LR-N07-0099, Response to Request for Additional Information Request for License Amendment, Extended Power Uprate

Response to Request for Additional Information Request for License Amendment, Extended Power Uprate

ML071290559
Person / Time
Site:	Hope Creek
Issue date:	04/30/2007
From:	Barnes G Public Service Electric & Gas Co
To:	Document Control Desk, NRC/NRR/ADRO
References
LCR H05-01, Rev 1, LR-N07-0099
Download: ML071290559 (306)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 PSEG Nuclear LLC 10 CFR 50.90 LR-N07-0099 LCR H05-01, Rev. 1 April 30, 2007 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:

Response to Request for Additional Information Request for License Amendment - 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, April 20, 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 (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.

Attachments 1 and 3 to this letter restate the NRC questions and provide PSEG's responses. As noted in Reference 2, responses to questions 3.58 through 3.65, 3.8 and question 13.16 will be provided by May 10, 2007.

The response to question 14.41 will be submitted on or before May 10, 2007. The response is delayed to ensure it includes the most current and accurate information available using inputs from the revised PEPSE analysis. The need to update the PEPSE analysis of the plant heat balance at EPU conditions to reflect plant specific turbine information was identified during the Independent Design Verification of the EPU Design Change Package. Efforts to update the PEPSE model were completed in April 2007.

4o EV

\92168+REv. 7/99

LR-N07-0099 LCR H05-01, Rev. 1 April 30, 2007 Page 2 PSEG has determined that the information contained in this letter and attachment does not alter the conclusions reached in the 10CFR50.92 no significant hazards analysis previously submitted.

There are no regulatory commitments contained within this letter. contains information proprietary to General Electric Company (GE). GE requests that the proprietary information in Attachment 1 be withheld from public disclosure in accordance with 10 CFR 9.17(a)(4) and 2.390(a)(4). An affidavit supporting this request is included with Attachment 1. Attachment 1 also contains information proprietary to Continuum Dynamics, Inc. (C.D.I.). C.D.I. requests that the proprietary information in Attachment 1 be withheld from public disclosure in accordance with 10 CFR 2.390(a)(4). An affidavit supporting this request is included with Attachment 1.

A non-proprietary version of PSEG's Attachment 1 responses is provided in .

Structural Integrity Associates report HC 31Q-301, "Hope Creek RPV Dome Dynamic Pressure Data Reduction," is provided in Attachment 4 in response to the NRC staffs request.

PSEG has updated Attachment 7 (Steam Dryer Evaluation) to the Reference 1 submittal. The updated Steam Dryer Evaluation is provided in Attachment 5 to this letter. Attachment 5 replaces, in its entirety, the previous version of the evaluation provided in Reference 1.

In response to NRC staff requests, updated C.D.I. Reports06-16P, "Estimating High Frequency Flow Induced Vibration in the Main Steam Lines at Hope Creek Unit 1: A Subscale Four Line Investigation of Standpipe Behavior," and CDI Report No. 06-17, "Hydrodynamic Loads on Hope Creek Unit 1 Steam Dryer to 200 Hz," are provided in Attachments 6 and 7 to this letter. C.D.I. Report 06-16P contains information which C.D.I. considers to be proprietary. C.D.I. requests that the proprietary information be withheld from public disclosure in accordance with 10 CFR 2.390(a)(4). An affidavit supporting this request is provided in Attachment 6. PSEG will provide a non-proprietary version of the report suitable for public disclosure by May 10, 2007.

The Power Ascension Test Plan is provide in Attachment 8 in response to the NRC staff's request.

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

LR-N07-0099 LCR H05-01, Rev. 1 April 30, 2007 Page 3 I declare under penalty of perjury that the foregoing is true and correct.

Executed on '/3 /67 (date)

Sincerely, George P. Barnes Site Vice President Hope Creek Generating Station

LR-N07-0099 LCR H05-01, Rev. 1 April 30, 2007 Page 4 Attachments (8)

1. Response to Request for Additional Information (proprietary)

2. Response to Request for Additional Information (non-proprietary)

3. Response to Request for Additional Information 14.46

4. SIA Report HC 31Q-301, "Hope Creek RPV Dome Dynamic Pressure Data Reduction"

5. LCR H05-01, Rev. 1, Attachment 7, Rev. 1

6. CDI Report 06-16P, Rev. 2

7. CDI Report 06-17, Rev. 3

8. Power Ascension Test Plan cc: S. Collins, Regional Administrator - NRC Region I J. Shea, Project Manager - USNRC NRC Senior Resident Inspector - Hope Creek K. Tosch, Manager IV, NJBNE

General Electric Company AFFIDAVIT I, George B. Stramback, state as follows:

(1) 1 am Manager, Regulatory Services, General Electric Company ("GE") and have, been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in Enclosure 1 of GE's letter, GE-HCGS-EPU-669, Edward D. Schrull (GE) to Larry Curran (PSEG), Transmittal

- Response to Request for Additional Information (RAI) Regarding Amendment Application for Hope Creek Generating Station Extended Power Uprate - RAIs 3.58, 3.59, 3.60, 3.63, 3.64, 7.11, 14.42, and 14.49, GE Proprietary Information, dated April 20, 2007. The proprietary information in Enclosure 1, which is entitled GE Responses to NRC RAIs 3.58, 3.59, 3.60, 3.63, 3.64, 7.11, 14.42, and 14.49, is delineated by a double underline inside double square brackets. Figures and large equation objects are identified with double square brackets before and after the object. In each case, the superscript notation 3) refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act,, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the.'narrower definition of "trade secret", within, the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Proiect v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product; GBS-07-02-afGE-HCGS-EPU-669 EPU RAIs 4-20-07.doc Affidavit Page I

c. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, resulting in potential products to General Electric;

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a., and (4)b, above.

(5) To address 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GE, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and

• bythe Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed information about the results of analytical models, methods and processes, including computer codes, which GE has developed, obtained NRC approval of, and applied to perform evaluations of loss-of-coolant accident events in the GE Boiling Water Reactor ("BWR"). The development and approval of the BWR loss-of-coolant accident analysis computer codes was achieved at a significant cost to GE, on the order of several million dollars.

The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GE asset.

GBS-07-02-af GE-HCGS-EPU-669 EPU RAIs 4-20-07.doc Affidavit Page 2

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm, to GE's competitive :position and foreclose or reduce, the.

availability of profit-making opportunities. The information is part of. GE's comprehensive BWR safety and technology base, and its commercial value extends, beyond the original development cost.. The..value.of the technology base goes..

beyond the extensive physical database and analytical methodology and includes' development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GE.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly, is substantial.

GE's competitive advantage will be lost if its competitors are able to use the results of the GE experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GE would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in:

developing these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated-therein are true and correct to the best of my knowledge, information, and belief.

Executed on this 20h day of April 2007.

/J9 George B. Stramback General Electric Company GBS-07-02-af GE-HCGS-EPU-669 EPU RAIs 4-20-07.doc Affidavit Page 3

4Continuum Dynamics, Inc.

(609) 538-0444 (609) 538-0464 fax 34 Lexington Avenue Ewing, NJ 08618-2302 AFFIDAVIT Re: Information Requested by "NRC Request for Additional Information Regarding the Request for Extended Power Uprate (TAC No. MD3002)"

and transmitted to the NRC by PSEG Nuclear Letter LR-N07-0099 I, Barbara A. Agans, being duly sworn, depose and state as follows:

-1. I hold the position of Director, Business Administration of Continuum Dynamics, Inc. (hereinafter referred to as C.D.I.), and I am authorized to make the request for withholding from Public Record the Information contained in the documents described in Paragraph 2. This Affidavit is submitted to the Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 2.390(a)(4) based on the fact that the attached information consists of trade secret(s) of C.D.I. and that the NRC will receive the information from C.D.I. under privilege and in confidence.

2. The Information sought to be withheld, as transmitted to PSEG Nuclear LLC as attachment to C.D.I. Letter No. 07078 dated 27 April 2007 Information Requested by "NRC Request for Additional Information Regarding the Request for Extended Power Uprate (TAC No. MD3002)" and transmitted to the NRC by PSEG Nuclear Letter LR-N07-0099.

3. The Information summarizes:

(a) a process or method, including supporting data and analysis, where prevention of its use by C.D.I.'s competitors without license from C.D.I. constitutes a competitive advantageover other companies; (b) Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product; (c) Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 3(a), 3(b) and 3(c) above.

4. The Information has been held in confidence by C.D.LI, its owner. The Information has consistently been held in confidence by C.D.I. and no public disclosure has been made and it is not available to the public. All disclosures to

third parties, which have been limited, have been made pursuant to the terms and conditions contained in C.D.I,'s Nondisclosure Secrecy Agreement which must be fully executed prior to disclosure.

5. The Information is a type customarily held in confidence by C.D.I. and thereis a rational basis therefore, The Information is a type, which C.D.I. considers trade, secret and is held in confidence by C.D.I, because it constitutes a source .of competitive advantage in the competition and performance. of-such work in the, industry. Public disclosure of the Information is likely to cause substantial harm to C.DI.'s competitive position and foreclose or reduce the availability of profit-,

making opportunities.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to be the best of my knowledge, information and belief.

Executedon this day of 2007.

l4arbara A. Agans Continuum Dynamics, Inc.

Subscribed and sworn before me this day: a-7o q) *

-ye un"ft rmei s'9,Wblhc EILEEN P BURMEISTER NOTARY PUBLIC OF NEW JERSEY MY COMM. EXPIRES MAY 6, 2007 LR-N07-0099 LCR H05-01, Rev. 1 Hope Creek Generating Station Facility Operating License NPF-57 Docket No. 50-354 Extended Power Uprate 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.

Each NRC question is restated below followed by PSEG's response.

14) Mechanical & Civil Engineering Branch (EMCB) 14.1 Question Superceded by new Steam Dryer Data 14.2 Please submit bias errors and uncertainties for:

(a) The steam dryer finite element model (FEM) as to how accurate are the stress/force transfer function amplitudes (note this does not question the damping assumed by PSEG, but is related to the accuracy of the modal masses computed by the model, which dictates the mean transfer function amplitude), and (b) the lack of rigorous analyses at time/frequency shifts of +/-2.5%, +/-5%,

and +/-7.5%.

Response

LR-N07-0099 LCR H05-01, Rev. 1 LR-N07-0099 LCR H05-01, Rev. 1 LR-N07-0099 LCR H05-01, Rev. 1

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14.3 ((

LR-N07-0099 LCR H05-01, Rev. 1 1]

Response

PSEG in CDI Report 06-17 section 3.2 it states that the HCGS and Susquehanna steam delivery systems are 'nearly identical', but in the same paragraph it clarifies the similarities and the differences. The differences are the location of the standpipes and that Susquehanna has dead-headed branch lines off the main steam lines whereas HCGS does not.

((

These sources and energy storage structures do not exist in the HCGS plant.

Refer to RAI 14.60 for further details.

14.4 Your assertion that the Hope Creek ACM accuracy is negatively (conservatively) biased does not appear to address significant ACM underpredictions of the Quad Cities Unit 2 (QC2) 156Hz dryer loading peak. Since the largest load in the Hope Creek interim EPU dryer analysis is a safety relief valve (SRV) singing peak (based on the scale model test (SMT)), provide an ACM bias error based on the tones observed in QC2 during the instrumented dryer ACM validation studies.

Response

((

+ i-LR-N07-0099 LCR H05-01, Rev. 1 I]

14.5 In the Hope Creek EPU submittal Attachment 7 (Steam Dryer Evaluation), you determined a net dryer loading uncertainty of 27.2%, this uncertainty is not applied to the interim EPU dryer stress estimates. Explain how the estimated 27.2% uncertainty is applied to the Hope Creek dryer stress assessment.

Response

As stated in page 12 of Attachment 7 to the EPU LCR, PSEG elected to report the stress ratios as calculated in the finite element analysis without adding the uncertainty factor. This avoids the necessity of changing the finite element analysis results if the uncertainty values are changed.

The discussion below applies to the alternating stress ratios which consist entirely of stresses from flow induced vibration whereas, the peak stress ratio has a significant component from dead weight which can be accurately calculated.

Considering uncertainty, the alternating stress ratio, would be the reported value in the Finite Element Report divided by (1 + uncertainty). For example, at nominal frequency, the minimum alternating stress ratio from CDI 06-27 R2 table 7a is 1.96. With an overall uncertainty of + 27.2%, this alternating stress ratio is between 2.70 and 1.54, (1.96 / (1 +.272)).

14.6 Confirm whether PSEG has planned for any dryer or reactor pressure vessel (RPV) dynamic measurements which would quantify the low frequency (less than 50 Hz) loads acting on the dryer. If so, please specify.

Response

No in-plant measurements are planned. As discussed in the response to RAI 14.3, HCGS does not have a geometry that is believed to contribute to significant low frequency loadings.

14.7 Please provide the computational fluid dynamics (CFD) analyses results for the Hope Creek steam dryer at current licensed thermal power (CLTP) and EPU conditions.

Response

The CFD calculation was not used in the steam dryer analysis. As stated in LCR Attachment 7, the CFD was only used to provide a qualitative assessment that no new flow phenomenon was being created by EPU.

In mid-2005 PSEG undertook an exploratory investigation with a vendor (not CDI) into the possibilities that Computational Fluid Dynamics (CFD) could be LR-N07-0099 LCR H05-01, Rev. 1 used to generate steam dryer loads. This effort was abandoned before the end of 2005. Towards late 2005, PSEG had discussions with its primary vendor, CDI, who is also knowledgeable on CFD calculations. CDI provided PSEG with the opinion that it would be difficult to validate a CFD analysis for this application.

Furthermore, the CFD simulations PSEG obtained from an earlier vendor could not be considered acceptable for defining loads since they were generated by a vendor without the prerequisite quality assurances for nuclear safety work and without any benchmarking of the CFD program. Since late 2005, no further effort was expended on CFD calculations.

PSEG elected to define the in-plant loads by Acoustic Circuit Modeling and to predict EPU changes/loads by the CDI 1/8th scale model test. The ACM has become the tool for estimating steam dryer loads from plant data used by all other pending EPU submittals and has been subject to benchmarking at QC2.

The CDI SMT has recently been benchmarked, at CLTP, against HCGS plant data (CDI Report 07-01 P). This is in line with the approach endorsed by the BWR Owners Group in late 2006.

14.8 In the Hope Creek EPU submittal Attachment 20 (Continuum [Dynamics], Inc.

(CDI) Report No.05-28P), CDI argues that ((

I]

Response

The strength of the ACM is that it does not need to identify or model the source of an input signal, e.g., SRV acoustic resonance. The ACM models the steam dome area, the steam dryer, and the portion of the MSL through the strain gage locations. The ACM converts the pressure pulses measured at the strain gages to pressure pulses across multiple locations on the steam dryer. It does not need to model any of the sources that create the pressure pulsations at the strain gage. The SRV standpipes are located downstream of the region modeled and the effect of the standpipes are present in the pressures detected by the strain gages.

LR-N07-0099 LCR H05-01, Rev. 1 SRV standpipe modeling is critical for scale model tests (SMT) since the SMT is used to predict the input to the ACM. Considerable effort was expended in the SMT SRV standpipe modeling to maximize the accuracy. This included taking measurements on a SRV under refurbishment to supplement the vendor drawings.

((

I]

14.9 Figures A.1 through A.26, of the Hope Creek EPU submittal Attachment 20 R]

Response

R[

LR-N07-0099 LCR H05-01, Rev. 1 1]

14.10 ((

Response

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14.11 ((

1]

Response

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11 LR-N07-0099 LCR H05-01, Rev. 1 14,12 Question Superceded by new Steam Dryer Data (February 28, 2007 Supplement) 14.13 ((

1]

Response

((

)) After rebenchmarking, the one-eighth SMT predicts the

-.10-LR-N07-0099 LCR H05-01, Rev. 1 onset of SRV acoustic at or about CLTP. This is consistent with plant data, which does not detect SRV acoustic resonance at CLTP ((

I]

14.14 In reference to the geometrical similarity between the SMT and the Hope Creek plant, in attachment 22 to the Hope Creek EPU submittal, clarify why the measured resonance frequency of SRV standpipes differs substantially from the estimated value.

Response

The initial estimate was made of the diameter of the SRV using an unscaled vendor drawing. This geometry gave ((

))

As part of the 1/8th scale model testing to predict the load, a more thorough review was made to maximize the details. This included taking measurements and photographs of a SRV under refurbishment to obtain more information on the diameter of the SRV. The result was a more complex shape for the chamber which was replicated in the 1 /8th SMT. The shape is as shown in Figure A-1 of CDI report 06-16. ((

))

14.15 Substantiate the use of the load increase factors cited in Page 30 (((185% in PRMS and 156% in differential pressure) instead of using the maximum values given in Table 8.2 of attachment 22 to the Hope Creek EPU submittal.))

Response

((

)) Using LR-N07-0099 LCR H05-01, Rev. 1 a SMT to provide a prediction of a load not present at CLTP reduces the uncertainty during the power ascension; however, it is not intended to replace the verification performed by the power ascension test plan (refer to the response to RAI 14.33).

14.16 In reference to attachment 22 of the Hope Creek EPU submittal, please discuss the effects of the Normalized Maximum Differential pressure on the dryer for the low and high frequency excitations separately. The objective is to produce two figures similar to Fig. 8.8, whereby one figure is for the low and the other for the high frequency excitations.

Response

LR-N07-0099 LCR H05-01, Rev. 1 11 14.17 In reference to attachment 22 of the Hope Creek EPU submittal, ((

11

Response

LR-N07-0099 LCR H05-01, Rev. 1

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14.18 In reference to attachment 22 of the Hope Creek EPU submittal, ((

I].

Response

((

LR-N07-0099 LCR H05-01, Rev. 1 LR-N07-0099 LCR H05-01, Rev. 1

))

14.19 In reference to Attachment 22 of the Hope Creek EPU submittal, demonstrate that the safety relief valves will operate properly at EPU conditions. PSEG's response to this request should include the assessment of (a) the acoustic -

pressure in the standpipes at EPU conditions, and (b) the effect of acoustic pressure at EPU on the structural integrity of the SRVs.

Response

HCGS reviewed the acoustic resonance concern and its potential impact on valve operability as a result of the failures at Quad Cities. Quad Cities had several relief valve designs including several Target Rock valves that were subject to very high acoustic resonance in the MSLs prior to the mitigation efforts taken in early 2006. The valves that failed were not the Target Rock valves. The failure on a non-Target Rock valve was due to excess wear on valve components that was directly attributable to the high vibration.

The photo is taken from the bottom of a Target Rock 7567F looking upward. The tube inside is the sensing line that leads to the seat of the valve's pilot operator.

The other valve components exposed to the valve chamber are major components that are not considered susceptible to the pressure pulsations from acoustic resonance.

As illustrated in the sketch, the sensing line opening is just above the interface between the upper and lower halves of the flanges. The opening is approximately halfway up the overall chamber. The smaller ID of the lower flange half is intended to shield the sensing lines from flow induced phenomena when LR-N07-0099 LCR H05-01, Rev. 1 the valve disc is opened.

(a) ((

)) This is well below the margin maintained between the lowest setpoint and the operating pressure. The HCGS SRVs maintain a nominal margin of > 108 psid between operating pressure and setpoint. The 14 SRVs fall into three setpoint groupings -four at 1108 psig, five at 1120 psig, and five at 1130 psig. During power operation with a steam dome pressure of 1005 psig, pressure drops in the MSL result in the pressure at the SRV flange dropping below 1000 psig. [HNT1]

(b) As discussed in attachment 8, HCGS installed accelerometers on the two representative MSLs and took measurements in 2005 to quantify the vibration levels at CLTP and to demonstrate that there is sufficient margin for increases due to operation at 115% CLTP. Also, as discussed in attachment 8, HCGS will add accelerometers on four of the SRVs to specifically monitor the SRVs during the power ascension. The Power Ascension Test Plan (PATP) specifies power ascension test plateaus (every 5%) which allow verification that the measured values are acceptable. Should measured values exceed the criteria, the PATP requires a reduction in power for evaluation.

Attachment 2 LR-N07-0099 LCR H05-01, Rev. 1 Sensing line 6.0-Inch ID 5.18-inch ID 6.82-inch ID Not to scale.

Sweepolet. Sweepolet curvature details not shown.

LR-N07-0099 LCR H05-01, Rev. 1 14.20 Question Superceded by new Steam Dryer Data (February 28, 2007 Supplement).

14.21 Question Superceded by new Steam Dryer Data (February 28, 2007 Supplement) 14.22 Question Superceded by new Steam Dryer Data (February 28, 2007 Supplement) 14.23 Question Superceded by new Steam Dryer Data (February 28, 2007 Supplement) 14.24 Question Superceded by new Steam Dryer Data (February 28, 2007 Supplement) 14.25 In the Hope Creek EPU submittal attachment 18 (CDI Report 06-17, Rev.2 "Hydrodynamic Loads on Hope Creek Unit 1 Steam Dryer to 200 Hz", September 2006), the results included in the report (e.g., Figures. 3.1 & 3.6) appear to contradict CDI's conclusion that the dryer load determined from SMT is conservative. PSEG is requested to address this apparent contradiction.

