are conservative.
(3)
The operational conditions of the analysis are limiting.
B-25 5.0 De~onstration Tests GE perfor~ed a cold hydraulic test at its Sar. Jose facility to first determine the cause of in-core instrument tube vibration and channel box damage and secondly, to see that their proposed modifica-tions will perform satisfactorily as expected.
Thirty-two fuel bundles (4x8 array) were installed in a test tank with as-manufactured channel boxes, lower tie plates, control rod plates, fuel support castings and in-core instrllilent tubes.
Plan views are given on pages 5-64 through 5-86 in reference 4.
There are some differences between the test and an in-reactor configuration.
The LPRM tubes in the test are cut short to approxi-mately 15' and attached to a spring whereas these tubes are more than 40' long in-reactor.
All the internals in the LPRM tubes (TIP tube, fission chamber and cables) were removed to facilitate installing an accelerometer.
The flow orifices of the fuel support castings were slightly altered to simulate the bypass flow volume.
In some tests, fuel rods were removed from the channels and replaced by dummy weights.
Also, the top of the fuel bundle is sealed (due to limited pump capac-ity) to simulate only bypass region flow and not flow through the fuel.
5.1 Mechanical For the initial BWR/4 simulation, GE was able to produce significant impacting of an LPRM tube and channel box.
When the pro-posed modification for operating reactors was tested, the impacting level was considerably reduced.
The staff monitored these tests and observed them on several different occasionso B-26 Add '._
- l t
~
d
'- th M
L d
- f
- l
- t ( 6) l:.,J.or.a ests were perror::1e at-
.e 1.oss 2.r. 1n3 aci 1. y.
The test facility consisted of sixteen fuel bundles (4x4 array),
one 0.750 inch OD LPRM tube, four control rod blades, a shroud and a pressure vessel. It simulated in-reactor temperatures and pres-sures but no two phase flow was introduced.
Two conclusions were drawn from the tests. First, the amount of bypass flow measured was more than expected.
Secondly, the impact level between fuel bundle and LPRM tube was higher than the value observed in the previous cold tests at San Jose.
GE reduced the lower tie plate hole size from the original to correct for the desired bypass flow.
The reasons for the higher "g" level observed by the acceler-ometer in the LPRM tube were also investigated.
The difference can be attributed to the in-bundle flow.
In the cold test, in-bundle flow was sealed off because of a limited pump capacity thus only simulating bypass flow between channels.
When the flow was allowed to pass through the fuel bundle in a channel box at Moss Landing it caused a slight excitation of the fuel bundle thus adding to the LPRM tube vibration and impact.
GE confirmed bundle vibration at the cold facility by opening the flow seal to four fuel bundles.
Further tests were performed at Moss Landing for both the BWR/3 simulated configuration and the fully plugged B\\*ffi/ 4 mockup.
GE found that the impact levels are the same as B-27 that of the BWR/4 with the complete modification (ranging be-tween 4 to 8 g's).
They also confirned, at the same facility, that the BWR/4 with bypass flow holes in the core support plate produced accelerations about an order of magnitude higher.
GE concluded that since the impacts for the BWR/3 and for the modi-fied BWR/4 were equivalent and since no significant wear was observed in the BWR/3 channel inspections after full service life, the proposed BWR/4 modifications should eliminate the significant wear.
The Moss Landing tests employed those core components for use in both the BWR/3's and the BWR/4's (both modified and Utliilodi-fied).
Although the scale of the entire core was not simulated in the tests, the relative effects for the hydraulic and mechanical responses of the components were measured at Moss Landing.
The measured impactings for tests from both the BWR/3 components and the modified BWR/4 components were significantly improved relative to those from the unmodified BWR/4 components.
Based upon the above observations and the assumption that the outreactor tests are a scaled equivalent of reactor hydraulic and mechanical environments, we conclude that the instrument and source tube impact levels in the modified BWR/4's are expected to be equivalent to the BWR/3-s.
General Electric reported data to show that no significant wear from impacting has bee~ observed in their BWR/3 surveillance program.
B-28 To provide verification of the expectations on actual operating reactors, we believe that a cociprehensive surveillance program is needed which is further discussed in section q.
Final confirmation of the modification can only occur after the alternative flow path configuration has experienced a full fuel cycle of service.
The plants employing this modified configuration need to schedule a post-irradiation surveillance on the channels at each outage for that purpose (see section 6).
5.2 Thermal and Hydraulic Alternate flow paths and finger spring flow tests were performed by General Electric in the ATLAS facility which si:nulated the inlet geometry and bypass region for one fuel bundle under typical BWR operating conditions.
