ML20203M026

Safety Evaluation Supporting Amend 157 to License NPF-49

ML20203M026
Person / Time
Site:	Millstone
Issue date:	02/12/1998
From:	NRC (Affiliation Not Assigned)
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NUDOCS 9803060359
Download: ML20203M026 (6)

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UNITED STATE 8 g

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WASHINGTON, D.C. 30006-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REG,ULATION RFI ATED TO AMENDMENT NO.157 TO FACILITY OPERATING LICENSE NO. NPF-49 NORTHEAST NUCLEAR ENERGY COMPANY. ET AL.

Mit i RTONE NUCLEAR POWER STATION. UNIT NO. 3 QQQ'ET NO. 50-423

1.0 INTRODUCTION

By letter dated August 29,1997, as supplemented by letters dated September 25 and l

November 14,1997, the Northeast Nuclear Energy Company, et al. (the licensee), submitted a requerl for changes to the Millstone Nuclear Power Station, Unit No. 3 Technical Specifications (TS). 3ased on a review and subsequent calculations of the cold overpressurization protection (COPS) enabling temperature and the emergency core cooling system (ECCS)/ charging system rnode 3 requirements, the licensee proposes to reduce the COPS enabling temperature. As a result, the licensee proposed the following TS changes: add new heatup and cooldown pressure /

temperature (P-T) limit curves and their associated requirements; add new power operated ratief valve (PORV) setpoint curves and their associated requirements; revise the reactor coolan+ '. ps and coolant circulation, ECCS, boration systems, and COPS to incorporate the lower enabling temperature and new restrictions for cold overpressure protection system, FORV undershoot, and residual heat removal (RHR) relief valve bellows; add a footnote to allow an reactor coolant pump (RCP) to substitute for an RHR pump during heatup f >m Mode 5 to 4 which is consistent with the improved standard technical specification (STS); reword TS 3/4.4.9.3 and its serveillence requirement to be consistent with the improved STS; and revise the affected Bases

- sections to ne consistent with the proposed changes. The September 25 and November 14, 1997, letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination.

2.0 - - PRESSURE / TEMPERATURE LIMITS

2.1 Background

The August 29,1997, license amendment request was intended to update the Millstone Unit 3 P-T curves for 10 effective full power years (EFPT, using the latest vessel beltline material and

- fluence data. In addition, the licensee propcsed more restrictions to the P-T limit limiting condition for operation (LCO), and proposed changes to tne surveillance requirements in the TS for botter clarity.

The staff evaluates the P-T limns based on the following NRC regulations and guidance: 10 CFR Part 50, Appendix G; Generic Letter (GL) 88-11; GL 92-01, Revision 1; GL 92-01, Revision 1, Supplement 1; Regulatory Guide (RG) 1.39, Revision 2 (Rev. 2); and Standard Review Plan 9803060359 980212 l

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2 (SRP) Section 5.3.2. GL 88-11 advised licensees that the staff would use RG 1,91, Rev. 2, te review P-T limit curves. RG 1.99, Rev. 2, contains methodoiogies for determining the increare in transhion temperature and the decrease in upper shelf energy (USE) resulting from neutron 7diation. GL 92-01, Rev.1, requested that licensees submit their reactor pressure vessel (RPV) data for their plants to the staff for review. GL 92-01, Rev.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 ths basis for the staffs saview of P T limit submittals, and as the basis for the staffs review of pressurized thermal shock (PTS) assessments (10 CFR 50.61 assessments). Appendix G to 10 CFR Part 50 requires that P-T limits for tha RPV be at least as conservative as those obtained by applying the methodology of Appendix G to Section XI of the American Society of Mechanical Engineers Boller and Pressure Vessel Code (ASME Code).

SRP 5.3.2 providet an acceptable method of calculating the P-T limits for ferritic materials in the beltline of the RPV based on the linear elastic fractere 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 of Appendix G pstulate the existence of a sharp surface flaw in the RPV that is normcl to the direction of the maximum stress. This flawis 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) locClons, which correspond to the depth of the maximum postulated flaw, if initiated and grown from the inside and outtide surfaces of the RPV, respectively.

