LR-N08-0003, Supplement to License Amendment Request for Extended Power Uprate

Supplement to License Amendment Request for Extended Power Uprate

ML080290663
Person / Time
Site:	Hope Creek
Issue date:	01/16/2008
From:	Barnes G Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LCR-H05-01, Rev 1, LR-N08-0003
Download: ML080290663 (9)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG Nu clear LLC 10 CFR 50.90 LR-N08-0003 LCR H05-01, Rev. 1 JAN 16 2008 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Hope Creek Generating Station Facility Operating License No. NPF-57 NRC Docket No. 50-354

Subject:

Supplement to License Amendment Request for Extended Power Uprate

Reference:

1) Letter from George P. Barnes (PSEG Nuclear LLC) to USNRC, September 18, 2006

2) Letter from USNRC to William Levis (PSEG Nuclear LLC),

October 16, 2007

3) Letter from George P. Barnes (PSEG Nuclear LLC) to USNRC, March 22, 2007 In Reference.1, PSEG Nuclear. LLC (PSEG) requested an amendment to Facility...

Operating License NPF-57 and the Technical Specifications (TS) for the Hope Creek Generating Station to increase the maximum authorized power level to 3840 megawatts thermal (MWt).

Amendment No. 172 (Reference 2) removed values for turbine first stage pressure associated with Pbypass from the TSs. Pbypass is the reactor power level below which the turbine stop valve closure and the turbine control valve fast closure reactor protection system trip functions and the end-of-cycle recirculation pump trip are bypassed automatically. This change was also included in the proposed TS changes in Reference 1. Attachment 1 to this letter provides revised marked up TS pages reflecting issuance of Amendment No. 172. updates PSEG's response to NRC request for additional information (RAI) 5.3 (Reference 3) to reflect weld overlay repairs performed during the most recent refueling outage.

2 o(

95-2168 REV. 7/99

LR-N08-0003 LCR H05-01, Rev. 1 JAN 16 2008 Page 2 revises the description of changes to the reactor recirculation system runback logic.

PSEG has determined that the information contained in this letter and attachments does not alter the conclusions reached in the 1 OCFR50.92 no significant hazards analysis previously submitted.

There are no regulatory commitments contained within this letter.

Should you have any questions regarding this submittal, please contact Mr. Paul Duke at 856-339-1466.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on

/ -/6

- 0 (date)

Sincerely, George P. Barnes Site Vice President Hope Creek Generating Station Attachments (3)

1.

Revised Markup of Technical Specification Pages

2.

Updated Response to Request for Additional Information 5.3

3.

Supplement to Request for License Amendment cc:

S. Collins, Regional Administrator - NRC Region I J. Lamb, Project Manager - USNRC NRC Senior Resident Inspector-Hope Creek P. Mulligan, Manager IV, NJBNE

ATTACHMENT I Hope Creek Generating Station Facility Operating License NPF-57 Docket No. 50-354 Extended Power Uprate Revised Markup of Technical Specification Pages The following pages reflect changes to the Technical Specifications included in License Amendment No. 172:

TS Page 3/4 3-5 3/4 3-47

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a)

A channel'may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system.in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(b)

This function shall be automatically bypassed when the reactor mode switchi is in the Run position.

(c)

Unless adequate shutdown margin has been demonstrated per Specification 3.1.1, the "shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn*.

(d)

The non-coincident NMS reactor trip function logic is such that all channels go to both trip systems.

Therefore, when the "shorting links" are removed, the Minimum OPERABLE Channels Per the Trip System are 4

APRMS, 6 IRMS and 2 SRMS.

(e)

An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel.

(f)

This function is not required to be OPERABLE when the reactor pressure vessel head is removed 'per Specification 3;l0.1.

(g) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.

(h)

This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

Mi) With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(j) This function shall be automatically bypassed when turbin first.

stage pressure is equivalent to THERMAL POWER less tha(

of RATED THERMAL Wk) Also actuates the EOC-RPT system.

Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

HOPE CREEK 3/4 3-5 Amendment No.172

TABLE 3.3.4.2-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION TRIP FUNCTION

1.

Turbine Stop Valve - Closure

2.

Turbine Control Valve-Fast Closure MINIMUM OPERABLE CHANNELS PER TRIP SYSTEM"' 1 2 (b)

(b) 2 (a)A trip system may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance provided that the other trip system is OPERABLE.

(b)This function shall be automatically bypassed when turbine first stage pressure is equivalent to THERMALI POWER less tha fRATED THERMAL POWER.

HOPE CREEK 3/4 3-47 Amendment No. 1 7 2

ATTACHMENT 2 Hope Creek Generating Station Facility Operating License NPF-57 Docket No. 50-354 Extended Power Uprate Updated Response to Request for Additional Information In Reference 1, PSEG Nuclear LLC (PSEG) requested an amendment to Facility Operating License NPF-57 and the Technical Specifications (TS) for the Hope Creek Generating Station (HCGS) to increase the maximum authorized power level to 3840 megawatts thermal (MWt).