Response

((

LR-N07-0099 LCR H05-01, Rev. 1 Hope Creek CLTP I

1 ' I' ' ' ' ' ' '

0.8 A Upper: Subs N A Upper: 200 --------- -- -

0.6 -- - - -- - - -

0.4 - - - - - - - - --

0.2 0

0 50 100 150 200 Frequency (Hz) 1]

14.26 In reference to the spectral peak near 135 Hz which appears at nodes 7 and 99 of Figure 4.6 in attachment 18 of the Hope Creek EPU submittal, please specify the source generating this 135 Hz peak. If it is the standpipe resonance frequency, then PSEG should explain why it is different from the frequency measured in SMT. PSEG should also explain why this frequency peak is not observed in the strain gage signals.

Response

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14.27 Since the field experience shows that strain gages do fail during operation, please explain whether PSEG has any plans to provide adequate redundancy so that sufficient amount of strain gage measurements can be made, despite failure of some strain gages, for reliable prediction of stresses in the steam dryer.

Response

- 20 -

LR-N07-0099 LCR H05-01, Rev. 1 PSEG obtained considerable input from Exelon and SIA in the selection and installation of strain gages. The failed strain gages were suspected of a common failure mode. In January 2007, PSEG performed a maintenance outage to restore, among other work, the 8 failed strain gages and confirmed that the failure was caused by a common mode failure. The splices between the strain gages and the containment cable were located by hot piping, causing the failure of the splices. The repairs were made and all channels provided readings on start-up.

Nevertheless, for prudence, PSEG is currently planning to install additional strain gages to provide redundancy and/or improve accuracy. The second set (of 32 strain gages) would be installed at 45' rotation from the existing strain gages.

14.28 In the Hope Creek EPU submittal attachment 19 (CDI Report No. 06-24, "Stress Analysis of the Hope Creek Unit 1 Steam Dryer for CLTP, Rev. 3," September 2006), the prediction of fluctuating pressure loads using a separate acoustic circuit analysis is discussed. The licensee asserts that the 80-Hz signal is not present in the MSL strain gage signals at CLTP but it is fictitiously introduced by ACM. The licensee is requested to explain whether ACM could have fictitiously modified the other peaks that are present in the predicted fluctuating loads. The licensee is also requested to explain the ACM mechanisms responsible for introducing the 80-Hz signal and ((

I))

Response

See responses to RAI 14.18 for an explanation ((

))

14.29 In the Hope Creek EPU submittal Attachment 19, the licensee describes the model simplifications made in developing the finite element model of the Hope Creek steam dryer. The report also discusses modeling of certain specific components such as perforated plates and the vane bank model. Then the details of the finite element mesh and the types of elements are described.

Please describe how the finite element model developed preserves the dynamic properties (mode shapes and natural frequencies) of the actual Hope Creek steam dryer.

LR-N07-0099 LCR H05-01, Rev. 1

Response

There are no HC1 steam dryer mode shapes or frequencies to compare against.

The HC1 steam dryer has been exposed to high radiation levels for over 20 years of power operations. Due to ALARA considerations, the steam dryer must be kept submerged. Inspections are done with submergible, remote operated cameras. It would be dose prohibitive to perform hammer tests to measure the dynamic properties of the HC1 dryer. Although the steam dryers as delivered to the site were confirmed as identical, the Unit 1 dryer was modified on-site whereas the abandoned Unit 2 dryer was never modified. Thus the mode shape information can not be used since the modifications alter mode shapes and frequencies. While many of the modes will remain virtually unchanged between the dryers, it is not obvious which ones are altered by the modifications and which ones are not. The HC1 steam dryer finite element model was prepared using a standard industry program, ANSYS, which is believed to be the same as used by other utilities for their steam dryer modeling. Finite element models are routinely used in the industry and standard practices are followed in the formulation of the model (e.g., 1% damping, etc.). ((

1]

14.30 The alternating stresses at the welds calculated by the finite element analysis are multiplied by the weld factor to determine the actual fluctuating stresses at the welds. In Hope Creek, steam dryer plates of different thicknesses are welded together. However, the weld factor accounts for the stress concentration introduced by a weld between components of equal thickness. Explain why the stresses at the welds are not modified to include size effects, which account for the welding of the components having different thicknesses.

Response

((

LR-N07-0099 LCR H05-01, Rev. 1

))

14.31 In the Hope Creek EPU submittal Attachment 21 (CDI Report No. 06-27, Rev. 0, "Stress Analysis of the Hope Creek Unit 1 Steam Dryer at CLTP and EPU Conditions Using 1/8th Scale Model Pressure Measurement Data, September 2006), it is questionable that a 50-step time history analysis can identify the frequency shift for the fluctuating loads that would impose the highest loads on the dryer under EPU conditions. For example, the lower Figure 6b page 11 (one without the 80 Hz signal) shows that for 0% frequency shift the maximum fluctuating load occurs at about 700 steps (1 second) into the time history analysis. The NRC staff requests the licensee to explain whether any frequency shift other than + 10% would impose higher loads and generate higher stresses in the steam dryer under EPU conditions for a longer time duration (i.e., 2 or 3 seconds).

Response

LR-N07-0099 LCR H05-01, Rev. 1 In the revised analysis for EPU conditions, CDI Report 06-27 (Rev 2.) submitted to the NRC on February 28, 2007 by LR-N07-0034, a complete 2 second run was performed for every frequency shift. Table 7b identifies the limiting component and identifies the frequency where the limit is reached for that component. The most limiting component is again at the +10% frequency shift.

Also, see response to RAI 14.2 for added information.

14.32 There is an assumed damping ratio of 1% with anchor points at 10 and 150 Hz for the transient pressure loading without any frequency shift described in section 3.5 of attachment 21 page 13. Explain whether the anchor points were changed when considering + 10% frequency shift in the loading.

Response

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14.33 In Section 5, attachment 21 of the Hope Creek EPU submittal, the licensee reports that the minimum stress ratios, which correspond to alternating stresses, are less than 1.0 when +10% frequency shifts are considered. These stress LR-N07-0099 LCR H05-01, Rev. 1 ratios are calculated without accounting for +27.2% uncertainty, which is reported in Attachment 7 of the Hope Creek EPU submittal. The licensee is requested to explain the effect of this uncertainty on the minimum stress ratios and identify the corresponding steam dryer components having stress ratio less than 1.0. Also, the licensee is requested to explain whether it plans to make any structural modifications to these components.

Response

CDI Report 06-27 RO was first supplemented by Tech Evaluation 80090626 and was then superseded by Report 06-27 Rev 2. After preparing Tech Evaluation 80090626, PSEG undertook the margin recovery effort, documented in CDI Report 06-27 Revision 2, which was submitted to the NRC on February 28, 2007. Revision 2 shows that all stress ratios are well above 1.0. The response to RAI 14.5 addresses how the load magnitude uncertainty is considered. All alternating stress ratios at the nominal frequency are > 1.96, which provides significant margin for uncertainty. All frequency variations (-10%, -7.5%, -5%, -

2.5%, + 2.5%, +5%, +7.5% and +10%), with one exception are > 1.62, which still provides significant margin for the load magnitude uncertainty.

The one exception is at the +10% frequency variation. The alternating stress ratio drops to 1.33 for the middle hood to end plate reinforcement strip.

When considering both the frequency uncertainty and the load magnitude uncertainty, the two can be combined by the square root of the sum of the squares (SRSS) since the uncertainties are independent of each other.

It is evident from CDI Report 06-27 R2, Figure 27 that the frequency shift for this node results in a substantial stress increase. At the plus10% frequency shift, the stress is 2.34 times the stress at nominal frequency, or 134% higher than at the nominal frequency. This equates to a frequency uncertainty of 134%. When the load uncertainty (27.2%) is combined by the SRSS with this frequency uncertainty, the combined uncertainty is only 2% larger than the frequency uncertainty. Thus, the stress ratio at this node for the 10% frequency shift remains above 1.30.

PSEG is not proposing or planning any steam dryer modifications. PSEG will perform a power ascension test plan which includes holding at specific power plateaus to verify that the actual, measured loads will not result in alternating stresses that exceed the fatigue allowable. The PATP is outlined in attachment 23 of the EPU LCR submittal.

14.34 Several dryer resonances are apparent in the stress power spectral densities (PSDs) shown in Figures 23-26, Attachment 21 of the Hope Creek EPU submittal. Please provide images and frequencies of the dryer resonances that respond significantly to the dryer loading.

LR-N07-0099 LCR H05-01, Rev. 1

Response

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LR-N07-0099 LCR H05-01, Rev. 1 LR-N07-0099 LCR H05-01, Rev. 1 LR-N07-0099 LCR H05-01, Rev. 1 14.35 The stresses based on Hope Creek in-plant MSL data show strong low frequency peaks (Figure 6.1 Attachment 19 of the Hope Creek EPU submittal),

whereas the stresses based on SMT data are dominated by the SRV singing near 120 Hz (Figures 23-26 Attachment 21 of the Hope Creek EPU submittal).

Please provide graphs which compare directly the dryer stress PSDs computed using in-plant Hope Creek data and SMT data at CLTP at the locations of maximum stress. These locations are given in Table 5.1, Attachment 19, and the nodes listed on page 78 of Attachment 21.

Response

((

LR-N07-0099 LCR H05-01, Rev. 1

-l i 4 +

LR-N07-0099 LCR H05-01, Rev. 1

-I 4 -4 4 -1 4

- 31 LR-N07-0099 LCR H05-01, Rev. 1 LR-N07-0099 LCR H05-01, Rev. 1 LR-N07-0099 LCR H05-01, Rev. 1 LR-N07-0099 LCR H05-01, Rev. 1 LR-N07-0099 LCR H05-01, Rev. 1 LR-N07-0099 LCR H05-01, Rev. 1 LR-N07-0099 LCR H05-01, Rev. 1 LR-N07-0099 LCR H05-01, Rev. 1 LR-N07-0099 LCR H05-01, Rev. 1 14.36 Question superceded by the new Steam Dryer Data (February 28, 2007 Supplement).

14.37 Attachment 8 (Flow Induced Vibration) of the Hope Creek EPU submittal, describes an assessment of flow induced vibration of systems and components in support of the EPU amendment request for Hope Creek. You indicate that some analyses remain underway for vibration susceptibility. Please provide the results or the progress of those analyses and discuss any resulting modifications or procedure changes.

Response

The following evaluations were completed following the transmittal of Attachment 8.

MS small bore piping connections: PSEG stated that small bore piping in the vicinity of the Turbine Stop Valve (TSV) and Turbine Control Valve (TCV) would receive further evaluation or alternatively be modified to improve vibration resistance. Twenty-four (24) MS small bore connections in the vicinity of the TSV and TCVs, consisting primarily of vents and drains, will be modified to improve vibration resistance by strengthening sockolet welds, as recommended in EPRI document TR 107455. This is being done during the EPU implementation outage, fall 2007.

EHC connections to TSV and TCV: PSEG performed a more detailed evaluation of the EHC connections, which based on industry data has been problematic at some plants even without power uprate. The evaluation concluded that with the exception of leakage from O-rings, there has been no tubing leakage/failures at HCGS. The root cause of O-rings leaks was a problem with identification of the proper O-ring as opposed to vibration.

In addition, the HCGS design initially or subsequent to commercial operation implemented all the GE suggested recommendations to reduce the overall vibration problems including but not limited to addition of accumulators, relocation of the EHC pumps off the skid, noise reduction (to eliminate internal pressure pulsations). Test Data on vibration tests done in 1989 as part of the conversion from full arc to partial arc high pressure turbine steam admission showed that the baseline, full arc vibration levels on EHC piping attached to the TSV and TCV were < 30% of allowable. With partial arc implementation, these levels were reduced to almost half of that value. PSEG has accordingly elected not to add flexible hoses on the tubing connections to the TCV and TSV.

All thermowells and sample probes evaluations have been completed. None require modification. See RAI 14.48.

-40 -

LR-N07-0099 LCR H05-01, Rev. 1 No procedure changes are required.

14.38 Attachment 23 (Power Ascension Test Plan Overview) of the Hope Creek EPU submittal states that a detailed Power Ascension Test Plan will be provided to the NRC Staff before increasing power above CLTP. Please provide the draft or completed test plan along with the limit curves for power ascension, including the margin available from the fatigue stress limit if the curve is reached during power ascension.

Response

The power ascension Test Plan (PATP) is attached. The steam dryer limit curves are being developed. PSEG expects to provide the curves to the NRC on or before May 10, 2007.

14.39 In reference to Attachment 8 of the Hope Creek EPU submittal, provide (a) applicable types of accelerometers/instruments for each of Vibration Monitoring Groups, and (b) locations of instruments and technical basis for selection of the monitoring locations.

Response

PSEG monitoring consists of Vibration Monitoring Groups (VMG) 1 and VMG 3 as defined in ASME Operation and Maintenance (OM) Standards and Guides (S/G) Part 3, "Requirements for Preoperational and Initial Start-up Vibration Testing of Nuclear Power Plant Piping Systems".

VMG 1:

The locations for the MS and FW accelerometers are provided in Table 2 of attachment 8 and are shown in "RAI 14.39 Figures 1 thru 7". The accelerometer selected is ENDEVCO Model 7703A-100.

As described in attachment 8, the instrumentation for VMG 1 consists of accelerometers. The accelerometer locations on this piping were selected based on detailed analyses of the piping, which identifies the resonance frequencies for the piping, correlates vibration levels to flow induced vibration (FIV) stress levels, and establishes the maximum allowed vibration (i.e., acceptance criteria). To determine the accelerometer locations, models of the piping were created, using PIPESTRESS which represents the dynamic characteristics of each monitored line. To provide the maximum number of natural frequencies with the minimum number of instruments, modal analyses of the piping systems were performed.

PSEG will add accelerometers on four (4) Safety Relief Valves (SRVs) (two SRVs on main steam lines "A" and 2 SRVs on MSL "B") to monitor acceptable vibration levels during power ascension.

LR-N07-0099 LCR H05-01, Rev. 1 VMG 3:

As part of the Power Ascension Test Plan (PATP), PSEG is selecting locations to be visually monitored. The selection considers ALARA/accessibility during power operation.

As stated in attachment 8, the condensate and reactor feed pumps will operate at an improved efficiency point and are not expected to increase in vibration.

Thus, no additional instrumented monitoring is planned for these pumps as vibration monitoring already exists on the pumps. In addition, the condensate pump rooms are accessible during power operation. The walkdowns will include the condensate pump rooms.

LR-N07-0099 LCR H05-01, Rev. 1 M'S-D 3.6. AUG 21 0 1: (WIGDC-WS)P MS. A -DP 14 EL 154' X, z

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i .fcr .*.3t C-1 /3 /  : Q 04ý07 RAI 14.39 Figure 1. Main Steam Line A Accelerometer Locations - Drywell

- 43 -

LR-N07-0099 LCR H05-01, Rev. 1 PSG--D 3. 0 AUG 2CO1 (KIND:)WS)P "S" A S" " J " -. ". / "

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A13, 06: 3411 R04.063 RAI 14.39 Figure 2. .Main Steam Line A SRVDL J Accelerometer Locations - Drywell

-44 -

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RAI 14.39 Figure 3. Main Steam Line B Accelerometer Locations - Drywell LR-N07-0099 LCR H05-01, Rev. 1

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RAI 14.39 Figure 4. Main Steam Line B SRVDL P Accelerometer Locations - Drywell

-46 -

LR-N07-0099 LCR H05-01, Rev. 1 RAI 1c .: :

RAI 14.39 Figure 5. Feedwater Piping Accelerometer Locations - Dryweli LR-N07-0099 LCR H05-01, Rev. 1 A2G. 2021.

2*G.D20 [W0DCW0S)P 8i from support

@260, AB-001-H07-DR Z018 X,Y,Z - - ------ 0 5 10.5' from support 7AB-001-H07 DR Z0 .

@236, AB-001-H14 DR Z013.":

x,67, 11 from support

-X, Y . A @60AB-002-H13 A02-H13 DR Z003 x, y Figure A: STMTDARD VIEW I 200.1k> 2221... IUNOONIAIR 4 \.eIL3' es/rs rol*. 2IC--.4QoM*OC'w\ CP0*A. -sO 03/05/04 1i*:*

04.13 RAI 14.39 Figure 6. Main Steam Piping Accelerometer Locations - Turbine Building LR-N07-0099 LCR H05-01, Rev. 1

ý,eý r~v~ A ~ V1~w RAI 14.39 Figure 7. Feedwater Piping Accelerometer Locations - Steam Tunnel LR-N07-0099 LCR H05-01, Rev. 1 14.40 In Attachment 8 of the Hope Creek EPU submittal, it is indicated that the Hope Creek acceptance criteria for vibration level of the main steam, feedwater, and other piping systems are based on the guidance of ASME OM S/G Part 3 Code where it requires that the calculated stresses due to steady state vibration shall not exceed the allowable stress limit as specified by the Code. However, Tables 3 to 5 show that the acceptance criteria are based on the root-mean-square (RMS) of acceleration for each piping. The licensee is requested to demonstrate how the piping vibration level will be within the OM Code stress limit using the RMS acceleration value in place of the peak response spectrum used in the classical piping analysis.

Response

The acceptance criteria for the main steam, feedwater, and other piping systems were developed by performing a dynamic response spectrum piping analysis for each of the piping systems. The acceptance criteria for steady state operation were generated prior to collecting actual plant data so the frequency content of the piping systems was not known. The frequency content due to steady state vibration is typically broad band so the acceptance criteria was developed using an equivalent RMS input that would contain energy over the entire frequency range of interest. This differs from the "classical piping analysis" such as seismic analysis where the magnitude of the input amplified response spectra (ARS) varies with frequency. The calculated stresses used to develop the acceptance criteria are conservative since every frequency is assumed to have the same input magnitude. Comparison of the acceptance criteria to the measured acceleration RMS values is consistent with the broad band input used to develop the criteria. Details of the method used to develop the acceptance criteria are provided in the following paragraphs.

Broad band ARS of 1 g were inputted in each of the three orthogonal directions.

The displacements, accelerations, and stresses due to the broad band ARS in each direction were calculated at each node of the piping system. The total response was obtained by combining the results from each of the three directions by the square root of the sum of the square (SRSS) method. This dynamic analysis provides the response of the piping system to a broad band ARS.

The steady state vibration is flow induced so the actual vibration loading is only applicable to the sections of piping normally containing flow. Since the flow is mostly parallel, the flow induced vibration provides excitation along the direction of the flow, not in all directions at each location. The stresses calculated due to the broad band ARS applied in all three orthogonal directions would result in a higher calculated stress than if only applied axially to each segment of pipe containing flow. This method yields lower allowable accelerations at each LR-N07-0099 LCR H05-01, Rev. 1 location than if only an axial load were applied to each non-stagnant section of pipe. This method produces conservative acceptance criteria.

The acceptance criteria for the measured accelerations were determined by multiplying the calculated acceleration at each sensor location in a unit load analysis by the ratio of the allowable steady state stress to the maximum calculated stress in the piping system. This ensures that the maximum steady state stress for each piping system does not exceed the ASME Operation and Maintenance (OM) Standards and Guides (S/G) Part 3.

14.41 [The response to question 14.41 will be submitted on or before 5/10/2007]

14.42 In the PUSAR Section 3.5.2, "Balance of plant piping," page 3-23, "Pipe Stresses" states that "No new postulated pipe break locations were identified."

Provide a justification for this statement. Confirm whether and how the determination of line break locations is based on SRP Section 3.6.2, MEB 3-1 criteria.

Response

Pipe Break criteria were evaluated based on the requirements of Section 3.6.2 of the UFSAR, which is based on B.1.a and B.1.c of NRC Branch Technical Position MEB 3-1 (SRP 3.6.2). Where required, percentage temperature increases were applied to the existing stress levels for the applicable piping system at all node points due to CPPU conditions. The combinations of stresses were evaluated to meet the requirement of pipe break criteria. Based on these criteria, no new postulated pipe break locations were identified.

14.43 In the PUSAR Section 10.1.2, Liquid Line Breaks, states that "Only the mass and energy releases for HELBs in the Reactor Water Cleanup (RWCU) and FW systems may be affected by CPPU and were re-evaluated at CPPU conditions."

Provide summaries of the evaluations for RWCU and FW line breaks.

Response

The RWCU and FW line break blowdown mass flow rate profiles at EPU conditions are determined by calculating the ratio of critical mass flux at CLTP and EPU conditions, then applying this ratio to the CLTP profile from the analysis of record. In the evaluation the Moody slip critical flow is used to determine blowdown mass flux values.

The RWCU and FW line break evaluations address operation at the following points on the Hope Creek EPU power flow map.