The Appendix G, ASME Code methodology requires that 1%nsees determine the adjusted reference temperature (ART or RT ) and the Charps * "iE at the maximum postulated flaw depth. The ART is defined as the sum of the initial (u.arradiated) reference temperature (initial RTc), the mean value ci the adjustmont in referer.co temperature caused by irradiation (ARTc), and a margin (M) term. The ART,c7 s a product of a chemistry factor and a fluence i

factor. The chemistry factor is dependent upon the amount of copper and nickelin the material and may be determined from tables in RG 1.99, Rev 2, or from surveillance data. The 'luence factor is dependent upon the neutron fluence at the maximum postulated flaw depth. Tne margin term is dependent upon whether the initial RTc is a plant specific or a generic value and whether the chemistry factor was determined using the tables in RG 1.99, Rev. 2, or surveillance data. The margin term is used to account for uncertainties in the values of initial RTo, copper and nickel contents, fluence and calculational procedures. RG 1.99, Rev. 2, describes the methodology to be used in calculating the margin term.

2.1 Evaluation The staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the reactor vessel of Millstone Unit 3. The amount ofirradiation embrittlement was calculated in accordance with RG 1.99, Rev. 2. The staff has determined that the material with the highest ART at 10 EFPY is the intermediate shell B9805-1, with 0.05% copper (Cu), 0.64% nickel (Ni),

and an initial RTc of 60 'F. The ART calculated by the staff is 111.8 *F for the limiting material.

This ART was calculated at 1/4T at 10 EFPY with corresponding neutron fluence of 2

0.556E19 n/cm. The ART calculated by the licenae, using the Chemistry Factor Table in Section 1.1 of RG 1.99, Rev. 2. is 111.9 *F. Both the staff and the licensee did not include the cladding thickness in calculating the attenuation of the fluence through the vessel wall.

____ _ __ The ARTS calculated oy the staff and the licensee are almost identical. However, substituting

- the ART of 111.9 'F into equations in SRP 5.3.2, the staff could not verify the proposed P T limit curves. In its response to the staff's request, the licensee, in a letter dated September 25,1997, provided details of the methodology used. This information revealed that Millstone Unit 3 has some unittue features built into the P-T Hmits: (1) the proposed P-T limits considered indiutor uncertainties,22 'F for temperature ano 129 psia for pressure; (2) it considered a pressure drop of 28.3 psi between the pressure transmitter and the reactor vessel beltline for one pump operation and a pressure drop of 74 psi for four pump operation; (3) the cooldown cuive is made

' of the conservative portions of the P-T limits at a cooldown rate of 0 'F/hr and the P-T limits at a cooldown rate of 80 'F/hr to 160 'F and 40 'F/hr to 60 'F; and (4) the pressure was expressed it' " psia," with a value of 10 added to the gage pressure to account for the primary containment pressu:e. Aften considering the extra conservatism in (1) (2), and (3), the staff confirmed the proposed P-T liinits (for 10 EFPY) for hostup, cooldown, and hydrotest and concluded thst they meet the beltline material requirements in Appendix G of 10 CFR Part 50. The staff found that all equations involving the conversion of the pressure in psig to psia in Reference 2 had a sign error.

However, from the intermediate results presented, the staff verified that the correct sign was used in actual calculations.

In add, tion to beltline materials, Appendix G of 10 CFR Part 50 imposes P-T limits based on the reference temperature for the reactor vessel cloture flange materials.Section IV.A.2 of Appendix G states that when the pressure exceeds 20% of the preservice system hydrostatic test pressure, the temprature of the closure tiar.ge regions highly stressed by the bolt preload must exceed the reference temperature of the materialin those regions by at least 120 'F for normal operation and by 90 'F for hydrostatic pressure tests and leak tests. - Based on the -

limiting r,oule shell reference temperatures of 40 'F, the minimum allowable temperature of this region is 160 'F. As mentioned previously, the licensee adders 22 'F to 160 'F to account for l

indicator uncertainty. These limits are shown as straight-line segments on Figures 3.4 2 and 3.4-3 of the submittal, and the staff has determined that the proposed P-T limits satisfy the -

requirements in Section IV.A.2 of Appendix G.

In the fS, the licerasee proposed to further re,,trict the LCO under "3/4.4.9 Pressure / Temperature Limits" by limiting one pump in operation when the cold leg temperature is less than 160 'F and

. no pumps in operation when the cold leg temperature is less than 120 'F. The staff considered these measures usefulin preventing overpressurization of the RPV at low temperatures. The proposed changes to the surveillance requirements in the TS are for more clarity, and are also acceptable.

