ML20216C624

Safety Evaluation Supporting Amend 124 to License NPF-30

ML20216C624
Person / Time
Site:	Callaway
Issue date:	04/02/1998
From:	NRC (Affiliation Not Assigned)
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ML20216C617	List: ML20216C611 ML20216C624
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NUDOCS 9804140518
Download: ML20216C624 (8)

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4 UNITED STATES j

,j NUCLEAR REGULATORY COMMISSION "o%*****/

2 WASHINGTON, D.C. 2066tWJ001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REf,tULATION RELATED TO AMENDMENT NO.124 TO FACILITY OPERATING LICENSE NO. NPF-30 UNION ELECTRIC COMPANY CALi AWAY PLANT. UNIT 1 DOCKET NO. 50-483

1.0 INTRODUCTION

By letter dated October 17,1997, as supplemented by letters dated March 3,1998, and March 17,1998 Union Electric Company (UE) requested changes to the Technical Specifications (Appendix A to Facility Operating License No. NPF-30) for the Callaway Plant.

The proposed changes would revise the Technical Specifications (TS) to modify the plant heatup and cooldown curves and the maximum allowable power operated relief valve setpoint for cold overpressure protection. On March 30,1998, the staff approved UE's request for an exemption from the requirements of 10 CFR 50.60, " Acceptance Criteria for Fracture Prevention for Light Water Nuclear Power Reactors for Normal Operation" in order to apply the American Society of Mechanical Engineers (ASME) Code Case N-514," Low Temperature Overpressure Protection." The Code case was used in developing the cold overpressure mitigation system setpoints.

The March 3,1998, and March 17,1998, supplemental letters provided additional clarifying information that did not change the staff's original no significant hazards ~ consideration dete:mination that was published in the Federal Reaister on January 14,1998 (63 FR 2282).

2.0 EVALUATION 2.1 Materials and Fluence UE's requested amendment is intended to extend the validity of the Callaway Unit 1 P-T limit curves to 20 effective full power years (EFPY). The current P-T limit curves are valid for a service period of 17 EFPY.

1 The fluence evaluation which is the basis for the proposed revised P-T curves was performed when the third surveillance capsule (V) was removed and evaluated at the end of the eighth cycle. The results are documented in WCAP-14895," Analysis of Capsule V from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program," and includes updates for capsules U and Y which were removed at the end of the first and fourth

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• The staff evaluates the P-T limits based ori the following NRC regulations and guidance:

L 10 CFR Part 50, Appendix G, " Fracture Toughness Requirements;" Generic Letter (GL) 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its impact on Plant

, Operations," July 12,1988; GL 92-01, Revislan 1,'" Reactor Vessel Structural integrity, March 6, 1992; GL 92-01, Revision 1, Supplement 1; Regulatory Guide (RG) 1.99, " Radiation I

Embrittlement of Reactor Vessel Materials," Revision 2 May 1988; and NUREG-0800, Standard Review Plan (SRP), Section 5.3.2, " Pressure-Temperature Limits." GL 88-11 alivised licensees that the staff would use RG 1.99, Revision 2 to review P-T Limit curves. RG 1.99, Revision 2

. contains methodologies for determining the increase iri transition temperature and the decrease

. in upper-shelf energy (USE) resulting from neutron radiation.; GL 92-01, Revision 1, requssted that licensees submit their reactor pressure vessel (RPV) data for their plants to the staff for -

review.- GL 92-01, Revision 1, Supplement 1, requested that licensees provide and assess data -

from other licensees that could. affect their RPV integrity evaluations. These data are used by

< the staff as the basis for the staff's review of P-T limit curves, and as the basis for the staff's review of pressurized thermal shock (PTS) assessments (10 CFR 50.61 assessments).

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' Appendix G to 10 CFR Part 50 requires that P-T limit curves for the RPV be at least as b

consentative as those obtained by applying the methodology of Appendix G to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME) Code, Protection Against Non-ductile Failure."