LR-N07-0099 LCR H05-01, Rev. 1 EPU Standard IUZL/o -ru I nefrlia[ ruwer kr-ru I r) I Iu1 Normal FW Temperature 102% EPUTP / 100% Core Flow / Reduced FW EPU Standard w/FWTRTmprte Temperature 102% EPUTP / 105% Core Flow / Normal FW EPU ICF Temperature EPU ICF w/FWTRTeprte 102% EPUTP / 105% Core Flow / Reduced FW Temperature EPU minimum pump speed 66.2% EPUTP / 39.2% Core Flow / Normal FW (mps) Temperature 66.2% EPUTP / 39.2% Core Flow / Reduced FW EPU mps w/FWTRTemperature The evaluations do not specifically model the EPU / MELLLA condition (102%

EPUTP, 99% Flow). The evaluations performed for the EPU Standard conditions 102% EPUTP / 100% Core Flow / Normal FW Temperature and 102% EPUTP /

100% Core Flow / Reduced FW Temperature) are representative of the normal feedwater temperature and reduced feedwater EPU / MELLLA conditions.

The results of the RWCU line break evaluation are summarized in the Tables 14.43-1 and 14.43-2.

Table 14.43-1 CLTP to EPU Power Flow Condition Enthalpy Change (1)

(Btu/Ibm)

EPU Standard -1.1 EPU Standard w/FWTR -4.7 EPU ICF 0.0 EPU ICF w/FWTR -3.6 EPU mps -28.8 EPU mps w/FWTR -32.6 (1) As compared to the current license basis value

- 52 -

LR-N07-0099 LCR H05-01, Rev. 1 Table 14.43-2 RWCU Pump RWCU HX RWCU F/D RWCU Pump Discharge Inlet Mass Inlet Mass Suction Power Flow Mass Flux Flux Flux Mass Flux

% increase  % increase  % increase  % increase (1) (1) (1) (1)

EPU Standard 1.0% 0.6% 0.1% 0.6%

EPU Standard 4.4% 2.3% 0.6% 2.6%

w/FWTR EPU ICF 0.0% 0.0% 0.0% 0.0%

EPU ICF w/FWTR 3.4% 1.8% 0.4% 2.0%

EPU mps 26.4% 12.5% 3.3% 30.4%

EPU mps w/FWTR 29.5% 13.9% 3.7% 35.2%

(1) As compared to the current license basis value Response for RWCU line break Environmental Conditions The licensing basis RWCU line breaks outside containment were evaluated during original plant design using the COPDA computer code. Pressure and temperature profiles for the various postulated RWCU breaks were evaluated.

The design basis analysis was based on a 17 node reactor building model.

A benchmarking analysis was performed using the COMPARE Mod 1 computer code using the same 17 node model as performed for the design basis model.

The benchmarking analysis was performed to evaluate the pressure and temperature response for the ARTS/MELLLA revised mass and energy releases.

The results of the benchmarking study demonstrated that the results using the COMPARE computer code closely follow the results using COPDA. Therefore the COMPARE models developed using the original design basis COPDA physical input parameters were used for both the ARTS/MELLLA mass and energy releases and the EPU mass and enthalpy changes described above.

In addition to the benchmarking cases six different RWCU rooms were evaluated. The cases are;

1. Break in Room 4403 - RWCU discharge downstream of check valve

2. Break in Room 4506 - downstream of Regen Hx Outlet

3. 6" RWCU break upstream of pump in Room 4402

4. 6" RWCU upstream of pump in Room 4505

5. RWCU break in Room 4621 F/D Inlet

6. RWCU break in Room 4503 - RWCU Valve and Pump Room Each of these cases were evaluated at the different power flow conditions defined above:

- 53 -

LR-N07-0099 LCR H05-01, Rev. 1 The peak compartment pressure and temperature following the design basis breaks in the RWCU piping inside the Reactor Building were determined using the COMPARE computer code. The mass and energy releases were modified to incorporate changed RWCU initial conditions for the different flow conditions defined above. The COMPARE results were either the same or less than the design basis pressure and temperature results.

In all cases evaluated the COMPARE results at EPU flow conditions are enveloped by the licensing basis pressure and temperature results (See Table 14.43-3).

For the FW line break case, the energy release (mass flux

• enthalpy) results at the power flow conditions identified above are compared to the energy releases from a MSLB in the MST at current licensed conditions. The comparison demonstrates that the FWLB energy releases at CPPU conditions are bounded by energy releases from a MSLB in the MST at current licensed conditions.

Table 14.43-3 Summary Results RWCU Line Breaks Case Break Room Pressure Temperature

1. (psig) (° Licensin EPU Licensin EPU g Basis g Basis 1 Pump dischg. 4403 1.7 1.7 217 217 1 Pump dischg. 4405 1.7 1.7 217 217 2 Regen Hx 4506 2.1 1.7 217 217 Outlet 3 6" Pump suction 4402 1.7 1.7 218 206 3 6" Pump suction 4321 . .1.6 1.6 217 206 4 6" Pump suction 4505 1.8 1.8 218 214 5 F/D (3" line) 4620 6.4 3.8 231 224 5 F/D (3" line) 4621 6.4 3.8 231 224 6 F/D (4" line) 4502 1.9 1.9 218 218 6 FID (4" line) 4503 1.9 1.9 218 218 LR-N07-0099 LCR H05-01, Rev. 1 14.44 In the PUSAR Section 3.5.1, "Reactor Coolant Pressure Boundary Piping," page 3-19 "Feedwater Evaluation," states that the "CPPU does not have an adverse effect on the FW piping design. A review of postulated pipe break criteria concluded that at three locations, cumulative fatigue usage exceeds postulated pipe break criteria limit. (question was revised following conference call) a) If the CPPU does not have an adverse effect on the FW piping, why is it necessary to perform structural modifications to ensure ASME code compliance prior to the implementation of the CPPU? Please clarify.

Response

The initial conclusions in PUSAR are based on a conservative GE screening analysis. The GE screening analysis reported that three locations may have a cumulative usage factor (CUF) greater than 0.1 at CPPU conditions, exceeding HCGS pipe break design criteria. PSEG subsequently completed reanalysis of Feedwater (FW) piping is at CPPU conditions as described below in 14.44(b).

The piping meets the ASME code requirements of HCGS (ASME B&PV Code, Div. 1, Section I11,1977 Edition through Summer 1979 Addenda) without any structural modification.

b) Provide the summary of the piping reanalysis results and the schedule that the structural modifications will be completed where required to ensure that ASME Code stresses and fatigue usage factors will not exceed the criteria limit, prior to implementation of CPPU. If structural modifications are required, confirm that the structural analysis model (piping and/or pipe support) reflects any required structural modifications.

Response

Two of the three locations referenced in PUSAR Section 3.5.1 are at FW Containment Penetration Nozzles. The flued head penetrations were originally qualified by detailed finite element analysis performed by Basic Technology, Inc.

in 1986. The ASME piping code of record was used to calculate the CUF based on the original loading specification and compared with the analyzed value at CPPU conditions. This comparison shows that the analyzed value of CUF at CPPU conditions is less than the CUF calculated on the basis of original loads, thereby meeting the HCGS pipe break design criteria.

The third location corresponds to FW Loop A data point 45 in the HCGS FW piping model. The FW piping reanalysis shows CUF less than 0.1 for this location, thereby meeting the HCGS pipe break design criteria.

Attachment 2 LR-N07-0099 LCR H05-01, Rev. 1 The following table lists the highest stress and cumulative usage factor vs.

allowables. The piping meets all Hope Creek ASME criteria without any structural modification.

LINE AE-035 Data Equation(9) Equation(9) Equation(9) Equation(9) EQ.12 EQ.13 Fatigue Points Level A Level B Level C Level D Usage (KSI) (KSI) (KSI) (KSI) (KSI (KSI Factor All 21.028 19.138 19.194 28.656 22.857 47.306 .599 Allowables 29.59 35.51 44.38 59.18 59.18 59.18 1.0 LINE AE-036 Data Equation(9) Equation(9) Equation(9) Equation(9) EQ.12 EQ.13 Fatigue Points Level A Level B Level C Level D Usage (KSI)

AlI KSI) (KSI) (KSI (KSI Factor All 1 21.331 27.382 20.635 28.742 32.728 45.463 .573 Allowables 29.59 35.51 44.38 59.18 59.18 59.18 1.0 14.45 In the PUSAR Section 10.1.2, "Liquid Line Breaks," page 10-3 "pipe whip and jet impingement," it is indicated that the FW piping was evaluated for temperature increases associated with CPPU conditions but it did not result in pipe stress levels above the thresholds required for postulating HELBs, except at locations already evaluated for breaks. Provide a summary of this evaluation for the NRC staff's review.

Response

UFSAR section 3.6.2.1.1 "Break Location in High Energy Fluid System Piping" states that high energy piping system line breaks and effects have been evaluated in accordance with sections B.1 .a and B.1 .c of BTP (Branch Technical Position) MEB (Mechanical Engineering Branch) SRP (Standard Review Plan) 3.6.2. For the Feedwater Piping Inside Containment SRP 3.6.2 is consistent with the piping stress requirements described in UFSAR 3.6.2.1.1.1.

The two feedwater piping trains inside containment are identified as AE035 and AE036. Pipe-break stress values at locations where the stress (EQ. 10) or cumulative fatigue usage factor (0.10) exceeds these criteria for the AE035 and AE036 are tabulated in the two tables below.

In all instances where Equation 10 exceeds 2.4 Sm, equations 12 and 13 are less than 2.4 Sm. At all locations where the cumulative usage factor exceeds 0.1, a pipe break has already been postulated. Therefore, based upon the reanalysis of the feedwater piping inside containment for EPU conditions, no new pipe break locations need to be postulated.

- 56 -

LR-N07-0099 LCR H05-01, Rev. 1 LINE AE-035 Node New New New Pipe New Pipe Break Basis for Node Type Stress Stress Stress Break Cumulative Cumulative Break Point EQ. 10 EQ. 12 EQ. 13 Stress Usage Usage Selection (ksi) (ksi) (ksi) Limit Factor Factor 2.4 Sm Limit (ksi) 15 TTJ 57.42 NR NR 47.34 .1995 0.10 TE 25 TTJ 60.821 NR NR 47.34 .2192 0.10 MIB 30 Bran. 54.939 NR NR 47.34 .2416 0.10

• 60 TTJ 56.566 NR NR 47.34 .1946 0.10 MIB 70 TTJ 56.502 NR NR 47.34 .1938 0.10 MIB 95 TEE 75.819 13.188 42.467 47.34 .2551 0.10 MIB 108 LUG 23.039 NR NR 47.34 .9253 0.10 MIB 130 TEE 65.425 18.392 34.920 47.34 .2141 0.10 MIB 140 TTJ 63.344 14.908 41.996 47.34 .0297 0.10 178 LUG 18.296 NR NR 47.34 .3756 0.10 MIB 195 TTJ 50.366 22.857 16.258 47.34 .0427 0.10 196 TTJ 49.373 22.23 17.187 47.34 .0430 0.10 200 TTJ 62.214 8.443 30.831 47.34 .5988 0.10 TE 205 TTJ 66.23 13.986 42.702 47.34 .1006 0.10 **

225 LUG 19.007 NR NR 47.34 .1673 0.10 MIB 250 LUG 14.941 NR NR 47.34 .3593 0.10 MIB 260 TTJ 48.542 16.889 24.078 47.34 .0359 0.10 261 TTJ 49.624 17.98 23.389 47.34 .0357 0.10 265 TTJ 60.591 NR NR 47.34 .5209 0.10 TE 270 TTJ 62.216 15.065 34.848 47.34 .0964 0.10 288 LUG 22.196 NR NR 47.34 .3799 0.10 MIB 315 TTJ 62.615 6.329 31.520 47.34 .5348 0.10 TE NR NR LR-N07-0099 LCR H05-01, Rev. 1 LINE AE-036 Node New New New Pipe New Pipe Break Basis for Node Type Stress Stress Stress Break Cumulative Cumulative Break Point EQ. 10 EQ. 12 EQ. 13 Stress Usage Usage Selection (ksi) (ksi) (ksi) Limit Factor Factor 2.4 Sm Limit (ksi) 15 TTJ 57.555 NR NR 47.34 .2289 0.10 TE 25 TTJ 61.086 6.952 30.476 47.34 .2539 0.10 MIB 30 BC 55.224 NR NR 47.34 .2754 0.10

• 60 TTJ 62.547 5.260 34.35 47.34 .2651 0.10 MIB 70 TTJ 56.313 NR NR 47.34 .2202 0.10 MIB 90 TTJ 36.855 NR NR 47.34 .1017 0.10 **

100 TTJ 37.557 NR NR 47.34 .1033 0.10 **

95 TEE 88.219 17.263 47.654 47.34 .54 0.10 MIB 108 LUG 24.86 NR NR 47.34 .9436 0.10 MIB 130 TEE 58.803 NR NR 47.34 .2369 0.10 MIB 140 TTJ 57.639 14.625 35.396 47.34 .0332 0.10 178 LUG 17.39 NR NR 47.34 .3744 0.10 MIB 180 BC 54.733 NR NR 47.34 .3055 0.10 MIB 195 TTJ 58.29 25.949 20.837 47.34 .054 0.10 200 TTJ 61.754 6.954 30.773 47.34 .5725 0.10 TE 206 TTJ 67.48 14.606 43.867 47.34 .1021 0.10 **

210 TTJ 53.083 25.025 36.351 47.34 .0092 0.10 225 LUG 19.921 NR NR 47.34 .1677 0.10 MIB 250 LUG 15.589 NR NR 47.34 .3602 0.10 MIB 265 TTJ 60.392 NR NR 47.34 .5219 0.10 TE 270 TTJ 57.987 12.888 32.245 47.34 .0861 0.10 275 TTJ 55.106 32.09 26.869 47.34 .0266 0.10 ***

288 LUG 22.504 NR NR 47.34 .3795 0.10 MIB 315 TTJ 61.228 5.328 30.762 47.34 .5313 0.10 TE

• This is next point to DP 25 where break is postulated.

• • These points are on TEE where DP 95 break is postulated.

These points are below the stress and Cumulative Usage Factor criteria and therefore no break is postulated.

NOTES:

TTJ: Tapered Transition Joint TE: Terminal End MIB: Mandatory Intermediate Break NR: Not Required 14.46 [See Attachment 3]

14.47 In the PUSAR Sections 3.2.2 through 3.2.2.3, "Reactor Vessel Structural Evaluation," states that "The effect of CPPU was evaluated to ensure that the LR-N07-0099 LCR H05-01, Rev. 1 reactor vessel components continue to comply with the existing structural requirements of the ASME B & PV Code. For the components under consideration, the 1968 code with addenda to and including winter 1969, which is the code of construction, is used as the governing code. However, if a component's design has been modified, the governing code for that component is the code used in the stress analysis of the modified component. The Hope Creek CPPU utilizes the original code of construction as the governing code for all components for CPPU conditions. New stresses are determined by scaling the "original" stresses based on the CPPU conditions (temperature and flow).

The analyses were performed for the design, the normal and upset, and the emergency and faulted conditions. If there is an increase in annulus pressurization, jet reaction, pipe restraint or fuel lift loads, the changes are considered in the analysis of the components affected for normal, upset, emergency and faulted conditions."

a) Provide a summary of the analyses stated above which justify that loading changes due to CPPU in the analysis of the components affected at the normal, upset, emergency and faulted conditions do not affect the structural integrity of these components.

Response

As defined in the LTR for CPPU, the only components that require evaluation are: (a) those with a pre-existing (OLTP) CUF >0.5 and (b) those that experience an increase in flow, temperature, Reactor Internals Pressure Differences (RIPDs), and other mechanical loads due to CPPU. Therefore, the only components in the HCGS vessel that require evaluation for CPPU operating conditions are the main closure studs, shroud support, feedwater nozzle safe end, and core spray nozzle.

For the main closure studs and the core spray nozzle, there is a small (0.2%)

change in temperature, but no change in pressure or flow. Hence, this small change in temperature has an insignificant effect on primary plus secondary stresses (P+Q) and fatigue Cumulative Usage Factor (CUF).

For the shroud support, there is no increase to the thermal stresses due to CPPU. The dynamic loads and recirculation LOCA loads do not change for CPPU conditions. The only changes are due to an RIPD increase of 11.1%

across the shroud support. This causes an increase in the primary stresses (P) of the same percentage. As a result, (P+Q) increases approximately 3%. This is an insignificant increa-se; it is judged that the CUF remains unchanged.

With regard to the Feedwater Nozzle, the safe end is the limiting component.

The only increases for this component are the temperature and flow rate, which have been assessed. System plus rapid cycling events are included in the LR-N07-0099 LCR H05-01, Rev. 1 CPPU CUF, which is conservatively calculated. No leakage has been identified for this component. As such, the safe end is qualified for 23 years of operation.

Table 14.47-1 provides the Original Licensed Thermal Power (OLTP) CUF compared to the Constant Pressure Power Uprate (CPPU) results.

Similarly, (P+Q) values are presented in the Table 14.47-2.

b) Provide a list of components that have their design modified along with a description of the design modifications, governing code, and maximum stress summary versus allowable stress limits at critical locations.

Response

Table 14.47-3 provides a summary of the applicable Code, stresses, and allowable stresses for all components that have been modified since the original design.

In addition to the component evaluations for modifications summarized in Table 14.47-3, welded overlays on the core spray nozzle to safe end weld (N5B) and the reactor vessel recirculation inlet nozzle to safe end weld (N2K) were satisfactorily evaluated for EPU conditions as described in response to RAI 5.3 (PSEG letter LR-N07-0056 dated March 22, 2007).

c) Provide the procedural tensioning method of Main Closure Studs that includes loads due to CPPU conditions. Show that the tensioning stud load including tolerances is within code allowables when including the higher CPPU conditions.

Response

As provided above in response to RAI 14.47a, the effect of CPPU on main closure stud CUF and (P+Q) is not significant. Therefore, no change to the stud tensioning method or elongation tolerances is required.

Table 14.47-1 Component OLTP CUE CPPU CUE Conclusioni Main Closure Studs 0.755 0.755 Component Acceptable Core Spray Nozzle 0.796 0.796 Component Acceptable ShroqudSupport 0.672 0.672 Component Acceptable Feedwater Nozzle Safe 0.950 0.999 Component Acceptable*

EndII This CUF is based upon 23 years of service. System plus rapid cycling events are included. The calculation for rapid cycling includes very

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Attachment 2 LR-N07-0099 LCR H05-01, Rev. 1 conservative assumptions regarding leakage. This component should be managed by inspection.

Table 14.47-2

~Component OLTF'jP+Q> CPPU P+(Q) Code Conclusion~

(ksi) ,(ksi) :, Allowab~le.

Main Closure 108.9 109.11 110.4 Component Studs Acceptable Core Spray 13.76 13.79 43.11 Component Nozzle Acceptable (See Note 1)

Shroud 22.03 24.41 69.9 Component Support Acceptable (See Note 1)

Note 1:

Thermal Bending has been removed as allowed by the ASME Code.

Table 14.47-3 Component Modification, ICode Modification 'Maximum Allowable

~Maximum Stress at Stress Stress at CPPU (k'§i),

OLTP (ksi)

Recirculation Safe End and ASME Boiler & 26.4 30.2 41.17 Inlet Nozzle Thermal Pressure Vessel Sleeve Code, 1974 Replacement Edition and Addenda up to Summer 1976 _____

ORID Nozzle ASME Boiler & 39.48 39.56 80.1 Hydraulic Capped Pressure Vessel System Code, 1974 Return Nozzle Edition and Addenda up to Winter 1975 Feedwater Safe End ASME Boiler & 18.6 23.46 53.1 Nozzle Safe Replacement Pressure Vessel (Thermal (Thermal End Code, 1974 Bending Bending Edition and Removed) Removed)

Addenda up to

____ ______ ____ ___ ___ Summer_1976 _ _ _ _ _ _ _ _ _ _ _

LR-N07-0099 LCR H05-01, Rev. 1 14.48 Pipe in-line components such as thermowells and sample probes experience increased vibrations due to flow past or through that component. In the PUSAR Section 3.4.1, "FIV Influence on Piping," states that "The safety-related thermowells and sample probes in the MS and FW piping systems have been identified and are being evaluated for the increased main steam and feedwater flow conditions, respectively, as a result of CPPU. Any thermowell or sample probe that is found to be unacceptable will be modified as necessary. Evaluation of non-safety related thermowells and sample probes is described separately."

Attachment 8 further discusses the effects of FIV on thermowells and sample probes. For FW and Condensate, attachment 8 of the Hope Creek EPU submittal states that "Based on a review of existing calculations, the frequency ratio (excitation frequency/natural frequency) at 115% CLTP remains well below the original design criteria of < 0.80. Quantify "well below" in the summary requested below. For FW and Condensate thermowells, Attachment 8 also states that the protrusion length "is 2.5-inches, and it is tapered from 1.5-inches to 1.0 inch. Due to their short protrusion, they are not considered susceptible."

Provide basis and justification of this statement. Provide a quantitative summary of the evaluation that supports the acceptability of the thermowells and sample probes in the MS, FW and related piping systems. Identify nonconforming component(s) and provide description of their modification(s).

