ML20206H286

Safety Evaluation Supporting Amends 133 & 131 to Licenses DPR-80 & DPR-82,respectively

ML20206H286
Person / Time
Site:	Diablo Canyon
Issue date:	05/03/1999
From:	NRC (Affiliation Not Assigned)
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ML20206H284	List: ML20206H282 ML20206H286
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NUDOCS 9905110136
Download: ML20206H286 (7)

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NUCLEAR REGULATORY COMMISSION I

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WASHINGTON, D.C. 20666 4001 s *****/

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l

RELATED TO AMENDMENT NO. 133 TO FACILITY OPERATING LICENSE NO. DPR-80 I

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- AND AMENDMENT NO. 131 TO FACILITY OPERATING LICENSE NO. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT. UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323

1.0 INTRODUCTION

By letter dated September 3,1998, as supplemented by letters dated January 22,1999, February 5,1999, and March 17,1999, Pacific Gas and Electric Company (licensee) submitted i

a license amendment request to revise TS 3.4.9.1, Figures 3.4-2 and 3.4-3 regarding the Appendix G pressure - temperature (P/T) limits to extend the applicability up to 16 effective full power years (EFPY). The licensee also updated the controlling materials for the P/T curves to reflect the current analysis for generating the new curves. The licensee also requested an exemption from 10 CFR 50.60 and Section IV.A.2 to Appendix G to 10 CFR Part 50, for use of the American Society of Mechanical Engineers (ASME) Code Case N-514 in determining the acceptable low temperature overpressure protection (LTOP) system setpoints. The licensee proposed revision to the TS Bases to reflect the above changes. The use of ASME Code Case N-514 would compensate for the more restrictive Appendix G P/T limits in the proposed TS affecting LTOP setpoints. The licensee has provided the justification to support its determination that the current LTOP setpoints are unchanged.

The January 22,1999, February 5,1999, and March 17,1999, supplemental letters provided additional clarifying information and did not change the initial no significant hazards consideration determination published M the Federal Reoister on December 16,1998 (63 FR 69345).-

2,0 BACKGROUND The LTOP system mitigates overpressure transients at low temperatures so that the integrity of

• the reactor coolant pressure boundary is not compromised by violating the 10 CFR Part 50, Appendix G. PTT limits under steady state operating conditions. Diablo Canyon Units 1 and 2 LTOP system use the pressurizer power operated relief valves (PORV) or a reactor coolant

_ system (RCS) vent with the reactor depressurized to accomplish this function. The system is manually enabled by operators and uses a single lifting setpoint for the PORV. The design basis of Diablo. Canyon Units 1 and 2 LTOP considers both mass-addition and heat-addition transients. The results of a licensee's evaluation indicated that the mass addition transients are 9905110136 990503 PDR ADOCK 05000275 P

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most limiting for the design of LTOP system. The limiting mass-addition analyses account for the injection from one centrifugal charging pump to the water solid RCS with letdown isolated.

The heat-addition analyses accounts for heat input from the secondary side of the steam

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generators into the RCS upon starting a single reactor coolant pump (RCP) when the steam generator secondary water temperature is less than 50*F above the RCS cold leg temperature.

The plant administrative controls in the current TSs and operating procedures provide restrictions in plant operation within the configuration assumed in the analysis for LTOP system design.

3.0 EVALUATION 3.1 Low Temperature Overpressure Protection The current Limiting Condition for Operation (LCO) in TS 3.4.9.3 requires that an LTOP shall be operable with two operable PORVs with a lifting setting of less or equal to 435 psig, or the RCS depressurized with an RCS vent of 2 2.07 square inches. This LCO is applicable when any i

RCS cold leg temperature is s 270*F when the head is on the reactor vessel. When any RCS cold leg temperature is below 107*F, a vent path greater than 2.07 square inches will be maintained per plant operating procedures. Also, the current TS 3.5.3 provides restrictions for a maximum of one operable centrifugal charging pump and no operable safety injection pump when any RCS cold leg temperature is less than or equal to 270*F. These TS restrictions in combination with other administrative controls in the plant operating procedures regarding RCP operation and ECCS flow path assure that the Diablo Canyon Units 1 and 2 will be operated within the configuration assumed in the analysis for LTOP system design. The method currently used for providing operational restriction in this area is consistent with NUREG-1431 and the licensing bases at Diablo Canyon Units 1 and 2.