In Reference 2, the NRC requested additional information concerning PSEG's request.

PSEG's original response to RAI 5.3 describing welded overlays was provided in Reference 3. During the most recent refueling outage, PSEG performed an additional weld overlay. The updated response to RAI 5.3 is provided below.

5)

Piping & NDE Branch (CPNB) 5.3 Identify all flawed components including overlay repaired welds that have been accepted for continued service by analytical evaluation based on American Society of Mechanical Engineers (ASME),Section XI rules. Discuss the adequacy of such analysis considering the effect of the EPU on the flaws.

Updated Response Hope Creek plant has three welded overlays on ASME Section Xl flawed components. There are no ASME Section Xl flawed components that have been accepted for continued service by analytical evaluation. The three overlays are on the reactor vessel core spray nozzle to safe end weld (N5B); the reactor vessel recirculation inlet nozzle to safe end weld (N2K); and the reactor vessel recirculation inlet nozzle to safe end weld (N2A).

The core spray overlay was verified adequate for EPU operation. At the core spray nozzle location, there is a slight (0.2%) change in temperature, but no change in pressure or flow due to EPU. Hence, the change in temperature has an insignificant effect on P + Q stresses and the fatigue usage for EPU.

The N2K recirculation inlet overlay was verified adequate for EPU operation. At the recirculation inlet nozzle location, there is a slight increase in pressure (1.1%), a slight decrease in temperature (-0.2%), and an increase in recirculation flow (3.4%). The pressure and temperature operating conditions used in the overlay analysis bound the EPU temperature and pressure conditions. For the

LR-N08-0003 Page 2 flow increase, the change in the heat transfer coefficient used in the analysis is 2.7%, which is considered insignificant.

The design inputs for the N2A recirculation inlet overlay bound the EPU operating conditions.

The three overlays on the Hope Creek plant reactor vessel nozzles are adequate for EPU conditions.

References

1)

Letter from George P. Barnes (PSEG Nuclear LLC) to USNRC, September 18, 2006

2)

Letter from USNRC to William Levis (PSEG Nuclear LLC), February 23, 2007

3)

Letter from George P. Barnes (PSEG Nuclear LLC) to USNRC, March 22, 2007

ATTACHMENT 3 Hope Creek Generating Station Facility Operating License NPF-57 Docket No. 50-354 Extended Power Uprate Supplement to Request for License Amendment In Reference 1, PSEG Nuclear LLC (PSEG) requested an amendment to Facility Operating License NPF-57 and the Technical Specifications (TS) for the Hope Creek Generating Station (HCGS) to increase the maximum authorized power level to 3840 megawatts thermal (MWt). Attachment 4 to Reference 1, NEDC-33076P, Rev. 2, "Safety Analysis Report for Hope Creek Constant Pressure Power Uprate," described planned changes to the reactor recirculation system (RRS) runback logic.

The planned changes included revising the Secondary Condensate Pump (SCP) permissive setpoint so that the RRS runback logic would be armed when feedwater (FW) flow is more than 85% of the EPU rated FW flow. However, PSEG subsequently elected to maintain the SCP permissive setpoint at approximately the same FW flow (in MIb/hour) as the pre-EPU setpoint. Thus, after EPU implementation, the RRS runback logic for both PCP and SCP trips will be armed when FW flow is greater than approximately 75% of EPU rated FW flow.

The instrument setpoint changes implemented for EPU are shown below. For EPU, the PCP and SCP permissive nominal setpoints are equal to the percent flow span when FW flow is 75% of rated flow. The setpoints provided previously in Table 5-3 in NEDC-33076P, Revision 2, included allowances for instrument loop uncertainties which are not required, given the flow capacity remaining after the trip of a single PCP or SCP.

Changes from the information provided in NEDC-33076P, Revision 2, are marked by a revision bar in the margin.

CPPU Parameter / Device Nominal Setpoint Primary Condensate Pump 75% Permissive 2 % FW flow span) 62.8 Secondary Condensate Pump 75% Permissive (% FW flow span) 62.8

2.

The Digital FW Control System processes the signals from both FW flow transmitters for the permissives for the Primary and Secondary Condensate Pumps. The setpoints are revised to set both the PCP and SCP trips at the same flow rate. Since both RRS runbacks will be intermediate runbacks at CPPU, the setpoints were chosen to be comparable to the current FW flow rates for the SCP trip.

LR-N08-0003 Page 2 These changes to the RRS permissive setpoints do not effect the results or conclusions of the analyses described in Reference 2 for trip of a reactor feed pump, PCP or SCP (RAI Responses 7.16, 7.17 and 7.18).

References

1)

Letter from George P. Barnes (PSEG Nuclear LLC) to USNRC, September 18, 2006

2)

Letter from George P. Barnes (PSEG Nuclear LLC) to USNRC, June 22, 2007