Response

None of the thermowells or sample probes require modification for the proposed power uprate (115% CLTP). With a constant pressure power uprate, the primary impact on in-line instrumentation is on systems which will see a significant increase in flow velocities.

All non-safety related FW and condensate sample probes are built from solid bars with a tapered OD from 1.5-inches to 1.0-inches (None are built from standard piping components). Lengths vary in order to obtain representative sampling points in the flow stream. The two most limiting sample probes had lengths of 11.6-inches and 12.9-inches. At 115% CLTP, the non-SR FW and condensate sample probes have a frequency ratio (excitation frequency/natural frequency) < 0.19, less than one-fourth of the screening criteria of 0.80.

All non-safety related FW and condensate thermowells are likewise built from solid bars with a tapered OD from 1.5-inches to 1.0-inches but with shorter protrusion lengths than sample probes. For those thermowells in the larger bore piping, the protrusion length is 2.5-inches. Thus, they are considered even less susceptible to FIV since their frequency ratio (excitation frequency/natural frequency) at 115% CLTP is well below 0.19.

The Safety Related (SR) thermowells and sample probes in the MS and the single SR thermowell in the FW system were analyzed by GE. They are all 62-LR-N07-0099 LCR H05-01, Rev. 1 tapered design, but of different OD's. Their frequency ratio at 115% CLTP is calculated as:

• SR FWthermowell: 0.21.

• SR MSL sample probes: 0.40.

• SR MSL thermowells (TE1004A-D): 0.435.

" SR MSL thermowell (TE040) had a frequency ratio of 1.00. It received a finite element modeling to confirm that the total vibration stresses were 52% of the allowable. The allowable used was 7690 psi.

14.49 In the PUSAR Section 3.4.2, "FIV Influence on Reactor Internal Components,"

states that "The increase in power may increase the level of reactor internals vibration. Analyses were performed to evaluate the effects of flow-induced vibration on the reactor internals at CPPU conditions. This evaluation used a reactor power of 3952 MWt and 105% of rated core flow. This assessment was based on vibration data obtained during startup testing of the prototype plant (Browns Ferry Unit 1). For components requiring an evaluation but not instrumented in the prototype plant, vibration data acquired during the startup testing from similar plants or acquired outside the RPV is used. The expected vibration levels for CPPU region were estimated by extrapolating the vibration data recorded in the prototype plant or similar plants and from GE Nuclear Energy BWR operating experience. These expected vibration levels were then compared with the established vibration acceptance limits. The following components were evaluated:

a) Shroud b) Shroud head and moisture separator c) Jet pumps d) Feedwater sparger e) Steam dryer f) Jet pump sensing lines The results of the vibration evaluation show that continuous operation at a reactor power of up to 3952 MWt and 105% of rated core flow will not result in any detrimental effects on the evaluated reactor internal components (except the steam dryer)." Provide a summary of the evaluation results for each of the above components except the steam dryer and the technical basis that supports the rational that reactor power of up to 3952 MWt and 105% of rated core flow will not result in any detrimental effects on the reactor internal components (based on the evaluated prototype plant).

Response

The acceptance criteria for the shroud, shroud head and separator, jet pumps and feedwater sparger is that the maximum stress anywhere in the component be less than the limiting stress as per GE criteria of 10,000 psi. This limiting LR-N07-0099 LCR H05-01, Rev. 1 value ensures that no fatigue is accumulated by the component, and is lower than the ASME Curve C fatigue limit of 13,600 psi. Table 14.49-1 lists the maximum stress values for each component under both OLTP and 3952 MWt reactor power and 105% of rated flow conditions.

Stress values for all these components are thus under GE acceptance criteria of 10,000 psi. For the Jet Pump Sensing Lines (JPSL) the acceptance criteria is the prevention of JPSL resonance with the Vane Passing Frequency (VPF). This no-resonance condition is met both under OLTP and 3952 MWt reactor power and 105% of rated flow conditions.

The OLTP data is obtained from the startup measurements made at the prototype plant (Browns Ferry 1) at various power levels and core flows. These are documented in Browns Ferry 1 Vibration Measurements, GENE document 386HA219, December 1975. The 3952 MWt and 105% of rated core flow values are obtained by extrapolation of the measured data. Summary of the evaluations is as follows:

Shroud For the shroud, the square of the increase in flow velocity due to power and core flow increase is used to calculate the increase in vibration. Maximum stresses during OLTP are very low and will remain well within acceptance criteria during EPU when the vibrations are increased to account for the flow increase. The vibrations amplitudes are calculated to increase by about 58% during EPU. The calculated maximum stress at 3952 MWt and 105% of rated core flow is less than (( ]

Shroud Head and Separator For the shroud head and separators, the square of the increase in separator velocity due to power and core flow increase is used to calculate the increase in vibration. Due to the increase in power, the separator flow velocity increases.

Using the flow velocity squared relationship, the vibration amplitudes are calculated to increase by about 58%, when operation at 3952 MWt and 105% of rated core flow is considered and result in a maximum stress of under ((

))

Jet Pumps Test point H22 at - 70% rod line and 100 % core flow and test point H24 at

-100% rod line and 100%- core flow from startup data on a prototype plant were selected and used to extrapolate to EPU at 105% core flow. This yielded a predicted maximum response of less than (( ))of the acceptance criteria at 3952 MWt and 105% of rated core flow conditions.

Feedwater Sparger LR-N07-0099 LCR H05-01, Rev. 1 Feedwater spargers are a triple thermal sleeve design. The maximum stresses with the triple thermal sleeve design are less than (( )) the stresses during EPU are calculated by using the square of the increase in FW flow (57%)

and are less than (( ]

Jet Pump Sensing Lines (JPSL)

For the JPSL's, finite element models were made and the natural frequencies of the lines under in-reactor conditions were found while the Hope Creek reactor was still under construction. The natural frequencies are dependent on the as-built line length. The as-built lengths of critical lines identified by analysis were measured while the Hope Creek reactor was still under construction. Impact tests on the critical lines were also performed. The analysis and test results indicated that the (( )) mode natural frequency of JPSL (( )) may be in resonance with the Vane Passing Frequency (VPF). Therefore, an additional support was added for JPSL (( )) while the Hope Creek reactor was still under construction. The rest of the lines had no problem with VPF resonance.

The JPSL resonance is due to recirculation pump speed and not due to power uprate, per se.

I Shroud 2 Shroud head and separator 3 Jet pumps 4 Feedwater sparger

_________________________]1 Table 14.49-1

14. 50 Question Deleted following Conference Call 14.51. In the past, small bore pipe failures due to high cyclic fatigue attributed to vibration have led to forced outages and plant shutdowns.

a) Attachment 8 of the Hope Creek EPU Submittal, Section 4.2.3, "FW Component Susceptibility", states that "Small bore lines attached to the FW headers were reviewed by experienced piping design engineers during RF13 (spring 2006)." Provide a summary of the review and a

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LR-N07-0099 LCR H05-01, Rev. 1 technical basis that justifies your conclusion that none of the FW small bore branch lines is susceptible to higher EPU vibration.

b) Attachment 8 of the Hope Creek EPU Submittal, Section 4.1.3, "MS System Component Review," indicates that similar review effort as in the FW small bore lines was performed for the small bore MS lines. Provide a summary of this review that qualifies the small bore MS lines for the higher EPU vibration.

c) Attachment 8 of the Hope Creek EPU Submittal, Section 4.1.3 states that "MS small bore connections in the vicinity of the TSV and TCVs will receive further evaluations to justify as-is, or alternatively, they will be modified to improve vibration resistance. Modifications may include strengthening of the sockolet welds or improvements in the supports.

This includes the Electro Hydraulic Control (EHC) unit connections."

Provide a summary of the evaluation for these small bore piping systems at the higher CPPU conditions. Also provide a description of any required structural modifications, as applicable.

Response

The EPU project team reviewed the piping isometrics, which include the small bore attachments and routing, and performed physical walkdowns during a refueling outage to determine actual conditions. They also had the quantitative vibration data collected in 2005 at 100% CLTP on the MS and FW main headers. This vibration monitoring program and CLTP data is summarized in Attachment 8 of the EPU submittal including the conclusion that there is substantial margins to address EPU increases. RAI 14.39 provides the diagrams for monitored lines and instrument locations.

The two individuals performing the evaluations had a combined 40 years of experience on piping design/stress evaluations and 35 years of experience specific to HCGS. The walkdown guidelines included noting any concerns on the condition of:

" The piping and pipe supports

" Insulation on subject and adjacent pipes.

" Attached components and branch lines

" Structures and components adjacent and below

• Valves, operators (actuators) and attached wiring, piping and tubing a) FW Small Bore Lines:

Inside Containment:

LR-N07-0099 LCR H05-01, Rev. 1 Each of the two 24-inch FW main headers has a check valve directly downstream of the containment penetration. A valve leakage test location is immediately downstream of each check valve. The containment penetration, an anchor, minimizes vibration at the check valve location.

Each of the two 24-inch headers splits into three 12-inch headers, which connect to the reactor vessel. Each of the six FW 12-inch headers has a cantilevered pressure instrument line. Although the highest FW vibration was recorded on these 12-inch headers, the maximum flow induced vibration (FIV) displacement was 1.6 mils (0.4 mils root mean square (RMS)) and the maximum vibration level was 0.108 g-RMS. These values reflect the limiting CLTP FIV case, recirculation pumps at maximum speed. Due to the low CLTP levels, adequate margin is available for EPU.

Reactor Building (Steam Tunnel)

As reported in Attachment 8 of the EPU submittal, the measured CLTP FIV levels on the 24-inch FW main headers in the steam tunnel are approximately 0.01 g-RMS, even lower than the FW values inside the containment.

The outboard FW check valves and their downstream FW valve leakage test connections are in close proximity to the containment penetration, an anchor which minimizes vibration at the check valve location.

The low vibration levels is judged not to pose any thereat to intermediate sized connections such as the 4-inch RWCU and 6-inch RCIC connections to the FW main headers.

Turbine Building:

As noted in Section 4.5 of Attachment 8, the'reactor feed pumps (RFPs) at EPU will be operating closer to their optimal design point. Since the RFPs are not expected to be a source of increased vibration levels and since the FW vibration levels (measured in the steam tunnel) are low, it was judged that the small bore connections in the turbine building by the 6 th point feedwater heater and RFPs the were not at risk.

b) MSL Small Bore Inside Containment:

There are four 26-inch main steam line (MSL) headers inside containment. The small bore connections are discussed below.

LR-N07-0099 LCR H05-01, Rev. 1 Inboard MSL/Main Steam Isolation Valve (MSIV) Drains: Piping is in close proximity to containment penetration, an anchor, that minimizes vibration on the lines.

Main Steam Flow Elements: Piping is designed for seismic events which would require the design to dampen any motion. The small bore connection to the large bore is a buttweld, which is more vibration resistant than a socket weld that is typically used for a small bore connection.

MSIV -Seal Leakoff: Piping was cut and capped in RF13. In addition, piping is in close proximity to the containment wall anchor.

MSLs Outside Containment thru to the Main Steam Stop Valves:

The MSLs are designed to seismic I criteria up to the Main Steam Stop Valves (MSSV).

MSIV -Seal Leakoff: A significant amount of piping was cut and capped earlier.

The piping is near the containment penetration, an anchor which minimizes vibration.

Main Steam Drain Pot Small-Bore Drains: One of the two small bore connections off of the 10-inch main steam drain pot was cut and capped. The second remains in service. The drain pot is near the containment penetration, an anchor which minimizes vibration.

Vent and Drain Piping Near Main Steam Stop Valves (MSSV): A double valve vent appears on each main steam line upstream of the MSSVs. This is a cantilevered mass. Downstream of each of the MSSVs is a 2-inch pipe that vents to the condenser through a 3-inch header pipe. This piping is analyzed for seismic loadings.

MSL connections downstream of the MSSVs are addressed in c) below.

c) The MS small bore connections downstream of the MSSVs, including the EHC connections, are addressed in the response to RAI 14.37.

LR-N07-0099 LCR H05-01, Rev. 1 14.52. Steam flow and feedwater flow will increase as a result of the CPPU implementation. Flow velocities affect pipe wall thinning due to flow accelerated corrosion (FAC). In the PUSAR Section 10.7 page 10-35 states that "Based on experience at CLTP operating conditions and previous FAC modeling results, it is anticipated that the CPPU operating conditions may result in the need for additional FAC monitoring points. The CHECWORKSTM FAC modeling techniques allow for the identification of additional monitoring points required for CPPU." Identify additional FAC inspection points that will result due to higher CPPU operating conditions.

Response

In December 2002, PSEG FAC Program performed a study using the CHECWORKS model software to evaluate the effect that a 20% OLTP uprate would have on the wear rates of the piping systems, and determine if the uprate would lead to significant pipe replacements. Increases in wear rates were found in portions of the condensate, feedwater, heater drain, and seal steam systems.

The study revealed that the power uprate operating conditions would have a minimal impact on FAC wear rates. The average predicted wear rate would not cause an increased need for physical modifications or replacements. Based on the 2002 study results, seven additional inspections have been added to the RF14 inspection scope to serve as a baseline to monitor the effects of post-CPPU wear rates. Components have been selected on the 'A' and 'B' feedwater trains downstream of the high pressure feedwater heaters, the seal steam system, the drain line from the #5C to #4C feedwater heater, and the drain line from the #4C to the #3C feedwater heater.

CHECWORKS is normally updated for best estimate operating conditions after the start of an operating cycle. This was done after RF1 3 (April 2006) for the RF14 (October 2007) inspection scope. Since operating conditions will change due to the CPPU after RF14, the CHECWORKS model is in the process of being updated with the new operating conditions. This process is being done prior to the outage to provide confirmation of the 2002 study that showed a minimal impact on FAC wear rates. Any additional monitoring or inspection required by the FAC program will be implemented based on the results of the updated model.

14.53 Question Deleted following Conference Call (previously addressed) 14.54 Discuss in detail how the verification and validation of the ACM computer code was pefrformed to satisfy th-e provisions of ASME NQA-1, "Quality Assurance Requirements for Nuclear Facility Applications," Subpart 2.7, "Quality Assurance Requirements for Computer Software for Nuclear Facility Applications," which has been accepted by the NRC in satisfying the requirements in 10 CFR Part 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel 69-LR-N07-0099 LCR H05-01, Rev. 1 Reprocessing Plants." Your discussion should include how the bench-marking of the ACM computer code was accomplished.

Response

ACM model development followed the procedures for analysis and calculation generation, and computer code generation, as found in the CDI Nuclear Quality Assurance Manual (Rev. 14). This manual requires compliance to applicable sections of ANSI N45.2, ANSI/ASME NQA-1, and reporting requirements of 10CFR Part 21. Model validation is detailed in CDI Report No.05-28P.

Model benchmarking is detailed in the attached document entitled "Bounding Methodology to Predict Full Scale Steam Dryer Loads from In-Plant Measurements."

LR-N07-0099 LCR H05-01, Rev. 1 BOUNDING METHODOLOGY TO PREDICT FULL SCALE STEAM DRYER LOADS FROM IN-PLANT MEASUREMENTS A Summary of Quad Cities Unit 2 Data Reduction and Model Validation LR-N07-0099 LCR H05-01, Rev. 1 LR-N07-0099 LCR H05-01, Rev. 1 i

1- -I-

+/-

LR-N07-0099 LCR H05-01, Rev. 1 i i.

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LR-N07-0099 LCR H05-01, Rev. 1 4 +

i i 4 +

I t

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LR-N07-0099 LCR H05-01, Rev. 1

• 1 1

-4 1

.4 .4

.4 1

.4 .4

))

14.55 Provide the validation documentation of 1/8th SMT, as well as the uncertainty and bias error associated with the use of SMT. The validation should include the benchmarking of the scale model by using the Quad cities 2 in-plant test data.

This validation should consider the low and the high frequency loading separately (e.g. comparison of the total RMS amplitude without considering the frequency content is not recommended).

Response

The CDI 1/8"' SMT for HCGS was benchmarked against the HCGS in-plant data at the same reactor power, CLTP. This is a better benchmark than a comparison with Quad Cities 2. Refer to the response to RAI 14.11 and CDI Report 07-01 P.

The data provided includes the PSD comparison between the HC SMT and the HC in-plant data for 0-200 Hz.

PSEG is not relying solely on the SMT. The FEA based on in-plant data at CLTP, 06-24 R3, demonstrates that the steam dryer has significant margin to accommodate an increase in any of the loads present at CLTP. The key purpose of the SMT was to provide as much information on SRV acoustic resonance that existing technology could provide since SRV acoustic resonance is not present at CLTP but is anticipated above CLTP. PSEG believes that the CDI 1 / 8 th SMT is the best technology presently available for this purpose. PSEG selected this

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LR-N07-0099 LCR H05-01, Rev. 1 option after the CDI SMT was able to replicate the QC2 in-plant relief valve acoustic resonance, which allowed CDI to design and test a fix on the 1 / 8 th SMT that successfully mitigated the QC2 relief valve acoustic resonance.

Note that when the margin recovery effort was undertaken, the SMT re-benchmarking assumed, conservatively, that SRV acoustic resonance commences immediately above CLTP. Using a SMT is a proactive step that PSEG undertook to reduce the power ascension unknowns. However, this does not substitute for verification during EPU power ascension. PSEG will perform the same key verification steps during the power ascension to ensure the steam dryer integrity as was done by earlier EPU power ascension that was done without the benefit of a SMT.

14.56 The licensee indicated in the public meeting that the Mach number used for the SMT data were re-evaluated and revised in Revision 2 of the CDI 06-27stress report. Discuss how you measure, calculate and validate the Mach number used in scale model testing for both 1/5th and 1/8th scales, including uncertainty.

Please submit this report to the NRC for the staff review.

Response

Please refer to the response to RAI 14.13. The rebenchmarking effort is discussed in CDI Report 06-16P Revision 2, in Attachment 6 to this letter.

14.57 Discuss how you monitor loading on the steam dryer due to vibration at the re-circulation pump bypass frequency.

Response

PSEG understands that the question was intended to read "...at the pump vane passing frequency".

HCGS as part of the remedial actions on recirculation system vibration had GE provide an analysis of the effects of the vane passing frequency (VPF) on the reactor internals. GE concluded that the components in the upper zone of the reactor, such as the moisture separators and dryers, are primarily affected by the increased steam flow rather than recirculation pump drive flow.

LR-N07-0099 LCR H05-01, Rev. 1 14.58 Discuss how you monitor the low-frequency pressure loading.

Response

The HCGS MSL strain gages include frequency response in the low range. If there is low frequency content in the MSL, it will be detected. CDI Report 07-01 P Figure 3.3 shows the PSD in-plant data for all strain gage locations.

Also, refer to RAI 14.60 on concerns as to the presence of high magnitude, low frequency loads at units, such as HCGS, without dead headed MS branch lines.

14.59 The data presented in the public meeting did not include the bias and uncertainty that are involved in the Strain Gage measurement, ACM analysis and Finite Element analysis. The licensee is requested to provide final stresses and fatigue margins in the steam dryer after all the structural modifications, uncertainties and bias errors are taken into account.

Response

No steam dryer modifications are planned. Refer to the response to 14.33 for the alternating stress ratio values considering the strain gage measurement and ACM uncertainty.

Refer to the response to RAI 14.2 for the uncertainty on the finite element analysis (FEA). The uncertainty in the FEA frequency is addressed by varying the frequency of the load by up to +/- 10% at 2.5% intervals. Other than frequency, the uncertainty in the FEA is small, and when combined by the square root of the sum of the square (SRSS) with the uncertainty in the strain gage and ACM, it is negligible. Note that the FEA assumed 1% Raleigh damping which is considered low. Damping values above 1% would reduce the alternating stresses.

LR-N07-0099 LCR H05-01, Rev. 1 14.60 The licensee indicated in the March 2, 2007, public meeting, that the conservatism factor was derived based on the pressure loads on the dryer resulting from the ACM analyses. The licensee is requested to discuss in detail how these conservative factors were determined. The information should include the effects of frequency content that are significant in the dynamic analysis and structural response. Also describe the accuracy of the ACM analysis for Hope Creek for low frequency pressure loading.

Response

The conservatism in the steam dryer pressure loads, discussed in the March 2, 2007 meeting, is shown in CDI Report 07-01P, submitted to the NRC on March 13, 2007. This report shows the conservatism of the 2006 in-plant data that was used as the input the CLTP FEM. Also, it includes the SMT benchmark against in-plant data at CLTP. Both comparisons include a PSD comparison up to 200 Hz.

LR-N07-0099 LCR H05-01, Rev. 1 LR-N07-0099 LCR H05-01, Rev. 1

))

LR-N07-0099 LCR H05-01, Rev. 1 14.61 Based on the Public Meeting with Hope Creek on Friday, March 2, 200[7], the NRC staff requests that the licensee provide the following two documents for review.

a) Report on RPV dome dynamic pressure data recently collected by PSEG (Structural Integrity Report).