Since the use of ASME Code Case N-514 compensates for the more restrictive Appendix G P/T limits in the proposed TS affecting LTOP setpoints, the licensee has proposed to maintain the current LTOP setpoints unchanged. The staff's evaluation of the licensee's justification regarding the LTOP setpoints are discussed below.

1 3.1.1 Enable Temperature i

The LTOP system enable temperature is the temperature below which the LTOP system is required to be operable. The licensee has proposed to maintain the current LTOP enable temperature of 270*F. The licensee has performed an evaluation for the adequacy of the enable temperature of 270*F by using a methodology to: (1) account for instrument uncertainties associated with the instrumentation used to enable the LTOP system and, (2) implement ASME Code Case N-514 that uses an enable RCS water temperature corresponding to a metal temperature of at least RTwor + 50*F at the belt line location (1/4t or 3/4t). Therefore, the licensee calculates the enable temperature as RTuor + 50*F +

temperature difference between RCS and metal + instrument uncertainties. Using the above equation, the calculated minimum enable temperature is 264*F. The licensee proposed an enable temperature of 270*F that includes an additional margin of 6*F.

b li, 3-a The staff finds that this proposed LTOP enable temperature is conservative with respect to the l

enable temperature allowed by ASME Code Case N-514 and the methodology presented in WCAP-14040, Revision 1. Therefore, the staff finds it acceptable.

3.1.2 LTOP Actuation Setpoint-The LTOP system is designed to mitigate overpressure transients at low temperatures to l

prevent violating 10 CFR Part 50, Appendix G P/T limits. Additionally, since the licensee is using ASME Code Case N-514 to determine the acceptable LTOP system setpoints, the NRC staff has accepted the use of the P/T limits which are 10 percent above the steady-state Appendix G limits for the design of LTOP system. The LTOP system actuation setpoint is the pressure at which the PORVs will lift, when the LTOP system is enabled, to limit the peak RCS pressure within the acceptable limits during a pressurization transient.

Diablo Canyon Units 1 and 2 use PORVs to provide pressure relief capacity for LTOP system.

The methodology used for determining the PORV actuation setpoint is consistent with the methodology presented in WCAP-14040, Revision 1.

The licensee has proposed that the current PORV actuation setpoint of 435 psig in TS 3.4.9.3 will remain unchanged to protect the proposed Appendix G P/T limits in Figures 3.4-2 and 3.4-

3. In response to the staff's request, the licensee, in its letter dated January 2,1999, provided a tabulation to list PORV setpoints, transient pressures overshoot, instrumentation uncertainties l

for temperature and pressure and corresponding P/T limit under various temperature conditions below the LTOP system enable temperature. The data presented in this tabulation confirms that the proposed PORV setpoints will provide adequate protection to the 10 CFR Part 50, l

Appendix G P/T limits under steady state conditions during a design basis overpressure transient (mass-addition or heat-addition) as described in Section 2.0 of this report. Based on the above discussion, the staff finds the proposed PORV setpoint acceptable.

3.1.3 RCS Vent Size With the RCS depressurized, the results of the licensee's evaluation showed that a vent size l-of 2.07 square inches is capable of mitigating a most limiting low temperature overpressure transient. The vent size of 2.07 square inches is larger than the vent path area that a PORV, with a minimum throat diameter of 1.625 inches, would provide in its fully open position. The staff finds it acceptable.

l 3.1.4 LTOP Review i

The staff has reviewed the licensee's justification for the unchanged LTOP system enable temperature and PORV actuation setpoint as discussed in Sections 3.1.1 and 3.1.2 above. The licensee has considered instrument uncertainties in its setpoint calculation using ISA S67.04-1994. The staff finds that the licensee's analyses were performed in a manner consistent with the approved methodology and that the results of the analyses conservatively demonstrated that the 10 CFR Part 50, Appendix G P/T limits up to 16 EFPYs will be adequately protected with these current LTOP setpoints, and, therefore, the staff finds the licensee's proposal acceptable.