. SRP 5.3.2 provides an acceptable method of calculating the P-T limits for ferritic materials in the beltline of the RPV based on the linear elastic fracture mechanics (LEFM) methodology of Appendix G to Section XI of the ASME Code. The basic parameter of this methodology is the stress intensity factor K,, which is a function of the stress state and flaw configuration. The methods pf Appendix G postulate the existence of a sharp surface flaw in the RPV that is

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normal to the direction of the maximum stress. This flaw is postulated to have a depth that is equal to one-fourth of the RPV beltline thickness and a length equal to 1.5 times the RPV,

beltline thickness. The critical locations in the RPV beltline region for calculating heatup and

^ cooldown P-T limit curves are the 1/4 thickness'(1/4T) and 3/4 thickness (3/4T) locations, which.

correspond to the depth of the maximum postulated flaw, if initiated and grown from the inside and outside surfaces of the RPV, respectively.

The Appendix G, ASME Code methodology requires that licensees determine the adjusted j

l reference temperature (ART or RTuor). The ART is defined as the sum of the initial i

~ (unirradiated) reference temperature (initial RTuor), the mean value of the adjustment in i

reference temperature caused by irradiation (ARTuor), and a margin (M) term.

The ARiuor is a product of a chemistry factor and a fluence factor. The chemistry factor is dependent upon the amount of copper and nickel in the material and may be determined from l tables in RG 1.99, Revision 2 or from surveillance data. The fluence factor is dependent upon the neutron fluence at the maximum postulated flaw depth. The margin term is dependent

- upon whether the initial RTuor is a plant-specific or a generic value and whether the chemistry i

factor was determined using the tables in RG 1.99, Revision 2 or surveillance data. The margin

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, L term is used to account for uncertainties in the values of initial RTuoy, copper and nickel -

L contents, fluence and calculational procedures. RG 1.99, Rev. 2 describes the methodology to be used in calculating the margin term.

2.1.1' Evaluation For the Callaway Unit i reactor vessel, the licensee determined that the most limiting material at the 1/4T and 3/4T locations is the lower shell plate, R2708-3. This plate was fabricated using plate heat C4499-1. The licensee calculated an ART of 100.4*F at the 1/4T location and -

84.2*F at the 3/4T location at 20 GFPY. The neutron fluence used in the ART calculation was 2

2 7.174 X 10 n/cm at the 1/4T location and 2.547 X 10 n/cm at the 3/4T location. The initial RTuor for.the limiting plate was 20*F. The margin term used in calculating the ART for the limiting weld was 34 at the 1/4T location and 32.1 at the 3/4T location. as permitted by Position 1.1 of RG 1.99, Revision 2. '

1 The ART is determined using the chemistry values for each beltline material of Callaway Unit 1.

The Reactor Vessel Integrity Database (RVID) contains chemistry values for each beltline material for all light water reactors in the U.S. The licensee provided updated chemistry data for the beltline materials of Callaway Unit 1 by letters dated March 17,1997 and October 17, 1997 (WCAP-14895). It should be noted that the staff used the updated chemistry values in the review for Callaway Unit 1. In addition, the staff compared the licensee's best estimate

~ hemistry data for weld wire heat 90077 against the best estimate chemistry values in the -

c CEOG Report CE NPSD-1039, Revision 2. The staff verified that the licensee's best estimate Cu and Ni values were the same as the values in the CEOG Report. It should also be noted that the staff is preparing a Request for Additional information (RAl) about certain aspects of the CEOG Report. In accordance with the RAI, the staff will expect the licensee to address any changes, as needed.

The beltline welds in the Callaway Unit 1 RPV were all fabricated using weld wire heat 90077..

The staff reviewed the initial RTuor values, in the RVID, for welds made of weld wire heat

- 90077 for all plants. The staff found that the initial RTuoy value of -60*F for the Callaway_ Unit 1 circumferential and axial welds was acceptable, since there were no other plants with the same l

weld wire heat.

_ The staff performed an independent calculation of the ART values for the limiting material using L

the methodology in RG 1.99, Revision 2.- Based on these calculations, the staff verified that the licensee's limiting matenal for the Callaway Unit i reactor vessel is the lower shell plate, R2708-

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3, that was fabricated using plate heat C4499-1. The staff's calculated ART value for the limiting material agreed with the licensee's calculated ART.value at 20 EFPY. Substituting the ART values for the Callaway Unit i limiting plate into the equations in SRP 5.3.2, the staff verifiaHf that the proposed P-T limits satisfy the requirements in Paragraph IV.A.2 of Appendix G j

t of 10 CFR Part 50.