Response

SIA report "Hope Creek RPV Dome Dynamic Pressure Data Reduction",

Report HC 31Q-301, is attached.

b) Stress report based on new in-plant strain gage data at CLTP with all gages working.

Response

HCGS did not rerun the CLTP Finite Element Analysis (FEA). At the March 2, 2007 meeting, PSEG stated that the CLTP FEA previously submitted (CDI Report 06-24 R3) was conservative since it used the 2006 loads, and CDI report 07-01, transmitted to the NRC on March 13, 2007 showed that there was significant conservatism in the 2006 loads. Since the CLTP FEA already showed significant margin in the alternating stress ratios, it was felt that there was no need to re-run the CLTP FEA.

14.62 In the conference call discussion on March 12, 2007, CDI indicated that the errors in flow rate were due to high friction in the piping in the subscale model.

The licensee is requested to provide:

a) Validation of Mach number of SMT for CLTP and EPU condition. The validation should be performed for the 1/8th and 1/5th subscale models.

Response

Refer to RAI 14.13.

b) ((

Response

R].

((I

))

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LR-N07-0099 LCR H05-01, Rev. 1 c) Revision or additional information to clearly indicate all the changes required in the earlier submittal (i.e., Attachment 7 to the September 18, 2006, EPU amendment request, Attachment 1 to the October 10, 2006 supplement, CDI reports 06-16, 06-17, and 06-26) which result from the new measurement and Mach number revision.

Response

Attachment 7 has been updated and is attached.

Attachment 1 to the October 10, 2006 supplement was an estimate of the EPU alternating stress ratios if the excess load conservatism was removed. It was done with the information then available. Subsequent to transmitting it, PSEG undertook a margin recovery effort, which is documented in CDI Report 06-27 R2 transmitted to the NRC on February 28, 2007. This was a more rigorous analytical effort to remove excess conservatism. It included re-benchmarking the SMT and rerunning the interim EPU FEA to recalculate the stress ratios. Thus, CDI Report 06-27 R2 is considered a replacement for Attachment 1.

CDI Reports06-16P and 06-17 have been updated to revision 2 and 3, respectively and are attached.

Technical Note 06-26 was submitted to the NRC in support of the algorithm developed to bound in-plant loads using data from only 2 MSLs.

Subsequently, in early February 2007, in-plant data was taken after all strain gage channels were restored to service. CDI Report 07-01P, submitted to the NRC on March 13, 2007 demonstrates that the 2006 data was conservative. Thus, there is no longer a need to demonstrate that the algorithm was conservative.

14.63 With regard to the 80 Hz the licensee indicated that RPV level sensor measurements show no strong acoustic mode existing at or near 80 Hz. It is requested to submit a report that documents these measurement data to the NRC for review as it uses these data to justify filtering 80 Hz peaks from their ACM model of the Hope Creek plant.

Response

SIA report "Hope Creek RPV Dome Dynamic Pressure Data Reduction", Report HC-31Q-301, is attached.

14.64 In a March 12 conference call, the licensee indicated that the new measurements show that the dryer load estimated from the first (May 2006) measurements at CLTP is conservative, in comparison to that determined from the new measurements. The licensee is requested to provide the frequency spectra of the new vs. old loads at CLTP conditions.

- 83 -

LR-N07-0099 LCR H05-01, Rev. 1

Response

CDI Report 07-01 P Figures 4.1 provides the dP comparisons across all low resolution nodes. Figure 4.3 provides the PSD comparison at the peak locations (outer hood) for 0 - 200 Hz. PSEG submitted CDI Report 07-01 P to the NRC on March 13, 2007.

LR-N07-0099 LCR H05-01, Rev. 1

7) Balance-of-Plant Branch (SBPB) (additional questions) 7.9 PUSAR Section 10.1.2 states that the mass and energy releases for high energy line breaks (HELB) in the feed water (FW) system were evaluated for CPPU and that the energy release from the FW line break at CPPU conditions is bounded by the energy release from a main steam line break at current licensed conditions. However, the PUSAR does not address the increased mass release from an HELB in the feed water system and its effect upon internal flooding. The Updated Final Safety Analysis Report (UFSAR) does address the mass release and its effect upon internal flooding for CLTP in Section 3.6.2.1.1.1.

a) Explain the effects of increased feed water flow from a feed water break at CPPU upon internal flooding including the effects stated in Section 3.6.2.1.1.1 of the UFSAR.

Response

The FW line break addressed in UFSAR Section 3.6.2.1.1.1 is postulated to occur in the reactor building main steam tunnel. The tunnel is designed for the effects of steam pressure and flooding from a double-ended break of the largest branch line of the feed water system. The existing FW line break flood analysis has been reviewed to determine the impact due to CPPU conditions. The flooding analysis is based on a mass release that results in a complete flooding of this tunnel. The mass release required to completely flood the tunnel does not change as a result of CPPU; consequently the UFSAR statement of Section 3.6.2.1.1.1 does not change for CPPU.

7.10 Section 6.4.1.1.1 of the PUSAR states in reference to the Station Service Water System (SSWS), that the CPPU effect is bounded by the loss of coolant accident (LOCA) analysis. Section 6.4.1.1.2 considers only a Safety Auxiliary Cooling System (SACS) LOCA heat load calculation when determining the adequacy of SACS for CPPU. Table 9.2-4 of the UFSAR for current licensed thermal power (CLTP) shows loss of offsite power (LOOP) as a greater heat load on SACS than LOCA in 3 of the 4 scenarios listed. The one scenario where Table 9.2-4 shows LOCA as a greater heat load than LOOP is for the one loop long term scenario (greater that 30 minutes for LOOP and greater than 10 minutes for LOCA).

Considering this information, please explain the following:

a) Explain in greater detail why the SACS is sufficient for recovery from a LOCA and from a LOOP for CPPU. To [facilitate] the staff's review, provide an updated Table 9.2-4 for CPPU.

LR-N07-0099 LCR H05-01, Rev. 1

Response

CPPU results in increased RHR heat loads during LOCA and LOOP design basis accident conditions due to increased suppression pool temperatures resulting from higher decay heat loads. For the design basis Case C LOCA at 102% of CPPU at 3840 MWt, the peak calculated post-LOCA suppression pool temperature is 212.3 0 F. For the design basis LOOP event, the peak suppression pool calculated temperature is 213.6 0 F.

It is assumed for the DBA LOCA (UFSAR Case C) that no offsite power is available coincident with a single failure of "A" EDG, resulting in a single RHR pump and heat exchanger available for suppression pool cooling. The same assumption is used in the LOOP scenario. In the LOCA, RHR is placed in service after 10 minutes. In the LOOP, RHR is placed in service after 15 minutes based on suppression pool temperature increasing above 120'F following manual SRV blowdown. The LOOP also assumes 100°F/hour cool-down rate, which exceeds the Hope Creek administrative cool-down limit of 90°F/hour.

The maximum calculated heat load removed at peak suppression pool temperature by a single RHR heat exchanger in the suppression pool cooling mode supplied with 100'F SACS cooling, K= 307 Btu/sec-°F, are calculated for the LOCA (212.3°F) and LOOP (213.6°F), as shown below:

LOCA QRHR = 3600 aRHR (TRHR - Tsacs)

= 3600 sec/hr (307 Btu/sec-°F) (212.30 F - 100'F)

= 124.1 MBtu/hr LOOP QRHR = 3600 0XRHR (TRHR - Tsacs)

= 3600 sec/hr (307 Btu/sec-°F) (213.60 F - 100'F)

= 125.6 MBtu/hr The evaluation of SACS heat loads conservatively uses a bounding value for both LOOP and LOCAs and is based on an assumed bounding suppression pool temperature of 215 0 F, as shown below:

QRHR = 3600 OCRHR (TRHR - Tsacs)

= 3600 sec/hr (307 Btu/sec-°F) (215°F - 100 0 F)

= 127.1 MBtu/hr The127.1 Mbtu/hr is a conservative upper bound heat rejection rate for the RHR heat exchanger for both LOCA and- LOOP conditions. The Hope -Creek-UHS analyses (calculation EG-0047) models the RHR heat load based on actual heat exchanger performance and does not assume a fixed heat load. Consequently, by observing the UHS technical specification temperature limits, the SACS HX outlet temperature will remain below its maximum post-accident value of 100°F.

LR-N07-0099 LCR H05-01, Rev. 1 A marked-up version of Table 9.2-4 is provided below.

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LR-N07-0099 LCR H05-01, Rev. 1 TABLE 9.2-4 STACS FLOWRATES AND HEAT LOADS FOR VARIOUS MODES OF OPERATION Normal Operation Normal Shutdown 6 & 6 gpm xlO Btu/h gpm xl0 Btu/h (2) (2) (2)

Motor-generator set coolers 1,800 11.2 11.27 800 11. 1,800 11.2 11.27 (1) (1) (1)

Main turbine lube oil coolers 1,725 9.36 1 725 .36 1,725 9.36 0 (1) (1) (1)

EHC hydro fluid cooler 10 0.04 10 .04 10 9.94 0 General H coolers 2,088 (4) *-,-21.95 2, 8 (4) 22.0 2,088 (4) 2*~ 0 2

(2) (2)

Generator stator coolers 1,866 1-6. 17.16 1,- 16.1 1,866 !,?89 i6.-,40 (1) (1)

Isophase bus coolers 100 &-a 9.-2 0.68 0.5 100 g--2 0 (1) (1)

Alterrex cooler 160 0.39 ] 0.39 160 -.39 0 (3) (3)

RFPT lube oil coolers 489 1.2 1.2 489 1.2 (3) (3)

Secondary cond pump coolers 30 0.13 0.13 30 0.13 (3) (3)

Turbine bldg chillers 10,500 .8 58.59 10, 58.5 10,500 8.5 58.59 (1) (1)

Process sampling 29 0.5 0.5 29 0.5 (1) (1)

Station air compressor 270 2.24 2.24 270 2.24 (1) (1)

Mezz pipe chase cooler 153 0.5 0.5 153 0.5 (4) (4)

Condenser compartment coolers 6 71 6ý 2-.G 2.205 2.0 671 64 2-4 2.205 (1) (1)

STACS Demineralizers 55 55 (11)

Htg sys cond. coolers 140 1.7 / 1.7 140 1.7 (Turbine Bldg.)

TACS total 4 4:24.68 19,723 124.68 9,723 24.-6 19,946 126.22 19,946 76.54 1 of HCGS-UFSAR Nc?;ember 1, 2003

- 88 -

LR-N07-0099 LCR H05-01, Rev. 1 TABLE 9.2-4 (cont)

STACS FLOWRATES AND HEAT LOADS FOR VARIOUS MODES OF OPERATION LOP < 30 Minut-e LOP > 30 Minutes One Loop Two Loops gpm xl0 6Btu/h 6 gpm x10 Btu/h Motor-generator set coolers Main turbine lube oil coolers EHC hydro fluid cooler General H2 coolers Generator stator coolers Isophase bus coolers Alterrex cooler RFPT lube oil coolers NOT NOT NOT NOT Secondary cond pump coolers OPERATING OPERATING OPERATING OPERATING Turbine building chillers Process sampling Station air compressor Mezz pipe chase cooler Condenser compartment coolers STACS Demineralizers (11)

Htg sys cond. coolers TAGS total la of 2 HCGS-UFSAR Septefebc.v 26, lO96

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LR-N07-0099 LCR H05-01, Rev. 1 TABLE 9.2-4 (cont)

LOCA < 10 Minutes LOCA > 10 Minutes One Loop Two Loops 6 6 gpm xl0 gpm xlO Btu/h gpm xl0 Btu/h Motor-generator set coolers Main turbine lube oil coolers EHC hydro fluid cooler General H2 coolers Generator stator coolers Isophase bus coolers Alterrex cooler RFPT lube oil coolers NOT N0TY NOT NOT Secondary cond pump coolers OPERATING OPERA OPERATING OPERATING Turbine building chillers Process sampling Station air compressor Mezz pipe chase cooler Condenser compartment coolers STACS Demineralizers (11)

Htg sys cond. coolers TACS total I

lb of 2 HCGS-UFSAR Spebr26, 1996

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LR-N07-0099 LCR H05-01, Rev. 1 TABLE 9-2-4 (cont)

STACS FLOW RATES AND HEAT LOADS FOR VARIOUS MODES OF OPERATION Normal Operating Normal Shutdown 6 4 6 gpm xl0 Btu/h -IG44 gpm xl0 Btu/h 26 (1) 0(1) 0 26 0 SACs Demineralizers (8) (1) +2+ *-2-(8)- 2.52 lE equipment chillers (Note 20) 408 2.52 408 8--6 8L(1)

(1) +-2+

RHR heat exchangers 8,65 7,000 47- 24.7 260

2) 1(15)

RHR pump seal coolers 10 18 23 (1/1) (15)4-64 _ 0-44 0.22 (7) (2) +4-)-

RHR motor bearing coolers 12 0.02 12 24 -.94 0.02 (2) (2 (2) (4-:2)- (12)

Fuel pool HXs (Note 20) 2,000 -6.1 17.6 2,000 16.1 2,000 i6. 17.6 (7)

Diesel generator coolers 1,600 19.75 (2) 7)

DG room coolers 760 1 3.83 (3)

FRVS cooling coils 1,140 (1) (7) (1) --2+

RHR pump room coolers 78 0.38 78 4-F6- 0.76 0.40 (1) (7) (1) +-2+

RHR pump and HX room coolers 89 0.38 89 4-7-9 0.46 0.42 (1)~ (7)

RCIC pump room coolers 136 0.06 136(1 (7)

Core spray pump room cooler 0.44 (1) (8) (1)

Control room chillers(Note 20) 1,588 7.53 3, 176; 7.53 1,588 7.53 (1) s (7 (8) (1)

Containment gas comp. (Note 20) 4 0.05 8j 0.05 4 0.05 (7)

HPCI pump room cooler Post LOCA sampling station 0.18 SACS total 4,026 4-099 26-29 27.70 1 50 11,228 2,749- 263.34 289.0

• 2+ *2+ +4+

(4) +4+ +4+

Grand total (STACS) 29,:723 15.988 3/2 176k.8(4) 4-3744-23,972 153.92 31, 174 365.54 2 of 2 HCGS-UFSAR Revirsion 12 LR-N07-0099 LCR H05-01, Rev. 1 TABLE 9-2-4 LOP 30 P inutcs LOP > 30 Minutes One Toon Two Loons Two Loons

/ 6 6 xl0 Btu/h gpm xl0 gpm xl0 Btu/h gpm xl0 Btu/h g\

(1) (2)

SACs Demineralizers 2**)(2) (8) 26 0 52 0 408* 2.52 816 2.52 (1) (1) +2+ +8+

1E equipment chillers 408 2 .52 408 4-0 2.52 (1) (17) (17) (2) 8,650 i93 .8- 127.1 18,000 1 ,(( 23 4.:79 209.2 RHR heat exchangers 8,650 26.00 17,300 52.00 (4) (15) (4) (i1 RHR pump seal coolers 20 (4) ( )0.36 20 ( (14) 0.36 20 --3 9 .62 0.36 20 4-6& 0 0.36 24 0.04 0.04 (4) (4)

RHR motor bearing coolers 24 0 .04 24 0.04 (2) (2) +4-+ (12) (2) (12)

Fuel pool heat exchangers 1,000 (2) *- 2, O00 2000 4-,Goo a .0- 17.6 2,000 6.9 17.6 (14) 3*(4) (4)({14)

(4) (14) (4) (14)

Diesel generator coolers 3,200 3.0 3 0 39.50 3,200 39 .50 3,200 39.50 (4)({16) (4) (16) (4) (16) (4) (16 1,520 7. 0 /520 7.60 1040 1,--2G 7. 60 1040 1,82 7.60 DG room coolers FRVS cooling coils (2) (2) (2) (2) 156 0.76 156 0.76 RHR pump room, coolers 156 0- 0.80 156 0.80

.-76 178 (2) 01) 0.7 178 (2) 0.76 178 (2) 0-.-4% 0.84 178 (2)

RHR pump and hx room coolers 0.84 (1) (1)

RCIC pump room cooler 13 (1) 0 6 13 (i) 0.06 13 - 0.09 13 0.09 Core spray pump room cooler (i) (2) (8) (1) (1) +2+ +8+

Control room chillers 1,588 7.53 3,6 7.53 1,588 7 .53 1588 7.53 (1) (1) (1) (1)

Containment gas compressor 4 4 0 .05 0.05

1) (1) (1)

HPCI pump room cooler 35 0.18 35 0.18 35 0 .48 0.23 35 0.23 Post LOCA sampling station SACS total 16~/6 85.31 28,442 \11.31 17,342 16,8090- 191.-4q 204.3 26,718 28,-468- 2-0.6 286.39

+-2+ +2+ +4+ 44+

Grand total (STACS) 796

,7(2) 85.31 28,442 (4) N\.31

[* (4) 16,809 28,468 310.68 17,342 204.3 26,718 286.39 2a of 2 HCGS-UFSAR Revisien 13

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LR-N07-0099 LCR H05-01, Rev. 1 TABLE 9.2-4 (cont)

(19)

LOCA < 10 Minutes LOCA > 10 Minutes I One Loop Two Loops 6 6 6 gpm xl0 Btu/h [Btu/h gpm xl0 Btu/h gpm xl0 Btu/h SACs Demineralizers (2) (2) 52(2) 52 52 (1) (8) (1) (1) +-2+ +8+

1E equipment chillers 408 2.52 81 2. 2 408 2.52 408 84-1-6 2.52 (1)(17) (17) (2)

RHR heat exchangers 8,650 i23.-89 127.1 18,000 -7,300 209.20 (4) (15) (4) (15) (4) (15)

RHR pump seal coolers 20 0.36 20 0.36 20 0.36 20 0.36 (4) (4 (4) (4)

RHR motor bearing coolers 24 0.04 24 .04 24 0.04 24 0.04 (2) -1+ 2 (2) -41+ (12) (2) (12)

Fuel pool heat exchangers 2000 1i904 8.96 17. 62,000 16.10 2000 ,-999 8.9s 17.6 2,000 16.19 17.6 (4) (14) (4 4) (4)(14) (4)(14)

Diesel generator coolers 3,200 39.50 3,200 39.50 3,200 39.50 3,200 39.50 (4) (16) (4) 16) (4) (16) (4) (16)

DG room coolers 1,520 7.60 1,520 7.60 1,520 7.60 1,520 7.60 (3) (18) (3 6 -4q-9-(18)

(4) (18) (6 (10)

FRVS cooling coils 1,400 4.40 2,280 .40 1,140 4.39 1140(3) ,89 4.39 (2) (2) (2)

RHR pump room coolers 156 G 0.8) 0 156 0.76 156 0--.-. 0.80 156 09-7 0.80 (2) 2) (2) (2)

RHR pump and hx room coolers 178 G.-- 0.8' 4 17 0.76 178 G.-6 0.84 178 0.76 0.84 (1) (1) (1)

RCIC pump room coolers 13 9.9- 0.09 13 0.06 13 O9-9 0.09 13 0.06 0.09 (4) -2+ (4) (2) (4)

Core spray pump room cooler 72 4-36 9- 4 .- 1 . 0 8 27 0.88 136 0.44 0.54 272 (-.88 1.08 (1) (2) (8) (1) (1) +(2- +-8+-

Control room chillers 1,588 P-.52 7.53 3,17 7.52 1,588 7-;2 7.53 1588 3,7--S: 7.52 7.53

+2+ (1) .0 (1) (1)

Containment gas compressor 4.0 0.05 4 0.05 (1) (1) (1) (1)

HPCI pump room cooler 35 G. 49 0.23 3 .18 35 1a 0.23 35 9.i& 0.23 (1) (1)

Post LOCA sampling station 10 0.11 10 0.11 SACS total 10,866 9,682 -2.24 82.62 73 19,134 18,082 196.44 209.33 28,620 2-,-94 29-93 291.97 I (4i4

+1 + +2+ (4) 4-2+ +2) (4) (4)

Grand total (STACS) 72.2'. 19,1Ga.- 96.33 31,004 290.03 10,866 82.62 19,134 209.33 28,620 291.97 NOTES:

(1) through (6) Number of units in operation, 1 through 6 units, respectively.

(7) Inter-mittent lead fer testlnq.

2b of 2 HCGS-UFSAR Refisiaion 13 November i4, 2003

- 93-LR-N07-0099 LCR H05-01, Rev. 1 TABLE 9.2-4 (cont)

NOTES: (cont)

I (9) Operator action is required to restart SFPCCS.

(11) Operation of this system occurs only during the heating season and corresponds with reduced load on the Turbine Bldg Chillers. Full flow and heat load will not occur coincident with full flow and load on the chillers. For this reason, this equipment is not included in the TACS System totals.

(12) If SAC HX outlet temp. cannot be maintained less than its max design temp., flow to fuel pool HX will be isolated to reduce total system heat load.

(13) The tables do not include the effects of a loss of instrument air.

(14) The min. flow rate to each EDG cooler is approximately 700 gpm; however, nuisance alarms may sound in the Control Room during periods of high SACS temperatures.