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4 3.2 Pressure-Temperature Limit Curves 3.2.1 Basis for the Staff's Assessment The NRC has established requirements in 10 CFR Part 50 to protect the integrity of the reactor i

coolant pressure boundary in nuclear power plants. The staff evaluates the P-T limits based on the following NRC regulations and guidance: 10 CFR Part 50, Appendix G; Regulatory Guide

' 1.99, Revision 2; Standard Review Plan Section 5.3.2 (SRP 5.3.2); and Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code (Appendix G to the Code).

' Appendix G of 10 CFR Part 50, requires that the P/T limits for an operating plant must be at.

least as conservative as those that would be generated if the methods of Appendix G to the Code were. applied._ The basic parameter in Appendix G to the Code for calculating P/T limit curves is the stress intensity factor K,, which is a function of the stress state and flaw i

configuration. The methodology of Appendix G to the Code postulates the existence of a sharp

.j surface flaw in the reactor pressure vessel (RPV) that is normal to the direction of the maximum stress. The maximum flaw size in the RPV 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 methodology of Appendix G to the Code requires that the licensees determine the K, factors, which vary as a function of temperature, from the RCS operating temperatures and from the adjusted reference temperatures (ARTS) for the limiting materials in the RPV beltline

- region. The critical locations in the RPV beltline region for calculating the ARTS used in the

_ generation of the 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.

RG 1.99, " Radiation Embrittlement of Reactor Vessel Materials," Revision 2, provides an acceptable method for calculating the ARTS for ferritic RPV materials. The ART is defined as the. sum of the initial (unirradiated) reference temperature (initial RTuo7), the mean value of the adjustment in reference temperature caused by irradiation (ARTuor), and a margin (M) term.

I ARTuor is a product of a chemistry factor and a fluence factor. The chemistry factor (CF)is dependent upon the amount of copper and nickel in the material and may be determined from the tables in RG 1.99, Revision 2, or from surveillance data obtained from the plant's applicable reactor vessel material surveillance program. The fluence factor is dependent upon the neutron fluence _at the maximum postulated flaw depths. The margin term is dependent upon whether the initial RTuor is a plant-specific or a generic value and whether the chemistry factor was determined using the tables in RG 1.99, Revision 2, or surveillance data. The margin term is used to account for uncertainties in the values of initial RTuor, copper and nickel contents, fluence and calculational procedures. RG 1.99, Revision 2, also describes the methodology to be used in calculating the M term.

Appendix G of 10 CFR Part 50 imposes the following restrictions on the calculation of P/T limits for an operating nuclear plant:

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. During normal operation', at times when the reactor core is not critical or during o

hydrostatic pressure or leak rate testing, Appendix G of 10 CFR Part 50 requires that

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the P/T limits must be at least as conservative as those which would be generated by

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applying the methods of Appendix G to the Code; During normal operations when the reactor is critical, the requirement in Appendix G of o

L 10 CFR Part 50 changes by adding 40*F to the values that would be obtained through application of the methods in Appendix G to the Code.

Appendix G of 10 CFR Part 50 also imposes certain minimum temperature requirements (MTRs) on a RPV at a nuclear power plant. The degree of conservatism in the MTRs specified in Appendix G of 10 CFR Part 50 are dependent upon mode of operation, criticality of the core, and on degree of pressurization of the RCS relative to the preservice hydrostatic test pressure (PHTP). The MTRs for PWRs are summarized below:

For normal operating or pressure testing conditions of the RCS, when the RCS pressure o

is less than or equal to 20 percent of the PHTP and the reactor core is not critical, the MTR is equal to the limiting ART for the RPV closure flange.

l For pressure testing conditions of the RCS, when the RCS pressure is greater than 20 o

percent of the PHTP and the reactor core is not critical, the minimum temperature requirement is equal to the limiting ART for the RPV closure flange plus 90*F.

For normal operations, when the RCS pressure is greater than 20 percent of the PHTP o

and the reactor core is not critical, the MTR is equal to the limiting ART for the RPV closure flange plus 120*F.

For normal operations, when the RCS pressure is less than or equal to 20 percent of the o

PHTP and the reactor core h critica', the MTR is equal to the larger of either the minimum temperature for the inservice hydrostatic test or a temperature that is equal to the sum of the limiting ART for the RPV closure flange and 40*F.

For normal operations, when the RCS pressure is greater than 20 percent of the PHTP o

and the reactor core h critical, the MTR is equal to the larger of either the minimum temperature for the inservice hydrostatic test or a temperature that is equal to the sum of the limiting ART for the RPV closure flange and 160*F.