In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes a minimum L temperature at the closure head flange based on the reference temperature for the flange i

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f o l-material.Section IV.A.2 of Appendix G states that when the pressure exceeds 20% of the f

'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 tests. Based on the flange RT, of 40*F for Callaway Unit 1, provided by the licensee, the staff has determined that the proposed P-T limits satisfy the requirement for the L

. closure flange region during normal operation and hydrostatic pressure test and leak test.

iL, 2.1.2 L Conclusion i

WCAP-14985 reports the values of the fluen::e resulting from the measurement of capsule V.

- In addition, it includes updated values for surveillance capsules U and Y. The updwe refers primarily to the revision of the cross sections which went into effect after these capsules were evaluated. The approximations used in the analysis of capsule V are the'same with those recommended by the staff and are thus acceptable. The calculated and measured values of-the fluence (from all the capsules) are in excellent agreement with the corresponding calculated values'. The proposed 20 EFPY fluence was derived imm these values and due to the good -

agreement are acceptable.

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. The staff concludes that the proposed P-T limits for the reactor coolant system for heatup, cooldown, leak test, and. criticality satisfy the requirements in Appendix G to Section XI of the ASME Code and Appendix G of 10 CFR Part 50 for 20 EFPY. The proposed P-T limits also satisfy GL 88-11 because the method in RG 1.99, Revision 2 was used to calculate the ART.

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' Hence, the proposed P-T limits may be incorporated into the Callaway Unit 1 Technical j

Specifications.

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' 2.2 Cold Overpressure Mitigation System 2.2.1 Evaluation UE also requested to modify the plant heatup and cooldown curves and the maximum allowable power operated relief valve (PORV) setpoint curve for cold overpressure protection, as found in -

TS Figures 3.4-2,3.4-3, and 3.4-4, respectively. These modifications are necessary for the plant to operate up to 20 effective full power years (EFPY), an increase from 17 EFPYs. In lL, addition, TS Bases 3/4.4.9 and 3/4.5.2 through 3/4.5.4, which stated one of the two centrifugal pumps is allowed by the TS to be operational in MODES 5 and 6 operation with the reactor vessel head installed, are revised by including the " normal" charging pump to the centrifugal l

charging pump allowed to be operational. The modifications to these TS Bases are acceptable h

because they provide a consistency with Limithg Condition for Operation (LCO) 3.5.4, which requires all safety injection pumps and one of the two centrifugal charging pumps, but not the o

• normal" charging pump, to be inoperable while in MODE 5 and MODE 6 opesation with the reactor vessel head on.

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l The cold overpressure mitigation system (COMS) uses PORVs located near the top of the l

pressurizer to supplement the water relief valves in the residual heat removal system (RHRS)

L suction lines for protection of the reactor vessel from being exposed to conditions of fast l

propagating brittle fracture. TS LCO 3.4.9.3 requires that, when the RCS temperature is below -

368 F and the RCS is not depressurized with a vent of greater than or equal to 2 square inches, i

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' either two RHR suction relief valves, or two PORVs with setpoints not exceeding the limit.

l established in Figure 3.4-4, or one' RHR relief valve and one PORV shall be operable. The TS l

amendment request will revise the PORV setpoints in Figure 3.4-4 to prevent the RCS pressure -

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. from exceeding the pressure-temperature limits in the revised Figuras 3.4-2 and 3.4-3 for operation up to 20 EFPYs.

i The COMS enable temperature and the PORV setpoints for 20 EFPYs are developed using the methodology described in the approved Westinghouse topical report, WCAP-14040-NP-A, j

Revision 2, " Methodology Used to Develop Cold Overpsssure Mitigation System Setpoir.ts and RCS Heatup and Cooldown Limit Curves," January 1990, as we I as the guidelines of ASME Code Case N-514. Code Case N-514 requires the lowdernperacure overpressure protection l

< (LTOP) systems to be enabled when the reactor coolant te mpera'ure is less than 200'F or at l

' temperatures corresponding to a reactor vessel metal temperatuie less than RTwor (nil-ductility.