(15) The required SACS flow for the RHR pump seal coolers is 18 gpm if the RHR pump is supporting shutdown cooling. If the RHR's pump's suction is from the suppression pool, the required SACS flow to the RHR pump seal cooler is only 5 gpm.

(16) Value for one room cooler per EDG. For certain single failure conditions, two room coolers per EDG are required at 260 gpm and 1.1 Mbtu/hr per cooler (220 gpm and 1.1 Mbtu/hr per cooler for SACS loop outage and 2 EDGs crosstied, see Section 9.4.6.2.7) . Both room coolers are available for these particular single failure scenarios.

(17) RHRHX flow and heat load assumes suppression pool cooling only (limiting condition).

(18) Post LOCA, either three or four FRVS coils operating is acceptable. If four FRVS coils are assumed operating, for the short term LOCA (<10 minutes) the minimum required flow is 1400 gpm based on three at full capacity 380 gpm and one degraded at 260 gpm. With a minimum of three operating in the long term, the required flow is 1140 (380 gpm each). Also, four FRVS coils at an average flow of 340 gpm to eare also acceptable in either the long or short term.

(19) Post LOCA short-term (< 10 minutes), RHR heat load can be transferred to SACS during automatic LPCI injection.

(20) During elevated river water temperatures, assuming two SACS loop operation, these loads may be placed on the SACS loop not supplying the TACS loads.

(21) Component flow and heat load information in the above table is bounding for any power level up to a rated thermal power (RTP) of 3840 MWt for all components except the Generator Stator and Hydrogen coolers which are evaluated at 3673 MWt.

2c of 2 HCGS-UFSAR Revision 13 November 14, 2003

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LR-N07-0099 LCR H05-01, Rev. 1 b) Why is the heat load on the residual heat removal (RHR) heat exchangers for long term recovery from a LOOP the same as for recovery from a LOCA for the One Loop case in Table 9.2-4?

Response

RHR heat loads change due to CPPU during design basis accident conditions (LOCA and LOOP) due to increased predicted suppression pool temperatures and improved modeling. CPPU containment response analyses show post-LOCA pool temperature increases to 212.3*F and post-LOOP to 213.6°F based on operation at a RTP of 102% of 3840 MWt. The RHR heat load during a LOCA increases from 121.7 to 124.1 MBtu/hr, and decreases during a LOOP from 132.5 to 125.6 MBtu/hr. The decrease during the LOOP is due to removal of excess conservatism in the original design calculations that did not credit containment heat sinks and a revised decay heat model based on ANS 5.1 methodology.

The heat load used to evaluate CPPU is conservatively set at 127.1 MBtu/hr for both DBA cases as described above in the response to 7.10.a, above 1 .

c) Why are both SACS loops required for LOCA recovery as stated in the third to last paragraph in Section 9.2.2.2 of the UFSAR and does this affect the explanation requested above.

Response

The statement in question (3rd to last paragraph of 9.2.2.2) is accurate and remains valid for CPPU. In the short term (injection phase), cooling water from SACS is required for LPCI pump cooling and for Filter, Recirculation, and Ventilation System (FRVS) cooling. With two independently powered SACS pumps in each separate SACS header, no single active failure will defeat adequate cooling to these components. A passive failure is not postulated in the short term. In the long term, as stated in the next paragraph of 9.2.2.2, the SACS system provides its essential functions with a single active or passive failure.

d) Why is the LOCA the bounding analysis for the SACS heat load when Section 5.4.7.2.2 of the UFSAR states that in effect the shutdown cooling mode is the largest duty for the RHR heat exchangers?

Response

Shutdown cooling mode provides the largest duty for the RHR heat exchangers since the appr1a-c-teii/eratiure is apprloii ately 110°F high th~n the ap-pr6ach temperature for suppression pool cooling. However, this is not considered the In 1999 when the maximum post-accident SACS temperature was increased from 95°F to 100°F, a conservative CLTP value of 123.8 MBtu/hr was applied to both LOCA and LOOP events, based on the bounding assumption of peak suppression pool temperature at 212°F. This bounding CLTP value currently appears in Table 9.2-4 but will be replaced upon CPPU implementation with the new bounding value of 127.1 MBtu/hr.

LR-N07-0099 LCR H05-01, Rev. 1 bounding load for SACS analyses because (1) shutdown cooling is not used post-accident and (2) the plant cool-down rate is manually controlled by plant operators.

During a plant cool-down with peak UHS temperature, cool-down rates below 90°F/hour may be required (during the early part of the cool-down) to prevent exceeding the SACS maximum outlet temperature of 95 0 F. Hope Creek calculation BC-0052 (Plant Cool-Down using One RHR Heat Exchanger) evaluates this scenario. By observing the 95 0 F SACS outlet temperature limit for normal operation, BC-0052 shows that RCS temperature can be reduced below 200°F under CPPU conditions in 13.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> using only one RHR heat exchanger, thereby satisfying the 24-hour requirement. Prior to CPPU, the BC-0052 calculation result was 9.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. Since a large margin remains to the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> limit (10.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of margin), the peak UHS temperature could be increased without exceeding shutdown cooling requirements.

Following a LOCA, cool-down rates are not operator controlled. Reactor decay heat is released to the containment which is subsequently removed by RHR heat exchangers in the suppression pool cooling mode. In this case, the Technical Specification limits on UHS temperature are selected to prevent exceeding the 100°F maximum post-accident SACS outlet temperature. The Technical Specification UHS limits incorporate accident assumptions that include failures that can result in only one available RHR heat exchanger. The UHS temperature limits govern shutdown cooling operation as well.

e) Does Technical Specification Limiting Condition for Operation (LCO) 3.7.1.3 for the ultimate heat sink require revision for CPPU. Explain.

Response

TS 3.7.1.3 was revised in 2006. This amendment request was based on Hope Creek UHS calculations that accommodate the additional heat from CPPU. The primary calculation EG-0047, Revision 4 (UHS Temperature Limitations) was submitted to the NRC as part of LCR H05-012 (Amendment 168). No further changes to this calculation are required for CPPU. TS 3.7.1.3 does not require revision for CPPU.

8) SG Tube Integrity & Chem. Eng Br (CSGB) (additional questions)

Protective Coating Systems (Paints) - Organic Materials 8.7 Based on the UFSAR, it is the NRC staff's understanding that, (1) Service Level I coatings were not procured and applied according to RG 1.54 because it was not yet issued at the time most plant NSSS equipment was ordered, (2) most NSSS equipment is coated with inorganic zinc qualified to ANSI N1 01.2, (3) the amount LR-N07-0099 LCR H05-01, Rev. 1 of unqualified coatings on NSSS equipment is less than about 26 lb (12 Kg), (4) the drywell and exposed metal in the drywell and torus are coated with phenolic epoxy qualified to ANSI N101.2, (5) the suppression chamber is coated with phenolic epoxy qualified to ANSI N101.2, and (6) the total amount of unqualified coatings for non-NSSS equipment is less than about 242 lb (110 Kg).

a) Please provide a discussion on the qualification requirements for original and repair coatings to confirm or correct the staff's understanding.

Response

The Staff understanding as stated above is correct. While most NSSS coatings are qualified to the standards and requirements of ANSI N101.2, the quality assurance requirements of RG 1.54 were not applied because it was not yet issued. Further confirmation of the Staff understanding can be found in responses to RAIs 8.9, 8.10 and 8.12.

As stated in UFSAR Section 6.1.2.1.1, GE specified that most Nuclear Steam Supply System (NSSS) equipment be coated with a prime coat of inorganic zinc.

This coating was qualified under ANSI N101.2 for equipment exposed to DBA conditions inside the containment. New Service Level 1 coatings and repairs to existing Service Level 1 coatings are performed in accordance with PSEG standard NC.DE-TS.ZZ-6006(Q) based on ASTM D5144-00,"Standard Guide for Use of Protective Coating Standards in Nuclear Power Plants". The requirements of this standard are further described in the response to RAI 8.12.

8.8 In the PUSAR Section 4.2.6 provides an estimate for qualified containment coating debris of 85 lb (39 Kg) from "paint chips." The staff infers that this quantity corresponds to the highest value from NEDO-32686, Rev. 1 for combinations of inorganic zinc and epoxy (the predominant coating systems in the Hope Creek containment). Please confirm the staff's understanding or describe the methodology used for determining this value.

Response

Debris quantities in the suppression pool are based NEDO-32686 "Utility Resolution Guidance (URG) for ECCS Suction Strainer Blockage" guidance and tabulated in design specification H-1-VAR-MDS-0357(Q).

The staffs understanding as stated above is correct. In NEDO-32686, Volume Ill, on performance of containment coatings during a LOCA, three estimates of representative coating systems are provided. The estimates range from 47 lbs.

to 85 lbs. In its strainer performance evaluation, Hope Creek has applied the largest (85 lb.) value.

8.9 Section 4.2.6 provides an estimate for unqualified containment coating debris of 270 lb (122 Kg) following a design-basis LOCA. Because this value is equivalent LR-N07-0099 LCR H05-01, Rev. 1 to the sum of the amounts stated in Section 6.1.2 of the UFSAR for unqualified coatings on NSSS and non-NSSS equipment, the staff infers that the analysis assumes all unqualified coatings in primary containment will become debris following a LOCA. Please confirm the staff's understanding, or discuss the methodology for determining the amount of debris from unqualified coatings.

Response

The staff understanding that all unqualified coatings become debris following a LOCA is correct. As stated in sections 6.1.2.1.1 and 6.1.2.2 of the UFSAR, the total amount of unqualified paint in the containment for the NSSS is estimated to be less than 12 kilograms. The total quantity of unqualified coatings in the containment for non-NSSS supplied components is estimated to be less than 110 kilograms. The total amount of unqualified coatings in containment is the sum of 12 + 110 = 122 kg (270 Ib).

8.10 Please discuss the conditions (temperature, pressure, radiological dose) used to qualify Service Level I protective coatings in containment for current operating conditions and whether they remain bounding for DBA conditions following the extended power uprate. Ifthe original qualification conditions were not bounding for any coatings, please discuss your plans to qualify those coatings.

Response

Per ANSI Standards N1 01.2, "Protective Coatings (Paints) for Light Water Nuclear Reactor Containment Facilities" Service Level 1 applies to coatings inside primary containment where coating failure may form solid debris products under DB LOCA conditions. As stated in the UFSAR section 6.1.2.1, original GE specifications for most Nuclear Steam Supply System (NSSS) equipment required that they be coated with a prime coat of inorganic zinc. The suppression chamber is coated with a phenolic epoxy (immersion) compound. The drywell and exposed structural metal inside the drywell are coated with modified phenolic epoxy. Generally, exposed metal surfaces of equipment located inside the drywell and torus are coated with a modified phenolic epoxy. All of these coatings have been qualified in accordance with ANSI N101.2.

As stated in ANSI Standards N101.2, "For general qualification, protective coatings shall be tested at conditions that simulate the DBA environment with regards to temperature, pressure, chemical solutions and radiation." As stated in N 101.2, paragraph 4.4.2, IrradiationTest, DBA integrated radiation doses typically range between 1 E8 rad to 2E9 rad. The total integrated dose for testing coatings at Hope Creek was 9.5E8 rad. The bounding integrated radiation leviel following a DBA LOCA after CPPU implementation increases inside the drywell from 7.24E7 rads (CLTP) to 8.4E7 rads (CPPU).

Coatings at Hope Creek were tested at 340'F and 62 psig. As tabulated in PUSAR Table 4-1, containment pressure increases from 48.6 psig to 50.6 psig LR-N07-0099 LCR H05-01, Rev. 1 and temperature increases from 291 OF to 2980F following a DBA LOCA after implementation of CPPU (analyzed at 102% of 3952 MWt). Both of these parameters are below testing limits of 340°F and 62 psig.

As shown above, test conditions used for qualifying level 1 coatings remain bounding for plant conditions after implementation of CPPU.

8.11 Section 4.2.6 of the PUSAR states that CPPU conditions do not affect the methods for calculating the amount of debris generated, but it does not discuss the results of the calculations. Section 4.2.6 also states that the existing calculations for zones of influence remain valid because the pipe break locations do not change. These statements do not directly address the effects of the change to CPPU conditions, such as the size of the zone of influence (ZOI) at pipe break locations already postulated, or a comparison of the amount of coatings debris generated under current and CPPU conditions. Please discuss your evaluation of the effects of changing to CPPU conditions on the ZOI, the amount of coatings debris generated, and the operation of emergency cooling systems.

Response

Jet impingement loads have been evaluated for CPPU and determined to be bounded by existing calculations and analyses. Jet impingement thrust forces are determined by the equation F = K (P) (A), where K is the thrust factor, P is operating pressure, and A is the area of the break opening. The K factors used in the original analyses were conservative, bounding values that remain valid for CPPU. Similarly, break areas do not change. Therefore, pressures at break openings were evaluated. In all cases, pressures used in the original analyses were bounding and remained valid for CPPU. Hence, jet forces resulting from high energy pipe breaks are bounding for CPPU conditions.

Based on the methods and guidelines provided by the Utility Resolution Guidance for ECCS Suction Strainer Blockage (URG), dated November 5, 1996, CPPU does not impact the variables used in the determination of calculated debris volumes. As such, ECCS pump head losses due to debris accumulation on ECCS suction strainers located in the suppression pool are not affected by CPPU.

8.12 Please discuss your requirements for inspecting, removing, and replacing degraded containment coatings, and the effects of CPPU conditions on these activities.

Response

New Service Level 1 coatings and repairs to existing Service Level 1 coatings are performed in accordance with PSE&G technical standard NC.DE-TS.ZZ-6006(Q) based on ASTM D5144-00,"Standard Guide for Use of Protective LR-N07-0099 LCR H05-01, Rev. 1 Coating Standards in Nuclear Power Plants". Additional guidance for application, maintenance and periodic assessment of safety related coatings is provided in procedure ER-AA-330-008. CPPU conditions do not change requirements of either the technical standard or procedure since level 1 coatings have been qualified for conditions that bound CPPU (see response to 8.10).

Technical standard NC.DE-TS.ZZ-6006(Q) establishes requirements for the selection, application (including surface preparation), touchup, repair, inspection, and testing of the Service Level I protective coatings within the primary containment. Allowable coatings for Service Level 1 use inside containment are tabulated in Attachment 2 of the technical standard. Inspection guidance is provided in Section 4.9.1 of ER-AA-330-008. This section relies on NUREG 1801 (Chapter XI), Protective Coating Inspection and Maintenance Programs, for inspection guidance.

8.13 Table 6.1-3 of the Hope Creek UFSAR lists organic materials other than paint in the primary containment, including Kerite and Hypalon electrical insulation.

Please describe the evaluations you performed, and discuss the results of those evaluations, to determine the effects of CPPU conditions on the generation of hydrogen and organic gases from paints and other organic materials in containment (e.g., cable insulation).

Response

The total conductor insulation and cable jacket material, including Kerite and Hypalon conductor/jacket, was determined to support assumptions made in the post-LOCA control room and offsite dose calculation. The mass of acid forming Hypalon from cables inside of the containment was used as an input to the post-LOCA pH calculation to show that the pH of the suppression pool following a LOCA remains above 7 for 30 days. The Hypalon cable evaluation was performed prior to the EPU Project and is not affected by EPU since there is a negligible change to cable quantities inside containment. The pH calculation and the control room and offsite dose calculation were revised for the EPU (due to greater fission product quantities present at the beginning of the accident).

The requirement to determine hydrogen and oxygen concentrations and to start recombiners currently remains in the Hope Creek Severe Accident Guidelines (SAGs) as a potential beyond design-basis activity. However, the revised 10CFR50.44 (68FR54123 dated 9/16/03) does not define a design basis LOCA hydrogen release and eliminates the requirements for hydrogen control systems to mitigate such releases. Hope Creek License Amendment No. 16-0,-issued on 8/9/05, eliminates the Technical Specification requirements for hydrogen recombiners and hydrogen/oxygen monitors.

Any additional gases that are produced would result in oxygen dilution and slow the approach to hydrogen flammability. The CPPU analysis did not include the

-100-LR-N07-0099 LCR H05-01, Rev. 1 generation of gases from paints and other organic materials in containment and this is conservative.

Flow-Accelerated Corrosion (FAC) 8.14 According to the PUSAR, page 10-34, your FAC inspection program is based on guidelines from the EPRI and American Society of Mechanical Engineers.

Please describe in more detail your criteria for scoping and prioritizing components in your FAC program, including how your criteria compare to the guidance in EPRI NSAC-202L-R2, "Recommendations for an Effective Flow-Accelerated Corrosion Program."

Response

The Hope Creek FAC program follows the guidelines described in station procedure ER-AA-430-1 001 "Guidelines for Flow Accelerated Corrosion Activities" which emulates NSAC-202L-R2. Section 4.2.1 of the procedure describes inspection selection (scoping). The parts of the selection process are:

1) CHECWORKS Model: The Hope Creek FAC Program selects components based on the results of the model's output (i.e., wear rate and remaining life). Components are selected from both lines that have not been inspected and from lines that have inspected components.

2) Industry Experience: The Hope Creek FAC program selects inspection components based on operating experiences (OEs) from the industry that are applicable to Hope Creek. Periodically, OEs are reviewed for Hope Creek applicability. If the event is applicable, suitable components are selected to address the issue.

3) Station Experience: The Hope Creek FAC program selects inspection components based on station experiences. Periodically, the corrective action program is reviewed to discover if any situations had occurred that would be applicable to the program, i.e., valve leak-bys, steam leaks, abnormal valve usage (e.g., open when should be closed), etc.

The thermal performance report is also reviewed periodically to identify any applicable leaking valves whose piping may need to be inspected.

Inspection components are also selected based on requests from system engineers or from design changes.

4) Re-Inspections: The Hope Creek FAC program selects inspection components based on the next scheduled inspection (NSI) number.

The component's NSI is based on the wear rate and the minimum allowable wall thickness. Components with an NSI that is equal to the next outage are inspected.

- 101 -

LR-N07-0099 LCR H05-01, Rev. 1

5) Susceptible Non-Modeled (SNM): The Hope Creek FAC program selects inspection components based on the susceptibility of the non-modeled piping. A large amount of FAC susceptible piping cannot be modeled because of a lack of operating parameter data. This includes almost all of the small-bore piping. This also includes feedwater heater shells. Lines that are deemed highly susceptible and could have detrimental consequences if failure occurred are slated for inspection.

6) Engineering Judgment: The Hope Creek FAC program also selects inspection components based on engineering judgment. Engineering judgment is used when selecting inspection locations through knowledge based on experience.

Priority is given to inspection locations which were inspected based on plant experience or on components that may require repair.

8.15 Please describe your most recent repair or replacement performed as a result of FAC. Include in your description the component replaced, the extent of degradation, actions to prevent recurrence, and how this experience was used to update the FAC program for existing and EPU conditions.

Description:

In December 2006, a 26-inch x 26-inch fabricated carbon steel tee on the #3A feedwater heater (FWH) extraction steam supply line developed a through-wall leak approximately 1.5 inches downstream of the upstream weld.

The tee is immediately downstream of a 26-inch check valve and immediately upstream of a 26-inch gate valve and connects to the #3 feedwater heaters extraction steam supply equalizer line. The through-wall hole was approximately 3/8-inch x 1-inch. The repair consisted of a carbon steel 1/2A-inch thick welded patch 4 inches wide extending 260 degrees around the upper main run of the tee.

Extent of Condition: Using a refined grid pattern (1-inch x 1-inch), wall thickness measurements of the tee showed degradation for the first two to three inches downstream of the check valve then an abrupt recovery to above nominal Wall thickness. Wall thickness measurements were taken on the #3C FWH tee and the results showed that degradation was present in the same area as the 'A' tee but the 'C' tee was above the allowable minimum wall thickness and would remiaini so throughout the opera ficiyc e. The 'A' and the 'C tes we~re ast inspected in 2000 (4.5 cycles) and 1999 (5.5 cycles) respectively. The 'B' tee had been inspected consecutively for 6 cycles. It was last inspected in 2004 and the wear trend indicated minimal degradation.

- 102 -

LR-N07-0099 LCR H05-01, Rev. 1 Recurrence Prevention: A permanent repair of the 'A' tee will be planned for RF14 (October 2007). PSEG also plans to address degradation to the 'C' tee and to inspect the 'B' tee during RF14.

This event will not cause any changes in the program. A reduced grid size was used to detect the extent of the degradation. A parallel component was inspected for extent of condition. The degraded tee was repaired and placed back in service. The tees will continue to be monitored.

8.16 Please discuss how components are inspected and evaluated with respect to the guidance in EPRI NSAC-202L-R2, in which suitability for continued service is based on current wall thickness, acceptable wall thickness, and predicted wall thickness at the time of the next inspection. Discuss how your acceptance criteria for minimum wall thickness are consistent with maintaining structural integrity.

