ML20086M132
| ML20086M132 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 07/20/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20086M128 | List: |
| References | |
| NUDOCS 9507240139 | |
| Download: ML20086M132 (5) | |
Text
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UNITED STATES E
iW I E NUCLEAR REGULATORY COMMISSION
' 'f WASHINGTON, D.C. 2055 5 0001
)
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.199 TO FACILITY OPERATING LICENSE NO. NPF-3 TOLEDO EDISON COMPANY CENTERIOR SERVICE COMPA_NJ ANQ THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DAVIS-BESSE NUCLEAR POWER STATION. UNIT NO. 1 DOCKET NO. 5G-346
1.0 INTRODUCTION
By letter dated January 30, 1995, the licensee submitted changes to pressure-temperature (P-T) limits in the Davis-Besse Unit 1 Technical Specifications (TS). The licensee proposed to revise the P-T limits and to extend the applicable period of the P-T limits from a current 10 effective full power years (EFPY) to 21 EFPY.
The revised P-T limits include an adjustment in the calculations to account for the pressure difference between the pressure j
transmitter and the reactor vessel midplane. Additionally, as required by License Condition 2.C.(3)(d), a reanalysis was performed, as necessary, to ensure continued means of protection against low temperature reactor coolant i
system overpressure events.
The staff evaluates the P-T limits based on the following NRC regulations and guidance: Appendix G to 10 C:R Part 50; Generic letters (GL) 88-11 and 92-01; i
Regulatory Guide (RG) 1.99, Revision 2; and Standard Review Plan (SRP) Section 5.3.2.
Appendix G to 10 CFR Part 50 requires that P-T limits for the reactor vessel must be at least as conservative as those obtained by Appendix G to Section III of the American Society of Mechanical Engineers (ASME) Code.
GL 88-11 provides that licensees may use the methods in RG 1.99, Revision 2, to predict the effect of neutron irradiation by calculating adjusted reference temperature (ART) of reactor vessel materials.
The ART is defined as the sum of initial nil-ductility transition reference temperature (RT of the material, the increase in RT,'he prediction method. cause an account for uncertainties in t The increase in RT is wor calculated from the product of a chemistry factor and a fluence factor.
The chemistry factor is calculated using surveillance data, obtained by the licensee's surveillance program, as directed by Regulatory Guide (RG) 1.99, Revision 2, Position 2.
SRP 5.3.2 provides guidance on calculation of the P-T limits using linear elastic fracture mechanics methodology specified in Appendix G to Section III i
9507240139 950720 PDR ADOCK 05000346 P
of the ASME Code.
The linear elastic fracture mechanics methodology postulates sharp surface defects that are normal to the direction of maximum stress and have a depth of one-fourth of the reactor vessel beltline thickness (1/4T) and a length of 1-1/2 the beltline thickness. The critical locations in the vessel for this methodology is the 1/4T and 3/4T locations, which correspond to the maximum depth of the postulated inside surface and outside surface defects, respectively.
The licensee determined that no changes were necessary to extend low temperature overpressure protection (LTOP) to 21 EFPY.
The Davis-Besse LTOP system consists of both active and passive subsystems.
The active subsystem uses relief valve (DH-4849) in the 4-inch suction line of the decay heat removal (DHR) system to provide overpressure protection of the reactor coolant system (RCS) temperature is less than 280 F.
The passive subsystem is based on the plant design and operating philosophy that prevents it from being in a water-solid condition (except for system hydrotests.)
The Davis-Besse RCS is i
always operated with either a steam or a gas space in the pressurizer; the steam bubble is replaced with nitrogen during plant cooldown when the RCS pressure is reduced.
2.0 EVALVATION For the Davis-Besse reactor vessel, the licensee determined that the middle circumference weld material, WF-182-1, is the limiting material for both the 1/4T and 3/4T locations. Using surveillance data', the licensee calculated an ARTof155'Fatthe1/4Tlocationandll4*Fatthe3/4Tlocationat41EFPY.
18 The neutron fluence used in the ART calculation was 4.365 x 10 n/cm at the 1/4T location and 1.588 x 10 a n/cm at the 3/4T location.
2 The staff used the surveillance data, as submitted in previous reports to the NRC', to perform an independent calculatian of the ART values for the limiting materials using RG 1.99, Revision 2, Position 2.