GE has stated that all components used in these tests were typical of those in production and currently operating in BWR's which incorporate finger springs in the fuel design.
The test results provided the applicant with flow loss coefficients for different hole sizes and leakage flow rates around the finger springs.
General Electric used these test results to determine the hole size to be drilled in the fuel bundle lower tie plates.
B-29 6.0 Post Reactor Modification Surveillance In the p~evious sections we have discussed the necessity of having a surveillance prograx:i during reactor operation to guard against the possible recurrence of channel box degradation.
We believe that two different types of sensors can be used to monitor vibrations durL~g power operations:
(1) in-core neutron detectors (TIPs), and (2) accelerometers attached on the tube beneath the reactor which detects the mechanical energy of i.!npact~
6.1
- TIPs, Excessive instrument tube-channel box interaction pre-viously has been determined from the neutronic noise level in unfiltered TIP traces.
The plant modifications, including the plugging of the bypass flow holes, are expected to affect the noise content of the TIP traces.
In particular, the noise in the 1 to 4 Hz frequency range caused by vibration of instrument tubes should be reduced relative to power dependent noise.
Based on our previous surveillance requirements, unfiltered TIP traces were taken prior to any plant modifications at the highest flow and power permitted.
For some plants, TIP traces were also taken at a nt,U11ber of power and flow conditions.
These data provide part of the basis for evaluating the efficacy of the reactor modifications.* After the reactor modification, comparison of similar measu*rements with the pre-modification data will be B-30 made to confirm that the mechanical vibration of the instrument tubes has been substantially reduced.
The unfiltered TIP traces taken during return to power operation will also provide baseli~~
data which can be used to monitor any changes in the 1 to 4 HZ noise level not attributable to such causes as power level, core flow and control rod pattern.
Therefore, we conclude that (1) surveillance using unfiltered TIP traces to monitor the efficacy of the plant modifications, and
( 2) the frequency of taking TIP trac-.es in accordance with GE Standard Technical Specifications (about 4 to 6 weeks of full power operation),*
are an acceptable means for monitoring neutronic noise of the type associated with instrument tube vibrations.
6.2 Accelerometer Since April 1975, when we first learned of in-core tube vibration, considerable experience has been accumulated both at various reactors and the San Jose facility regarding the capability of accelerometers to detect significant impact.
The Cooper, Duane Arnold and Peach Bottom reactors all demonstrated with acceleometers at different flow rates that there is a definitive transition in the flow rate below which no significant
- GE STS Table 4.3.1-1 Item 2e and footnote f (December 1, 1975 revision).
B-31 impact of the in-core tube can be detected.
This was the basis for allowing plants to operate at lower flow ev_en though we suspected that some reduced wear rate may continue.
GE performed an experiment with a full-length LPRH tube mounted upright in the air.
They then impacted the tubes with a hammer and monitored the stress wave with an accelerometer at various locations along the tube.
NRC consultants and personnel from Philadelphia Electric Company, TVA and GE jointly experimented with a piezo-electric accelerome':.er at the Brown 1 s Ferry plant during the current shutdown.
All came to the conclusion that the accelerometer is a viable sensor that detects any significant impact of the in-core tube.
The first two reactors to employ the modified configuration should install accelerometers on the in-core instrument tubes.
We regard this action necessary to provide further evidence of the efficacy of the modified reactor.
The applicants involved should establish a one month surveillance interval and report to us any anomalous behavior observed in the accelerometer.
GE has already accumulated some accelerometer experience in a BWR/3 plant.
This together with the experience obtained during power ascension flow tests at the Duane Arnold reactor( 3) and other reactors with plugs only provides a reference for comparison.
B-32 6.3 Internals GE presented a plan to inspect channel boxes at the earliest refueling outage.
The first two reactors which imple-mented the plant modification will be required to perform detailed visual examinations of a statistically significant number of channel boxes for the first two refueling* cycles after the modi-fication.
The results of current inspections ~ndicate that outer pheripheral bundles may be more susceptable to a corner wear.
The statistical sampling should emphasize channel boxes which appear more susceptable to wear.
GE provided a satisfactory progra:n for the plug surveillance.
It includes removal of two plugs each from the core after two, five and ten years of service.
The plugs will be examined for wear, spring force relaxation and any deformation.
As discussed in section 2. 1, all the i n-core instrument and source tubes should be inspected when the channel box i nspection indicates that there is significant corner wear i n the channels.
Furthermore, an in-core IRM or SRM tube must be replaced when its wear exceeds Oe01 inches.