Response

At Hope Creek, the components are inspected using the ultrasonic techniques (UT). Grid size and layout for the components are those recommended by NSAC-202L-R2. Grid size is reduced in areas of significant wear. Wall thickness data is received from the field and is downloaded into the "FAC Manager" evaluation software. "FAC Manager" is an evaluation tool produced by CSI Technologies, Inc., and is used at numerous nuclear plants. Various design parameters, including minimum allowable wall thickness as calculated by the mechanical design group, are loaded into the evaluation tool for each component. The data is reviewed and evaluated. The evaluation methods available through "FAC Manager" are those described in NSAC-202L-R2. The choice of the method is dependent on the type of component and its history.

Results of the evaluation reveal if the component will remain above minimum allowable wall thickness throughout the next operating and what the predicted minimum wall thickness will be at the end of the operating cycle. Additionally, the evaluation shows the remaining service life of the component (based on the calculated minimum allowable wall thickness) and the next scheduled inspection (NSI) outage. The NSI is an outage prior to the time that the component reaches minimum allowable wall thickness. Piping is analyzed using minimum wall thickness according to the Hope Creek design methodology and acceptable only if it meets all the design requirements of Hope Creek.

8.17 The PUSAR states on page 10-35 that the temperature of selected portions is predicted6to increase a maximum of 13'F. However, PUSAR Table 10-12 indicates the temperature of some components is predicted to increase by up to 19'F (e.g., condensate piping temperature changing from 154 - 365°F at the current power level to 153 - 384 0F at higher power). Please clarify this apparent discrepancy between your statement about the maximum temperature change and the values shown in the table.

103 -

LR-N07-0099 LCR H05-01, Rev. 1

Response

The 130 F is the calculated FW temperature difference between the reactor heat balances for 120% OLTP and for CLTP.

For the purposes of evaluating flow assisted corrosion (FAC), the HCGS program inputs the temperature predicted for the specific line, rather than a single generic value. The values are based on the plant's heat balance which includes the balance of plant power block components (for example, the turbine, feedwater heaters (FWH), condenser, condensate system, extraction steam, FWH drain lines, moisture separator, and feedwater system). As can be seen in Table 10-12, the power uprate temperature impact varies with the line and system. The temperatures reported in PUSAR Table 10-12 were obtained from the turbine heat balance for 120% OLTP (GE Turbine, AA-02-086.v.3 dated 09/10/02) and the CLTP heat balance of record for HCGS, which is based on the PEPSE model (H-1 -ZZ-M DC- 1847).

8.18 Table 10-12 in the PUSAR describes changes in the variables that affect FAC rates. However, because the FAC rate is determined by the interactions of these variables, comparing the parameters may not indicate how they affect the FAC rate. Please discuss the effect of the CPPU conditions on the FAC rates as predicted by your CHECWORKS model, for example by providing the FAC rates for several (e.g., 5 to 10) components representing the highest predicted FAC rates before the uprate and the highest rates after the uprate (i.e., two potentially different sets of components). Include the calculated corrosion rates for comparison.

Response

In December 2002, a study was performed using the CHECWORKS model software to evaluate the effect that the then proposed 20% power uprate would have on the wear rates of the piping systems and if the uprate would lead to significant pipe replacements. The study concluded that the average predicted wear rates would not cause an increased need for replacements. Increases in wear rates were found in portions of the condensate, feedwater, heater drain, and seal steam systems.

The following tables contain high wear rate components that are representative of various systems. This method was used to eliminate the possibility of listing several high wear components on the same line and/or the component's exact location on a parallel train.

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LR-N07-0099 LCR H05-01, Rev. 1 Before Uprate

• System Line Name Component ID Description Wear Rate

/Year (mil)

Seal Steam Steam Seal 1-AF-004-S06-T2 Tee 27.4 Evaporator Steam Supply Extraction Steam Cross Under 1-AC-012-S20-T1 Tee 20.3 Steam to #5 Feedwater Heater Header Heater Drains #6B Feedwater 1-AF-076-S06-L2 Elbow 18.3 Heater to #5B Feedwater Heater Extraction Steam Cross Under 1-AC-012-S22-T1 Tee 17.2 Steam Header to

1. 5B Feedwater Heater Heater Drains #4A Feedwater 1-AF-108-S01-L1 Elbow 14.9 Heater to #3A Feedwater Heater Feedwater Minimum Flow 1-AE-005-S03-T1 Tee 13.4 Line Cross Around Header Tee to 'A' AC33-N-001 Nozzle 13.4 Moisture Separator Moisture 'A' Moisture 1-AC-031-S04-T1 Tee 11.9 Separator Drains Separator to Header Tee

• Based on RF13 inputs and Cycle 13 operating conditions.

After Uprate#

System Line Name Component ID Description Wear Rate

/Year (mil)

Seal Steam Steam Supply to 1-AF-004-S05-P1 Pipe 25.9 Steam Seal Evaporator Heater Drains FWH 4B to FWH 1-AF-109-S01-L1 Elbow 16.1 3B Condensate Secondary 1-AD-111-SO1-N1 Nozzle 15.8 Condensate Pump Discharge Header Condensate 3A Feedwater 1-AD-031-S01-N1 Nozzle 15.1 Heater to 4A Feedwater Heater

- 105-LR-N07-0099 LCR H05-01, Rev. 1 System Line Name Component ID Description Wear Rate NYear (mil)

Heater Drains 3A Feedwater 1-AF-141-S02-L2 Elbow 13.6 Heater to 2A Feedwater Heater Heater Drains 5C Feedwater 1-AF-095-S03-L1 Elbow 12.3 Heater to 4C Feedwater Heater Condensate 2A Feedwater 1-AD-062-S01-L1 Elbow 10.2 Heater to #2 Feedwater Heater Header

1. After uprate wear rates are based on the operating conditions of the 120%

OLTP uprate and input data from refueling outage RF10 (Autumn 2002) 8.19 Please discuss your FAC program for small bore piping, including how it compares to the guidance in Appendix A of NSAC-202L-R2.

Response

The Hope Creek FAC Program is in the process of establishing a small-bore program as part of the Hope Creek Asset Management Initiative. NSAC-202L-R2 Appendix A will be used as guidance. A Susceptible Non-Modeled Analysis (SNM) of the FAC program's susceptible piping has been completed. Following the guidelines addressed in Appendix A of NSAC-202L-R2, the program's piping that could not be modeled by CHECWORKS was ranked based on consequence of failure and degree of susceptibility (based on assumed operating conditions).

Small-bore piping is included in the SNM. The small-bore piping whose consequence of failure is more than minimal and whose susceptibility to FAC is high (based on assumed operating conditions) have been identified and will be included in the program's pro-active pipe replacement program. The small-bore piping whose susceptibility is less than high will be scheduled for inspections and monitored. Numerous small-bore lines have already been replaced with non-susceptible material. HPCI/ RCIC steam supply drain lines, Reactor Feed Pump Turbine Steam Chest drain lines, valve drain lines, and feedwater heater operating vent lines are an example of the small-bore piping that has been replaced with non-susceptible material.

-106-LR-N07-0099 LCR H05-01, Rev. 1 8.20 Please discuss how your program addresses flow-related thinning other than FAC, such as erosion-corrosion due to high velocity fluids or suspended particles.

Response

The Hope Creek FAC Program inspects certain components for degradation caused by liquid droplet impingement (LDI). Indications that LDI may be present are valve leak-bys, or conditions (open valves, leaks) that cause the velocity of the two-phased mixture to increase dramatically. The FAC program also inspects for cavitation per system engineering requests.

References

1. PSEG letter LR-N06-0286, Request for License Amendment: Extended Power Uprate, September 18, 2006

2. NRC letter, Hope Creek Generating Station - Request for Additional Information Regarding Request for Extended Power Uprate (TAC NO. MD3002), April 20, 2007

- 107-LR-N07-0099 LCR H05-01, Rev. 1 Hope Creek Generating Station Facility Operating License NPF-57 Docket No. 50-354 Extended Power Uprate 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.

NRC question 14.46 is restated below followed by PSEG's response.

14) Mechanical & Civil Engineering Branch (EMCB) 14.46 The increased main steam flow results in increased forces and moments from the Turbine Stop Valve (TSV) closure transient at CPPU conditions. The TSV load was used in the design of MS piping system. Provide a quantitative summary of the main steam and associated piping system evaluation (inside and outside containment), including pipe supports, that contains the increased loading associated with the TSV closure transient at CPPU conditions along with a comparison to the code allowable limits. Include data at critical locations (i.e.

nozzles, penetrations, etc). For nonconforming piping and pipe supports, provide a summary of modifications required to ensure that piping and pipe supports are structurally adequate to perform their intended design function and the schedule for completion of these modification[s].

Response

Main steam piping supports and other critical piping components are evaluated for the Turbine Stop Valve Closure (TSVC) loads. The TSVC load diminishes rapidly for branch piping and is considered insignificant after the first support and/or elbow attached to MS piping. The attached pages from 14.46-3 to 14.46-101 provide the stress data at critical locations and allowable vs. calculated values for CPPU conditions.

All of the supports that are listed in pages 14.46-102 to 14.46-189 include the existing (CLTP) loads, extended power uprate loads ("CPPU" or "EPU"), and the ratio of extended power uprate loads to Design Load The supports that were modified for the CPPU condition are listed in page 14.46-190. No other supports require modification due to CPPU Loads.

- 14.46 LR-N07-0099 LCR H05-01, Rev. 1 References

1. PSEG letter LR-N06-0286, Request for License Amendment: Extended Power Uprate, September 18, 2006

2. NRC letter, Hope Creek Generating Station - Request for Additional Information Regarding Request for Extended Power Uprate (TAC NO. MD3002), April 20, 2007

- 14.46

=>

0 3

CD MS Line A I'-

0 A-2

-*C" CD0

• CO

0 CD Project: Hope Creek Extended Power Uprate Calculation No.: C-0122, Rev. 9 *Page No: 22, 34, 36, 253,293, 325, 329, 333,390 Calculation

Title:

Main Steam Line 'A', MSRV Lines A, J & R

Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable Power Uprate Stress* Stress Stress to Allowable (PSI) (PSI) (PSI) 0)

Equation : 9 Normal/Upset 011 24,637 26,041 31,860 0.82 Emergency 011 22,480 22,480 39,825 0.56 Faulted 011 24,941 26,363 53,100 0.50 Equation: 10 ** 045F 58,860 60,567 53,100 1.14 Equation :12 045F 40,724 40.724 53,100 0.77 Equation :13 045F 22,627 22,627 53,100 0.43 Equation :14 (Fatigue) CUF 200 0.04 0.041 0.100 OK 17-C-)

from GE Stress Report 23A6126 Rev. 1 dated 2/7/89 which is a part of Calculation No. C-0122 Rev. 9 0 Since equation 10 is not satisfied, the piping is qualified by equations 12 and 13. Cnl r--

A-3 CD

ý0 CD

0 3

CD Project: Hope Creek Extended Power Uprate Calculation No.: C-0122, Rev. 9 *Page No: 22, 34, 36, 253, 293, 325, 329, 333, 390 Calculation

Title:

Main Steam Line 'A', MSRV Lines A, J & R

Description:

Main Steam Piping - Line A-CRITERIA PER ASME Node Original Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable Power Uprate Stress* Stress Stress to Allowable

-I (PSI) PSI) (PSI 0P 4h. Equation : 9 Normal/Upset 011 24,637 26,041 31,860 0.82 Emergency 011 22,480 22,480 39,825 0.56 Faulted 011 24,941 26,363 53,100 0.50 Equation:10* 064-*" 58,421 60,115 53,100 1.13 Equation :12 064- 37,933 37,933 53,100 0.71 Equation 13 064* 22,327 22,327 53,100 0.42 Equation :14 (Fatigue) CUF 200 0.04 0.041 0.100 OK 0

• from GE Stress Report 23A6126 Rev. 1 dated 2/7/89 which is a part of Calculation No. C-0122 Rev. 9 Since equation 10 is not satisfied, the piping is qualified by equations 12 and 13.

• • • Elbow '064 to '063N

-C)

-4 A-4 (D C0

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Title:

Main Steam Line 'A', MSRV Lines A, J & R

Description:

Main Steam Piping - Line A***

CRITERIA PER ASME Node Original Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable Power Uprate Stress* Stress Stress to Allowable (PSI) (PSI) (PSI)

Equation: 9 Normal/Upset 011 24,637 26,041 31,860 0.82 0)

Emergency 011 22,480 22,480 39,825 0.56 Faulted 011 24,941 26,363 54,600 0.48 Equation :10 ** 064*** 56,809 58,456 53,100 1.10 Equation :12 064** 36,516 36,516 53,100 0.69 Equation :13 064** 22,293 22,293 53,100 0.42 Equation: 14 (Fatigue) CUF 200 0.04 0.041 0.100 OK 0

• from GE Stress Report 23A6126 Rev. 1 dated 2/7/89 which is a part of Calculation No. C-0122 Rev. 9 I Since equation 10 is not satisfied, the piping is qualified by equations 12 and 13.

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Project: Hope Creel Extended Power Uprate Calculation No.: C-0122, Rev. 9 *Page No: 22, 34, 36,253, 293, 325, 329, 333, 390 Calculation

Title:

Main Steam Line 'A', MSRV Lines A, J & R

Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable Power Uprate Stress* Stress Stress to Allowable (PSI) (PSI) (PSI)

0) 26,041 31,860 0.82 Equation :9 Normal/Upset 011 24,637

-!4 Emergency 011 22,480 22,480 39,825 0.56 Faulted 011 24,941 26,363 53,100 0.50 Equation :10 064*'* 56,803 58,450 53,100 1.10 Equation 12 064*.. 36,516 36,516 53,100 0.69 Equation :13 064*** 22,293 22,293 53,100 0.42 Equation :14 (Fatigue) CUF 200 0.04 0.041 0.100 OK 17-0

• from GE Stress Report 23A6126 Rev. 1 dated 2/7/89 which is a part of Calculation No. C-0122 Rev. 9

• • Since equation 10 is not satisfied, the piping is qualified by equations 12 and 13.

• • • Elbow 064 to 065F

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=r 3-4 Project: Hope Creek Extended Power Uprate Calculation No.: C-0122, Rev. 9 *Page No: 22, 34, 36, 253,293, 325, 329, 333, 390 Calculation

Title:

Main Steam Line 'A', MSRV Lines A, J & R

==

Description:==

Maximum Stress Intensities from GE Stress Report 23A6126 Rev. 1dated 2/7/89 which is a part of Calculation No. C-0122 Rev. 9 C)

Branch Connection Sweepolet CDi

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(D Project: Hope Creek Extended Power Uprate Calculation No.: C-0122, Rev. 9 *Page No: 22, 34, 36, 253, 293, 325, 329, 333, 390 Calculation

Title:

Main Steam Line 'A', MSRV Lines A, J & R

Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable Power Uprate Stress* Stress Stress to Allowable 0') (PSI) (PSI) (PSI)

(b Equation : 9 NormaVUpset 011 24,637 26,041 31,860 0.82 Emergency 011 22,480 23,761 39,825 0.60 Faulted 011 24,941 26,363 53,100 0.50 Equation :10 021 N** 47,988 49,380 53,100 0.93 Equation :12 021N- 25,673 25,673 53,100 0.48 Equation: 13 021N*** 22,694 22,694 53,100 0.43 Equation: 14 (Fatigue) CUF 200 0.04 0.041 0.100 OK

• from GE Stress Report 23A6126 Rev. 1 dated 2/7189 which is a part of Calculation No. C-0122 Rev. 9 C--

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=3 Project: Hope Creek Extended Power Uprate Calculation No.: C--0122, Rev. 9 -Page Nos.: 86, 87, 88 Calculation

Title:

Main Steam Une A',. MSRV Lines A, J & R

Description:

Main Steam Piping-Une A MSIV (Inboard) Bonnet Flange Loads Evaluation*

Maximum of Loading Existing Mxu ow Ratio=.

Inboard MSIV Node Pt. ASME Load Loading Combination EPU Loads Enveloped EPU & AllowableRemark Level Enveloped (Note 1) Allowable with TSV Load ALlads 0

0:) Level B Axial F 2,155 2,786 2,155 2,786 29,270 0.0952 OK (Note 2)

Moment 514,872 665,729 514,872 665,729 1,610,000 0.4135 OK (Note 2)

Level C Axial F 2,155 2,786 2.155 2,786 29,270 0.0952 OK (Note 2)

Inboard 073 Moment 514,872 665,729 514,872 665,729 1,610,000 0.4135 OK (Note 2)

Level D Axial F 2,155 2,786 2,566 2,786 29,270 0.0952 OK (Note 2)

Moment 514,872 665,729 610,529 665,729 1,6`1,000 0.4135 OK (Note 2)

Level A is not affected by TSV Loads.

Forces and Moments are In Lbs and in In-Lbs respectively.

nfromGE Stress Report 23A6126 Rev. I dated 217/89 which is a part of Calculation No. C-0122 Rev. 9 Note 1: Allowable are taken from GE Document 213A8411 Rev.0 r-)

0 Note 2: The load increase due to TSV is greater than the existing enveloped loads. However, the increase is lower than the allowable.

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Project: Hope Creek Extended Power Uprate Calculation No.: C-0122 Rev. 9 -Page Nos. 86, 89 through 92 Calculation

Title:

Main Steam Line' A', MSRV Lines A, J & R

Description:

Main Steam Line A - Inboard MSIV Inlet and Outlet Connection Loads Evaiuation Loading Existing Maximum of Ratio=

Inboard MSIV Node Pt. ASME LeeLoad Loading Combination EPU Loads Enveloped Enveloped Allowable Maximum/All Remark with TSV Load Loadsowabe Axial Stress 7,940 1 ,940 10.266 17,700 0.580 ( Note 1)

Level B Tor. Stress 768 993 768 993 17,700 0.056 (Note 1 Bend. stress 4,977 6,435 4,977 6,435 17.700 0.364 (Note 1)

Axial Stress 7,Q94 10,266 7,940 10,258 17,700 0.580 (Note 1 )

Inlet 069 Level C Tor. Stress 768 993 768 993 17,700 0.056 ( Note 1 )

Connection Bend. stress 4,977 6,435 4,977 6.435 17,700 0.364 ( Note I )

0D Axial Stress 7,940 10,266 7,940 10,266 17700 0.580 ( Note 11 Level D Tor. Stress 768 993 830 993 17*700 0.056 1Note I1 Bend. stress "77 6.435 4.977 6,435 17,700 0.364 INote 1I Axial Stress 7,937 10,263 7,937 10,263 17,700 0.580 (Note 1i Level B Tor. Stress 800 1.034 800 1,034 17,700 0.058 (Note 1 1 Bend. stress 3,557 4,599 3,557 4,599 17,7D0 0.260 (Note 1)

AxIal Stress 7,937 10.263 7.937 10,263 17,700 0.580 (Note 1) outlet 077 Level C Tar. Stress 800 1,034 800 1,034 17,700 0.058 1 Note 1I connection Bend. stress 3,557 4,599 3,557 4,599 17,700 0.260 1 Note 11 Axial Stress 7.943 10_2943 10 ,270 17,700 0.580 (Note I1 Level D Tor. Stress 708 915 764 915 17,700 0.052 (Note 1 Bend. stress 3,627 4,690 3,794 4,690 17,700 0.265 (Note 1)

• Level A is not affected byTSV Loads.

Stresses are in psi r-0

-from GE Stress Report 23A6126 Rev. 1 dated 2t7/89 which is a part of Calculation No. C-0122 Rev. 9 Note 1: The load increase due to TSV Is greater than the existing enveloped loads, However, the increase Is lower than the allowable.

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Project: Hope Creek Extended Power Uprate Calculation No.: C-0122 Rev. 9 *'PageNos. 47 through 50 Calculation

Title:

Main Steam Line' A', MSRV Lines A, J & R

Description:

Main Steam Piping - Une A MSIV Acceleration Loads Evaluation" Maximum of ASMELevelLoad Loading Direction Loading Combination EPU Loads Existing Enveloped EPU &

Enveloped Qualified by Test Remarks with TSV' Load* Loads

0) AH 0.388 0.502 0.388 0.502 (Note 1) OK Level B AV 0.171 0.221 0,235 0.235 (Note 1) OK 071 B _

r.'3 LevelD AH 0.417 0.539 0.545 0.545 (Note 1) OK AV 0.271 0.350 0.616 0.616 (Note 1) OK AH 3.290 4.254 3.290 4.254 (Note 1) OK Level 8 0.598 0.774 0.598 0.774 (Note 1) OK 076 Operator AV Level D AH 3.351 4.332 4.024 4.332 (Note 1) OK AV 0.864 1.117 0.864 1.117 (Note 1) OK Level B AH 0.610 0.788 0.613 0.788 2.0 (Note 1) OK 073 Bonnet AV 0.221 0.286 0.221 0.286 3.0 (Note 1) OK Flange OK LevelFD A- 0.615 0.795 1.233 1.233 4.5 (Note 1)

AV 0.265 0.342 0.420 0.420 5.0 (Note 1) OK

• from GE stress Report 23A6126 Rev. I dated 2/7/89 which is apart of Calculation C-0122 Rev.9

• -Acceleration Loads are in G's Note 1: NEDC 31020 - Hope Creek Environmental Qualification Report MSIV Actuator, Attachment 5 and Appendix H.