Appendix G further requirer. that the predicted Cl.arpy USE at end-of-license (EOL) for vessel beltline materials be above 50 ft lb or that licensees demonstrate that lower values of Charpy USE will provide margins of safety equivalent to those required by Appendix G of Section XI of the ASME Code. This USE requirement is satisfied because all beltline materials have EOL USEs above 50 ft lo.

2.3 Overall The staff concludes that the proposed P-T limits for the reactor coolant system for heatup, cooldown, leak test, and criticality are valid as indicated on the curves. The P-T limits satisfy the requirements of Appendix G of 10 CFR Part 50 for 10 EFPY. The proposed P-T limits also satisfy Generic Letter 88-11 because the method in RG 1.99, Rev. 2, was used to calculate the ART. Hence, the proposed P-T limits may be incorporated into the Millstene Unit 3 TS. The j

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proposed changes to a more restrictive P-T limit LCO and to the surveillance requirements in the T3 are riso acceptable.

3.0 COLD OVERP8@SURE PROTECTION SYSTEM 3.1 Eackground The COPS mitigates overpressurt, transients at inw temperatures so that the integrity of the reactor coolant pressure boundary is not compromised by violating the 10 CFR Part 50, Appendix G, P-T limits under steady state operating conditions. Millstone Unit 3 COPS uses the combination of FORV and the RHR suction relief valves or an RCS vent with the reactor depressurized to accomplish this function. The system is manually enabled by operators and uses Asrimble setpoints as the lift pressure for the PORV and a singla setpoint for RHR suction relief valves. The design basis of Millstene Unit 3 COPS considers both mass-addition and heat.

addition transients. The mass-addition analyses account for the injection from one centrifugal charging pump to the RCS. The heat addition analyses accounts for heat input from the secondary side of the steam generators into the RCS upon starting a single RCP under RCS operational configurations and temperatures.

In October 1996 and May 1997, the licensee identified inconsistencies and nonconservatisms in the current TS with respect to COPS at Millstone Unit 3. In response to its commitments made in Licensee Event Reports96-038 and 97-030, the licenseo has evaluated the TS and Coos at Millstone Unit 3 and by letter dated August 29,1997, proposed changes to TS 3.1.2,3.4.1, * ' 9, 3.5.3, and their associated Bases in the areas that are affected by the modified COPt it Millstone Ur,it 3. The proposed changes will ensure that the TS with its associuted bases are consistent and conservative with respect to COPS for Matone Unit 3 up to 10 EFPY of reactor operation. The proposed TS provided restrictions in plant operation within the configuration s

assumed in the analysis for COPS design.

The Millstone Unit 3 prop 0 sed COPS er,able temperature and the PORV actuation setpoint were established using a plant-specific mathodology similar to that presented in WCAP-14040, Revision 1. However, the licensee for Millstor,e Unit 3 chose to keep its COPS setpoints in its TS.

3.2 Evaluation The proposed LCO in TS 3.4.9.3 requires that a COPS shall be operable with a maximum of one centrifugal charging pump and no safety injection pumps capable of injecting into the RCS and (1) any combination of two operable relief valves (PORVs or RHR suction relief valves), or (2)

RCS depressurized with an RCS vent of 2 5.4 square inches. This LCO is applicable when any RCS celd leg temperature is s 275 'F when the head is on the reactor vessel. Consistent with this LCO, the licensee proposed modifications to TS 3.1.2, 3.4.1,3.4.9.1,3.4.9.2, and TS 3.5.3 to restrict the plant operational configuration consistent with the design of the COPS. The staff's evaluation of the COPS setpoints is presented below.

3.2.1 Enable Temperature The COPS enable temperature is the temperature below which the COPS system is required to be operable. The licensee proposed to establish an COPS enable temperature methodology to:

(t) account for instrument uncertaintins associated with the instrumentation used to enabis the COPS system and, (2) implement the NRC Weh technical position (BTP) RSB 5-2 of using an enable RCS water temperature corresponding to a metal temperature of at least RT,,m + 90 'F at

_ _ ___ the belt line location (1/4T or 3/4T) Therefore, the licensee proposed to calculate the enable temperature as rte + 90 'F + temperature difference between RCS and metal + instrument uncertainties. Using the above equation, the calculated minimum enable temperature is 244 'F.

The licensee proposed an enable temperature of 275 'F that includes an additional margin of 31 'F.