In addition, the staff verified that copper and nickel contents and initial RT, agreed with the NRC uo reactor vessel material database from the licensee's response to GL 92-01.
Based on the staff's calculation, the staff verified that the licensee's calculated ARTS for Davis-Besse are acceptable.
Substituting the ARTS into equations in SRP 5.3.2, the staff verified that the proposed P-T limits for heatup, cooldown, criticality, and inservice hydrostatic test satisfy the requirements in Paragraphs IV.A.2 and IV.A.3 of Appendix G of 10 CFR Part 50.
In addition to beltline materials, Appendix G of 10 CFR Part 50, also imposes a minimum temperature at the closure head flange based on the reference temperature for the flange material.
Section IV.A.2 of Appendix G states that when the pressure exceeds 20% of the preservice system hydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload must exceed the reference temperature of the material in those regions by at least 120*F for normal operation and by 90*F for hydrostatic pressure tests and leak ests.
Based on the flange RT of 60'F for Unit 1 provided by the licensee}, the staff has determined tha,o1 t the proposed P-T
' See References 1 through 4 2 See Reference 5
l limits have satisfied the requirement for the closure flange region during normal operation, hydrostatic pressure test and leak test.
The provisions for LTOP were reviewed. As discussed previously, the LTOP system utilizes both passive and active subsystems. psdiscussedinthe plant-specific safety evaluation for Amendment No. 57, the water level in the pressurizer must be restricted for a given RCS pressure. Administrative procedures and TSs require the high pressure injection system be disabled when the RCS temperature is below 280*F. The restriction on pressurizer water level for a given RC': pressure ensures that the pressure-temperature limit will not be exceeded if a charging pump pumps the contents of tie makeup water tank into the RCS.
The active LTOP system is the DHR system relief valve, DH4849, which is the primary means of LTOP during modes 4 and 5.
Tre setpoint of the relief valve is 330 psig. This is less than the allowable pre:sure limit of 360 psig at 140 F which assures adequate overpressure protection.
LTOP protection during Mode 3is provided by the passive LTOP subsystem which consists of administrative controls which are not affected by this proposed amendment.
The staff has performed an independent analysis to verify the licensee's proposed P-T limits. The staff concludes that the proposed P-T limits for heatup, cooldown, inservice hydrostatic test and criticality are valid for 21 effective full power years because (1) the limits conform to the requirements of Appendix G of 10 CFR Part 50 and the provisions of GL 88-11 and (2) the surveillance data used in calculating the P-T limits are consistent with data submitted to the staff in surveillance reports. Hence, the proposed P-T limits may be incorporated in the Davis-Besse Unit 1 TS.
In addition, the proposed editorial changes in the Bases section of the TS are consistent with the P-T limits changes; therefore, they are acceptab'a The LTOP evaluation performed by the staff has determined no substantive cnanges 3
have occurred to invalidate previous LTOP evaluations.
j
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Ohio State official was notified of the proposed issuance of the amendment.
The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes a surveillance requirement. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluent that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (60 FR 14029). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR Sl.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
3 See Reference 6
5.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: Glenn Dentel Chu-Yu Liang Linda Gundrum Date:
July 20,1995 l
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5-REFEREN;fi
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I.
BAW-2125, Analysis of Capsule TEl-D, The Toledo Edison Company, Davis-Besse Nuclear Power Station Unit 1 -- Reactor Vessel Surveillance Program-, December 1990.
2.
BAW-1882, Analysis of Capsule TEl-A, The Toledo Edison Company, Davis-l Besse Nuclear Power Station Unit 1 -- Reactor Vessel Surveillance Program-, June 1989.
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3.
BAW-1834, Analysis of Capsule TEl-B, The Toledo Edison Company, Davis-Besse Nuclear Power Station Unit 1 -- Reactor Vessel Surveillance Program-, May 1984.
4.
BAW-1701, Analysis of Capsule TEl-F, The Toledo Edison Company, Davis-Besse Nuclear Power Station Unit 1 -- Reactor Vessel Surveillance Program-, August 1982.
5.
Toledo Edison, Supplemental Information Regarding the License Amendment Request to Revise the Reactor Coolant System Pressure-Temperature j
Operating Limits and Reactor Vessel Material Surveillance Program, May 4,1988.
6.
Abnendment No. 57 to Facility Operating License No. NPF-3, May 5, 1983.
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