L reference temperature) + 50* F, whichever is greater; and to limit the maximum pressure in the reactor vessel to 110% of the pressure determined to satisfy Appendix G. In WCAP-14894, "Callaway Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," July 1997, the I

RTuor for 20 EFPYs is calculated to be 100.4*F. Based on the Code Case N-514 guidelines, the COMS enable temperature is 200* F. Therefore, TS LCO 3.4.9.1, which requires the COMS to be enabled at 368'F, is conservative.

The COMS PORV setpoints are determined based on the design basis analyses consisting of a l

mass input transient and a heat input transient, initiated with the RCS in a water-solid condition and the RHRS isolated from the RCS, disabling the relieving capability of the RHR relief valves.

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- The heat injection scenario is the startup of a reactor coolant pump (RCP) with the steam -

l generator secondary side hotter than the RCS temperature, resulting in a RCS pressurization i

from sudden heat input to a water-solid RCS from the steam generator. The mass injection scenario is caused by the simultaneous isolation of the RHRS, isolation of letdown and failure of the charging flow controls, resulting in an RCS pressurization from a net charging flow input to a water-solid RCS. The analyses of the heat and mass input transients were performed with the i.

LOFTRAN computer code to determine the RCS pressure overshoot after a PORV is actuated.

- In addition, the uncertainties associated with pressure and temperature instrumentation, the

' single failure assumption of a PORV, as well as the hydrostatic and dynamic effects of the RCP J are taken into account in the analyses.

In response to the staff's requests for additional information, the licensee in its letters of March 3 and March 17,1998, provided information related to the analyses performed for development of Callaway COMS setpoints, including input assumptions used in the analyses, i

the PORV relieving capacity, the injection flow rates of the centrifugal charging pump and the

" normal' charging pump. In the heat input transient scenario for the analysis of RCS pressure f

overshoots, an RCP is assumed to start when the steam generator secondary side temperature I

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_ is 50'F higher than the RCS cold leg temperature. This 50*F temperature difference assumption is consistent with LCOs 3.4.1.3 and 3.4.1.4.1, which prohibit a RCP from being started in MODE 4 and MODE 5 operation, respectively, unless the secondary water i temperature of each steam generator is less than 50*F above each of the RCS cold leg,

temperature. The mass input transient analysis assumes simultaneous injection of both a L

centrifugal. charging pump and the " normal" charging pump into the water-solid RCS while the RHRS and the letdown line are isolated. This assumption is consistent with LCO 3.5.4, which requires all safety injection pumps and one of the two centrifugal charging pump to be -

inoperable while in MODE 5 and MODE 6 operation with the reactor vessel head on, and l

3 ther' fore, allows a centrifugal charging pump and the " normal" charging pump te be' operable l

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. under these modes of operation. The combined mass injection rates from both charging.

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pumps at a full capacity over a range of RCS pressures are increased to 105% of the calculated values.

m 2.2.2 Conclusion l

The staff has reviewed the PORV setpoint analyses provided in the UE's March 3 and March 17,1998, letters, including the RCS pressure overshoots for both the mass and heat input transients for the assumed RCS initial conditions and PORV setpoints, the treatment of pressure and temperature uncertainties, and the selection of the PORV setpoints.with comparisons to the Appendix G limits and the limit established for maintaining _the integrity of

_ the PORV piping. The results show that, with the selected PORV setpoints at various RCS temperatures, the peak RCS pressures during both the mass and heat input transients are bounded by the Appendix G limits (multiplied by 110% per Code Case N-514 guideline) or the PORV piping pressure limit. Since the analyses were performed with the approved

- methodology of WCAP-14040, Revision 2, the staff concludes the PORV setpoint curves as specified in the revised Figure 3.4-4 are acceptable.

3.0 STATE CONSULTATION

l In accordance with the Commission's regulations, the Missouri State Official was notified of the proposed issuance of the amendment. The State official had no comments.

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4.0 ENVIRONMENTAL CONSIDERATION

i ll The amendment changes a requirement with respect to the installation or use of a facility i

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component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no-significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (63 FR 2282). Accordingly, the amendment meets the eligibility criteria for categorical i

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exclusion set forth in 10 CFR 51.22,'c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact _ statement or environmental assessment need be prepared in connection with the issuance of the amendment.

5.0 CONCLUSION

The Commission 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 ccnducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: M. Khanna L. Lambros Y. Hsii Date:

' Apr 11 2,1998

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