B-33 7.0 Conclusions We have reviewed the proposed reactor modification (both plcgging the 1-inch bypass holes in the lower core support plate and drilling 2 holes in the lower tie plate of all fuel bundles) and found that:
(1) the outreactor flow test sufficiently demonstrated that the modification will reduce significantly in-core tube impacting and hence channel box damage, and (2) the General Electric evaluation for the effects of the nodification on cech2nical, nuclear and theroal-hydraulic perfonnance in all modes of operations as well as accidents is satisfactory.
We conclude that the above observations are sufficient to permit this modification and subsequent reactor operations at rated power.
However, the first two reactors to employ the modification should implement the surveillance program as described in section 6.
The surveillance program can be summarized as requirements to:
(1) perform a detailed visual inspection of a statistically significant number of channel boxes for the first two refueling cycles after the reactor modification, (2;
monitor unfiltered TIP traces and report any anomalous behavior to the NRC, 8-34 (3) install accelerometers on a number of in-core instru-ment/source tub~s, oonitor the acceleroweter signal at least monthly and report any anomalous behavior to the the NRC, and (4) remove and examine two plugs each after two, five and ten years of service.
For those reactors in which the 1-inch bypass holes are plugged but not all fuel bundles are drilled we conclude that the outreactor flow test sufficiently demonstrated that the modification will reduce significantly in-core tube vibration and hence channel box damage.
However, the allowable reactor power level after such modifications must be reviewed individually for each reactor considering normal operation, anticipated transients, and accidents.
B-35 Het'e.:--e:1ces
- 1.
11Safety Evaluation Report for Parti~l Power Operation in BWRs with Channel Wear 11 from V. Stello to K. Goller, July 18, 1975.
- 2.
Letter from R. Engle, GE, to V. Stello, MRC, July 11, 1975.
- 3.
"Safety Evaluation Report for Duane Arnold Operation with Plugged Bypass Flow Holes", from V. Stello to K. Goller, June 30, 1975.
- 4.
11 Peach Bottom Atomic Power Station Units 2 and 3:
Safety An-alysis Report for Plant Modifications to Eliminate Significant In-core Vibration", NEDG-20994, GE, September 1975. (Proprietary)
- 5.
11A Study of the Effect of Position and Voids on BWR In-Core Detector Readings" by Hsiang-Sheu Cheng, BNL-20547, Sept. 1975.
- 6.
"Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibration", NEDE-21156, January 1976.
(Proprietary)
- 7.
Georgia Power Company, Letter to Mr. A. Giambusso, Director, Office of NRR, from I. S. Mitchell III, July 9, 1975.
- 8.
Hatch Unit 1 FSAR, Docket 50-321.
- 9.
Wilson, Grenda and Patterson, "The Velocity of Rising Steam in a Bubbling Two-Phase Mixture, ANS Transactions, 5(1), p. 151-152 (1962).
B-36
APPENDIX C SECTION 4.0 REACTOR REFERENCES
- 1.
Amendment 31 to the Brunswick Unit 1 Safety Analysis Report, November 1975.
- 2.
Letter from J. Jones, Carolina Power and Light Company to B. Rusche, USNRC, dated January 2, 1976.
- 3.
"General Electric BWR Thermal Analysis Basis (GETAB) Data Correlation and Design Applica-tion," NED0-10958 and NEDE-19058 (Proprietary).
- 4.
General Electric, Process Computer Performance Evaluation Accuracy," NED0-20340, and Amendment 1, NED0-20340-1, dated June 1974 and December 1974.
- 5.
"General Electric BWR Generic Reload Application for a 8 x 8 Fuel," NED0-20360, Revision 1, November 1974 and Supplement 2 to Revision 1, May 30, 1975.
- 6.
"Brunswick Steam Electric Plant, Unit No. 2, 10 CFR 50, Appendix K Calculations and Revised Technical Specifications." Docket No. 50-324.
- 7.
Letter from J. Jones, Carolina Power and Light Company to B. Rusche, USNRC, dated March 19, 1976.
- 8.
Brunswick Steam Electric Plant Unit 1 Safety Analysis Report for Plant Modifications to Eliminate Significant In-Core Vibrations, NED0-21215, March 1976.
- 9.
"Safety Evaluation Report on the Reactor Modification to Eliminate Significant In-Core Vibration in Operating Reactors with 1-inch Bypass Holes in the Core Support Plate," by Office of Nuclear Reactor Regulation, USNRC, February 1976.
- 10.
Letter from J. Jones, Carolina Power and Light Company to B. Rusche, USNRC, dated May 7, 1976.
C-1