'ihe staff finds that this proposed COPS enabie temperature is conservative with respect to the c

enable temperature allowed by NRC (DTP) RSB 5-2 and therefore, is acceptable.

S 3.2.2 COPS Actuation Setpoint COPS is designed to mitigate overpressure transients at low temperatures to prevent violating 10 CFR Part 50, Appendix G P-T limits. Additionally, since overpressure events most likely occur dur'ng 'sothermal con (tions in the RCS, the NRC has accepted the use of the steady-state Appendix G limits for the design of COPS. The COPS actuation setpoint is the pressure at which the pressure relief valves willlift, when the COPS ir enabled, to limit the peak RCS pressure during a presmrization transient.

Millstone Un.t 3 uses PORVs and RHR saction relief valves to provide pressure relief capacity for COPS. The methodology used for determining the PORV actuation setpoint is consistent v4th the methodology presented in WCAP-14040, Revision 1.

Tht 'icensee proposed PORV actuation setpoints in TS Figures 3.4-4a and 3.4-4b were calculated in accordance with the proposed methodology. In response to the staff request, the licensee, in its letter dated November 14,1997, provided a tabulation to list PORV setpoint, transient pressure overshoot, instrumentation uncertalrdies for temperature and pressure and corresponding P-T limit under various temperature conditions below the COPS enable temperature. The data presented in this tabulation confirms that the proposed PORV setpoints will provide adequate protection to the 10 CFR Part 50, Appendix G, P T limits under steady state conditions during a design basis c.corpressure transient (mass addition or heat-addition) as described in Section 3.1 of this safety waluation. Based on the preceding discussion, the staff finds thc proposed PORV setpoint acceptable.

Millstone Unit 3 usea the RHR suction relief valves as an attemative to PORVs for COPS. The RHR suction relief valves are designed with setpoint of s440 psig +/ 3%. The licer,see has performed an evaluation t. verify the adequacy of this relief valve setpoint. The results of its calculation confirmed that the Millstone Unit 3 RHR suction relief valve with the proposed setpoint will adequately mitigate the most limiting overpressure transient and protect the 10 CFR Part 50, Appendix G P-T limits under steady state operating conditions. Also, the design wessure of the RHR discharge pipin0 (600 psig) would not be exceeded during the overpressure tunsient. The staff finds the RHR suction relief valve setpoint acceptable.

3.2.3, RCS Vent Size With the RCS depressurized, the results of the licer.see's analysis show that a vent size of 5.4 square inches is capable of mitigating the most limiting cold overpressure transient. The capability of a vent this size is greater than the flow resulting from this limiting mass-addition transient. The staff finds this vent size acceptable.

6-3.3 Overall The staff has reviewed the licensee's proposed TS 3.4.9.3 for the COPS enable temperature and actuation setpoint. The staff also reviewed the licensee's analyses relateC to the proposed enable temperature of 275 *F and actuation setpoints as discussed in Sections 3.1 and 3.2 above.- The licenwe 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 will be adequately protected with

- these setpoints end, therefore, f;nds the proposed TS 3.4.9.3 with its associated Bases regarding COPS acceptatis.

The staff has also reviewed the licensee's proposed TS 3.1.2,3.4.1, 3.4.9.1, 3.4.9.2, and 3.5.3 and finds that these proposed TS provide adequate restrictions of plant operation to support the design of CCPS at Millstone Unit 3, and therefore, are acceptable.

4.0 STATE CONSULTATION

in accordance with the Commission's regulations, an attempt to cor. tact the Connecticut State

- official was made, however, the official could not be contacted.

5.0 ENVIRONMENTALSONSIDERATION The amendmen* changes a requirement with respect to installation or use of a facility componsnt l

located within the restricted area as defined in 10 CFR Part 70 and changes surveillance i

requirements. The NRC staff has deterrained that the amendment involves no significant increase in the amouats, 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 ha::ards considsration, and there has been no public comment on such finding ( 62 FR 52583 cated December 8,1997). Accordgly, the amendment meets the eligibility criteria for categorical exclusion set forth in 'i0 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared !n connection with the issuance of the amendment.

6.0 CONCLUSION

Tne 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 propon:1 manner, (2) such activities will be conducted in compliance with the Commission's regulatiLc,s, and (3) the issuance of the amendment will not be inimicci to the common defense and security or to the health and safety of the public.

Principal Contributors: S. Gheng C. Liang Date:

February 12, 1998 e

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