Table 1 of Appendix G of 10 CFR Part 50 summarizes these requirements in slightly more detail. The composite P/T limit curves are generated by superimposing the appropriate minimum temperature requirements over the most limiting generated P/T limit curves for the units, and selecting the most conservative P/T data to establish the limiting composite curves Appendix G to the Code considers normal operating conditions to include conditions of the plant during normal power operations of the reactor, during heatups and cooldowns of the reactor core either critical or not critical, and during anticipated operational transients.

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for the plants. Exemptions from complying with requirements of 10 CFR Part 50, Appendix G, must be requested when it is determined that the P/T limits for an operating plant do not meet the criteria stated in the rule.

3.2.2 Evaluation For the DCPP RPVs, PG&E provided the heatup, cooldown and pressure test curve figures for DCPP effective to 16 EFPY.2 The staff determined that PG&E opted to use the technical methods provided in non-mandatory Appendix A to Section XI of the ASME Code as the methodology for generating the DCPP P/T limit curves. PG&E's P/T limits are based on the ART values for the most limiting materials in the DCPP RPVs at the 1/4T and 3/4T locations.

To test the validity of PG&E's proposed curves, the staff performed an independent assessment of the licensee's submittal. The staff applied the methodologies of Appendix G to i

the Code and Appendix G of 10 CFR Part 50, as the bases for its independent assessment.

The assessment included an independent calculation of the 1/4T and 3/4T ART values effective to 16 EFPY for the limiting materials in the DCPP RPV beltlines, and independent generation of l

the P-T limit curves for the DCPP RPVs effective to 16 EFPY, For the evaluation of the limiting l

beltline materials, the calculations of the ART values were based on the methodology in RG 1.99, Revision 2.'

PG&E's proposed P/T limit curves for normal operating and pressure testing conditions, effective to 16 EFPY, are slightly more conservative than the P/T limit curves generated by the l

staff in accordance with the methods of Appendix G to the Code. The curves are in compliance with Appendix G of 10 CFR Part 50 and provide sufficient assurance that the DCPP reactors will be operated in a manner that will protect DCPP RPVs against brittle fracture. The staff confirmed that PG&E's P/T limit curves included the appropriate MTRs that were at least as conservative as those required by Appendix G of 10 CFR Part 50. Given these considerations, the staff therefore concludes that the proposed P/T limit curves, effective to 16 EFPY, are in compliance with Appendix G of 10 CFR Part 50 and acceptable for incorporation into the DCPP TS.

DCPP TS Figure 3.4-2 includes the proposed leak rate test curve, the heatup curve for the DCPP reactors when the cores are not critical, and heatup curve for the reactors when cores are critical.

The heatup curves are based on heatup rates up to 60*FMr. DCPP TS Figure 3.4-3 includes the cooldown curves for DCPP reactors at cooling rates of 0*FMr,25*FMr,50*FMr,75*FMr, and 100*FMr.

For the limiting DCPP beltline material, the staff confirmed that at16 EFPY, DCPP Unit 1 Lower Shell Longitudinal Weld 3442 C was the limiting beltline location for the assessment at the RPV 1/4T location, and that the ART value for this weld was appropriately calculated to be 183.7*F. This value is based on a fluence of 0.434E19 n/cm' and the latest weld chemistry provided in CEOG Task Report CE NPSD-1039, Rev. 2. The staff similarly confirmed DCPP Unit 2 Intermediate Shell Plate B5454-2 was the limiting beltline location for the assessment at the RPV 3/4T location, and that the ART value for this plate was appropriately calculated to be 151.4*F, This value is based on a fluence of 0.152E19 n/cm and the latest chemistry for the plates provided by PG&E. All calculations were confirmed to be in accordance with the methods in RG 1.99, Rev.2.

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4.0 STATE CONSULTATION

i in accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the amendments. The State official had no comments.

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. ENVIRONMFNTAL CONSIDERATION These amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve 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 amendments involve no l

significant hazards consideration, and there has been no public comment on such finding (63 FR 69345). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmentalimpact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

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

Principal Contributors: C. Y. Liang

- J. Medoff Date: May 3, 1999 I

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