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Project: Hope Creek Extended Power Uprate Calculation No.: C-0122 Rev. 9 *Page Nos. 44, 45 and 46 Calculation

Title:

Main Steam Line' A', MSRV Lines A, J & R

Description:

Main Steam-Line A RPV Nozzle Loads Evaluation ASME Load Loading Loading EPU Loads Ratio=

Component Node Pt. Aeve LNoad L)(Noad Combination Allowable Maximum/ Remark Level (Note 1) (Note 2) with TSV (ANote3) lowable Secondary HR 81,970 105,987 674,200 0.157 OK (Level B)

W~

Secondary MR 3,513,508 4,542,966 14,384,000 0.316 OK (Level B) I I RPV 001 _________

Primary HR 47,989 62,050 374,520 (Level D) 0.166 OK Primary (Level D) MR 2,047,157 2,646,974 7,417,000 0.357 OK Forces and Moments are In Lbs and in In-Lbs respectively.

• from GE Stress Report 23A6126 Rev.1 dated 2f7/89 which is a part of Calculation No. C-0122 Rev. 9 Note 1 Levels A and C are not affected by the TSV loads Note 2 For Secondary load level, HR and MR are the envelop of case 1 and case 3 HR and MR are calculated at the RPV I NOZZLE junction using given FX, FY, FZ, MX, MY, MZ load components.

Note 3 For the EPU loads , only the load combination with TSV load is assumed to be impacted, C:)

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CD Project: Hope Creek Extended Power Uprate Calculation No.: C-0122 Rev. 9 -Attachment 9, Page Nos. 19 and 20 Calculation

Title:

Main Steam Line' A', MSRV Lines A, J & R

Description:

Main Steam Piping. Line A Penetration Flued Head Evaluation*

ASME -Loading **Loading Existing EPU & Allowable Ratio = Remark Anumber Node Pt. Le Load in Combination EPU Loads Envelopd EU& Alwbe Rtoemr NumbLevel Direction with TSV Load Enveloped (Note 1) Max/Allowable Level_____ with__Load Loads Axial (P) 50,391 65,156 50,391 65.156 79,000 0.825 OK OBE + Shear (V) 3.963 5,111 3,953 5,111 28,300 0.181 OK 0)

TSVC Moment (Ml 574,563 742,910 574,563 742,910 2,212,000 0.336 OK ANG 83 -

Axial (P) 118,210 152,846 118,210 152.846 564,300 0.271 OK Shear(V) 19,079 24,669 19 079 24,669 338,600 0.073 OK Level D Moment (M) 1,374,266 1,776,926 1,374,266 1,776,926 5,078,000 0.350 OK "Level A and Level C are not affected by TSV Loads.

Forces (P), (V) and Moments (M)are in Lbs and in In-Lbs respectively.

Note 1: Penetration Flued Head Allowable are taken from Attachment 9, Pages 19 & 20 in Calc. No. C-0122 Rev. 9

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(D Project : Hope Creek Extended Power Uprate Calculation No.: C-0122 Rev. 9 Page Nos.: 38, 476 Calculation

Title:

Main Steam Line A', MSRV Lines A, J & R

Description:

SRV Piping - Line A Maximum Stress Intensities 0) 0)

• from GE Stress Report 23A6126 Rev. 1 dated 217/89 which is a part of Calculation No. C-0122 Rev. 9

• Maximum load for TSV load combination.

Note 1: Not affected by Power Uprate.

Note 2: Since Equation 10 is not satisfied, Equation 11 is used to satisfy the required criteria

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=r 3D Project: Hope Creek Extended Power Uprate Calculation No.: C-0122 Rev. 9 Page Nos.: 38, 459 Calculation

Title:

Main Steam Line 'A', MSRV Lines A, J& R

Description:

SRV Piping - Line A Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio: Ratio:

CODE SECTION IIIND-3600 Number Maximum Uprate Allowable Power Uprate Power Uprate Stress* Stress Stress to Original to Original Equation : 8 Sustained Loads 381 8,119 8,119 22,500 0.361 Note 1 0)

Equation : 9 Occasional 382** 7,005 7,404 27,000 0.274 Level C 344F 10,393 10,985 33,750 0.325 Level D 382 26,559 26,559 45,000 0.590 Note 1 Equation :10 Thermal + 354 20,622 20,622 22,500 0.917 Note 1 OBED Equation 11 Thermal + 354 25,341 25,341 37,500 0.676 Note 1 Sustained (Note 2)

• From Stress Report

• • Maximum load for TSV load combination.

Note 1: Not affected by Power Uprate. 0 I

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CD Project : Hope Creek Extended Power Uprate Calculation No.: C-0122 Rev. 9 Page Nos.: 41, 531 Calculation

Title:

Main Steam Line 'A', MSRV Lines A, J & R

Description:

SRV Piping - Line J Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio: Remarks CODE SECTION III ND-3600 Number Maximum Uprate Allowable Power Uprate Stress* Stress Stress to Original

--1 Equation : 8 Sustained Loads 183F 7,649 7,649 22,500 0.340 Note 1 I

Equation: 9 Occasional 152F 24,047 25,418 27,000 0.941 Level C 183F 9,118 9,638 33,750 0.286 Level D 152F 24,295 24,295 45,000 0.540 Note 1 Equation :10 Thermal + 180 21,672 21,672 22,500 0.963 Note 1 OBED Equation 11 Thermal + 180 26,167 26,167 37,500 0.698 Note 1 Sustained I_

from GE Stress Report 23A6126 Rev. 1 dated 217/89 which is a part of Calculation No. C-0122 Rev. 9 Note 1: Not affected by Power Uprate.

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0 CD Project : Hope Creek Extended Power Uprate Stress Report No.: 23A6126- Rev. 1 Page Nos.: 41, 504 Calculation Title Main Steam Piping & Equipment Loads

Description:

SRV Piping - Line J Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio: Remarks CODE SECTION IIIND-3600 Number Maximum Uprate Allowable Power Uprate Stress* Stress Stress to Original

0) Equation: 8 Sustained Loads 183F 7,649 7,649 22,500 0.340 Note 1 Co Equation: 9 Occasional 152F 24,047 25,418 27,000 0.941 Level C 183F 9,118 9,638 33,750 0.286 Level D 152F 24,295 24,295 45,000 0.540 Note 1 Equation :10 Thermal+ 130N 20,253 20,253 22,500 0.900 Note 1 OBED Equation 11 Thermal+ 130N 25,316 25,316 37,500 0.675 Note 1 Sustained (Note 2)

From Stress Report Note 1: Not affected by Power Uprate. 17-

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0 CD Project : Hope Creek Extended Power Uprate Calculation No.: C-0122 Rev. 9 Page Nos.: 42, 614 Calculation

Title:

Main Steam Line' A', MSRV Lines A, J & R

Description:

SRV Piping - Line R Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio: Remarks CODE SECTION III ND-3600 Number Maximum Uprate Allowable Power Uprate Stress* Stress Stress to Original Equation : 8 Sustained Loads 291 8,127 8,127 22,500 0.361 Note 1 0)

Equation : 9 Occasional 282N 23,517 24,857 27,000 0.921 Level C 248 10,975 11,601 33,750 0.344 Level D 282N 23,517 23,517 45,000 0.523 Note 1 Equation :10 Thermal+ 282N 28,118 28,118 22,500 1.250 Note 2 OBED Equation 11 Thermal + 282N 33,187 33,187 37,500 0.885 Note 1 Sustained

• from GE Stress Report 23A6126Rev. 1 dated 2/7/89 which is a part of Calculation No. C-0122 Rev. 9 Note 1: Not affected by Power Uprate.

Note 2: Since Equation 10 is not satisfied, Equation 11 is used to satisfy the required criteria CD,;0 C-=

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3 CD Project: Hope Creek Extended Power Uprate Calculation No.: C-0122 Rev. 9 Page Nos :42, 604, 614 Calculation

Title:

Main Steam Line' A', MSRV Lines A, J & R

Description:

SRV Piping - Line R Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio: Remarks CODE SECTION III ND-3600 Number Maximum Uprate Allowable Power Uprate Stress* Stress Stress to Original 0)

Equation: 8 Sustained Loads 291 8,127 8,127 22,500 0.361 Note 1 Equation: 9 Occasional 282N 23,517 24,857 27,000 0.921 Level C 248 10,975 11,601 33,750 0.344 Level D 282N 23,517 23,517 45,000 0.523 Note 1 Equation :10 Thermal + 262 22,857 22,857 22,500 1.016 Note 2 OBED Equation 11 Thermal + 262 27,615 27,615 37,500 0.736 Note 1 Sustained

• from GE Stress Report 23A6126 Rev. 1 dated 2/7189 which is a part of Calculation No. C-0122 Rev. 9 Note 1: Not affected by Power Uprate.

Note 2: Since Equation 10 is not satisfied, Equation 11 is used to satisfy the required criteria 017 A-22 CD0 CD0

• (0

CD Project: Hope Creek Extended Power Uprate Calculation No.: C-0122 Rev. 9 Page Nos.: 42, 603 Calculation

Title:

Main Steam Line' A', MSRV Lines A, J & R

Description:

SRV Piping - Line R Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio: Remarks CODE SECTION III ND-3600 Number Maximum Uprate Allowable Power Uprate Stress* Stress Stress to Original

0) Equation: 8 Sustained Loads 291 8,127 8,127 22,500 0.361 Note 1 Equation: 9 Occasional 282N 23,517 24,857 27,000 0.921 Level C 248 10,975 11,601 33,750 0.344 Level D 282N 23,517 23,517 45,000 0.523 Note 1 Equation :10 Thermal + 260N 22,474 22,474 22,500 0.999 Note I OBED Equation 11 Thermal + 260N 27,523 27,523 37,500 0.734 Note 1 Sustained from GE Stress Report 23A6126 Rev. 1 dated 217/89 which is a part of Calculation No. C-0122 Rev. 9 Note 1: Not affected by Power Uprate. r-0 I

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0 CD Project: Hope Creek Extended Power Uprate Calculation No.: C-0122 Rev. 9 Page Nos.: 42, 578 Calculation

Title:

Main Steam Line' A ', MSRV Lines A, J & R

Description:

SRV Piping - Line R Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio: Remarks CODE SECTION III ND-3600 Number Maximum Uprate Allowable Power Uprate Stress* Stress Stress to Original Equation: 8 Sustained Loads 291 8,127 8,127 22,500 0.361 Note 1 0)

Equation : 9 Occasional 282N 23,517 24,857 27,000 0.921 Level C 248 10,975 11,601 33,750 0.344 Level D 282N 23,517 23,517 45,000 0.523 Note 1 Equation :10 Thermal + 240N 13,050 13,050 22,500 0.580 Note 1 OBED Equation 11 Thermal + 240N 19,181 19,181 37,500 0.511 Note 1 Sustained

• from GE Stress Report 23A6126 Rev. 1 dated 2/7/89 which is a part of Calculation No. C-0122 Rev. 9 Note 1: Not affected by Power Uprate. I--

0 I

0 C-CD A-24 0 Co Co

0 CD Project: Hope Creek Extended Power Uprate Calculation No.: C-0122 Rev. 9 'Page Nos. 60 through 73 Calculation

Title:

Main Steam Line A, MSRV Lines A, J & R

Description:

SRV Lines A, J & R SRV Inlet & Outlet Flange Moments Loads Evaluation ASME Loading Existing Maximum of Ratio=

SRV Line Node Pt. Load Loading Combination EPU Loads Enveloped EPU & Allowable Maximum/ Remark Level with TSV Load Enveloped Allowable Loada SRVA 306 Level B MR 263,066 366,004 308,618 366,004 800,000 0.458 ( Note 2)

0) Inlet Level D MR 294,495 380,782 4h, _ 316,614 380,782 925,000 0.412 ( Note 2)

N) SRVA 310 Level B MR 214,050 276.767 219 956 276,767 600,000 0.461 (Note 2 Outlet LevelTU MR K 220,697 285,361 226!526 285,361 600,000 0.476 ote SRVJ 106 Level B MR 132,408 171,204 250,254 250,254 800,000 0.313 OK ( Note 1)

Inlet -Level D MR 136,742 176,807 250,967 250N967 925,000 0.271 0 OK Note 1)

SRVJ 110 Level B MR 116,972 151,245 192.318 192,318 600,000 0.321 OK (Note 1)

Outlet Level 0 MR 119.692 154,762 192,742 192,742 600,000 0.321 OK (Note 1)

SRV R 206 Level B MR 174,369 225,459 198,150 225,459 800,000 0,282 (Note 2)

Inlet Level D MR 174,359 225,459 198,150 225,459 925,000 0.244 (Note 2)

SRV R 210 Level B MR 147,491 190,706 166,171 190,706 600,000 0.318 ot Outlet Level U M ... 147,491 190,706 166,171 0 600,000 0318 ote2 17-Forces and Momenta are in Lbs end in In-Lbs respectively. I"

'from GE Stress Report 23A6126 Rev. 1 dated 2/7/89 which is a part of Calculation No. C-0122 Rev. 9 Note 1: EPU Loads are less than the Existing Enveloped Loads and are within allowable.

Note 2: The load increase due to TSV is greater than the existing enveloped loads. However, the increase is lower than the allowable. 01 (pO A-25 X 0 CD C:)

0 CD Hope Creek Extended Power Uprate 0 Project:

Calculation No.: C-0122 Rev. 9 -Page Nos, 47, 51, 52 and 53 Calculation

Title:

Main Steam Line' A', MSRV Lines A, J & R

Description:

SRV Lines A, J & R Valve Acceleration Loads Evaluation

. LodingExising Maximum of Node Pt. Description ASME Load Level Loading Combination Direction Loading Existing EPu &

EPU Loads Enveloped Enveloped Allowable Remarks with TSV Load Loads Al-v B 0.340 0.440 2.182 2.182 8.00 OK (Note 1)

LevelB -B___

306 SRVLINEA AV 0.322 0.416 0.554 0.554 6.00 OK (Note 1)

Al 0.398 0.514 2.192 2.192 8.00 OK (Note 1)

Level D AV 0.365 0.472 0.580 0.580 6.00 OK (Note 1)

Level B AN 0.328 0.424 1.502 1.502 8.00 OK (Note 1 )

108 SRV LINE AV 0.163 0.211 0.547 0.547 6.00 OK (Note 1) 06 0,379 0.490 1.514 1.514 8.00 OK (Note 1)

LevelD AH AV 0.193 0.250 0.550 0.550 6.00 OK (Note 1) oN1 Level B AI 0.350 0.453 1.616 1.616 8.00 OK (Note 1) 206 SRV LINE R - AV 0.227 0.294 0,565 0.565 6.00 OK (Note 1)

Level D AH 0.403 0.521 1.629 1.629 8.00 OK ( Note 1) i - AV 0.259 0.336 0.578 0.578 6.00 OK(Note1j I

Level A and Level C are not affected by TSV Loads.

Acceleration Loads are in G's

-from GE Stress Report 23A6126 Rav.1 dated 217189 which is a part of Calculation No. C-0122 Rev. 9 Note 1: EPU loads are less than Existing Enveloped Loads. Therefore, the structural Integrity of the SRV valves are not affected.

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CD Project: Hope Creek Extended Power Uprate Calculation No.: C-0121, Rev. 7 *Page No: 22, 35-37, 206, 363, 411 Calculation

Title:

Main Steam Line' B', MSRV Lines B, F, K& P

Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Ext. Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable EPU Stress* Stress Stress to Allowable (PSI) (PSI) (PSI)

Equation : 9 NormalVUpset 009 25,788 27,258 31,860 0.86 009 25,464 25,464 39,825 0.64 0') Emergency Faulted 009 26,030 27,514 53,100 0.52 Equation 10 066F 59,230 60,948 53,100 1.15 Equation :12 066F 39,985 39,985 53,100 0.75 Equation :13 066F 27,852 27,852 53,100 0.52 Equation :14 (Fatigue) CUF 200 0.08 0.081 0.100 0.81

• from GE Stress Report 23A6127 Rev. 1 dated 217189 which is a part of Calculation No. C-0121 Rev.7

• • Since equation 10 is not satisfied, the piping is qualified by equations 12 and 13.

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(D Project: Hope Creek Extended Power Uprate Calculation No.: C-0121, Rev. 7 *Page No: 22, 35, 36-37, 206, 367, 411 Calculation

Title:

Main Steam Line' B ', MSRV Lines B, F, K & P

Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Ext. Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable EPU Stress* Stress Stress to Allowable (PSI) (PSI) (PSI)

Equation : 9 Normal/Upset 009 25,788 27,258 31,860 0.86 02 Emergency 009 25,464 25,464 39,825 0.64

0) Faulted 009 26,030 27,514 53,100 0.52 Equation: 10

• 066F 58,074 59,758 53,100 1.13 Equation: 12 066F 39,482 39,482 53,100 0.74 Equation: 13 066F 26,343 26,343 53,100 0.50 Equation: 14 (Fatigue) CUF 200 0.08 0.081 0.100 0.81

• from GE Stress Report 23A6127 Rev. 1 dated 2/7/89 which is a part of Calculation No. C-0121 Rev.7

-Since equation 10 is not satisfied , the piping is qualified by equations 12 and 13.

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Project: Hope Creek Extended Power Uprate Calculation No.: C-0121, Rev. 7 *Page No: 22, 35, 36-37, 206, 323, 411 Calculation

Title:

Main Steam Line' B', MSRV Lines B, F, K & P

Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Ext. Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable EPU Stress* Stress Stress to Allowable

- (PSI) (PSI) (PSI)

-N Equation : 9 Normat/Upset 009 25,788 27,258 31,860 0.86

0) Emergency 009 25,464 25,464 39,825 0.64 N)

Co Faulted 009 26,030 27,514 53,100 0.52 Equation : 10 049F 49,787 51,231 53,100 0.96 Equation : 12 049F 31,374 31,374 53,100 0.59 Equation: 13 049F 23,451 23,451 53,100 0.44 Equation : 14 (Fatigue) CUF 200 0.08 0.081 0.100 0.81

• from GE Stress Report 23A6127 Rev. 1 dated 2/7189 which is a part of Calculation No. C-0121 Rev.7 0 I-1 1-C-"

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CD Project: Hope Creek Extended Power Uprate Calculation No.: C-0121, Rev. 7 *Page No: 22, 35-37, 206, 359,411 Calculation

Title:

Main Steam Line' B', MSRV Lines B, F, K & P

Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Ext. Power Code Ratio:

SECTION IIINB-3600 Number Maximum Uprate Allowable EPU Stress* Stress Stress to Allowable (PSI) (PSI) (PSI)

Equation : 9 Normal/Upset 009 25,788 27,258 31,860 0.86 0) bA Emergency 009 25,464 25,464 39,825 0.64 0

Faulted 009 26,030 27,514 53,100 0.52 Equation: i0 065 49,537 50,974 53,100 0,96 Equation: 12 065 31,492 31,492 53,100 0.59 Equation :13 065 25,590 25,590 53,100 0.48 Equation :14 (Fatigue) CUF 200 0.08 0.081 0.100 0.81

• from GE Stress Report 23A6127 Rev. 1 dated 2/7/89 which is a part of Calculation No. C-0121 Rev.7 0

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CD Project Hope Creek Extended Power Uprate Calculation No.: C-0121, Rev. 7 *Page No: 22. 35-37, 411,422 Calculation

Title:

Main Steam Line' B 1,MSRV Lines 8, F, K & P

Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Ext. Power Code Ratio:

SECTION IIINB-3600 Number Maximum Uprate Allowable EPU Stress* Stress Stress to Allowable

,_(PSI) (PSI) (PSI)

Equation: 9 Normal/Upset 009 25,788 27,258 31,860 0.86

0) Emergency 009 25,464 25,464 39,825 0.64 Faulted. 009 26,030 27,514 53,100 0.52 Equation: 10 300* 55,557 57,168 54,600 1.05 Equation: 12 300 23,090 23,090 54,600 0.42 Equation: 13 300 30,445 30,445 54,600 0.56 Equation: 14 (Fatigue) CUF 200 0.08 0,081 0.100 0.81

• from GE Stress Report 23A6127 Rev. 1 dated 2/7/89 which is a part of Calculation No. C-0121 Rev.7

-Since equation 10 is not satisfied, the piping is qualified by equations 12 and 13.

'Branch Connection Sweepolet 300 to 39 C-)

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Project Hope Creek Extended Power Uprate Calculation No.: C-0121, Rev. 7 "Page Nos: 92, 93 and 94 Calculation

Title:

Main Steam Line' B', MSRV Lines B, F, K & P

Description:

Main Steam Piping Line B - MSIV (Inboard) Bonnet Flange Loads Evaluation' Maximum oat ASME Load Loading Existing EPU & Allowable Ratio=

inboard MSIV Node Pt. Load Loading Combination EPU Loads Enveloped Envelpe (Note Maximum/r Remark with TSV Load oad Ilowable

__________ __________ ~Loads ___________

Level B Axial F 2,030 2,625 2,030 2,625 29,270 0.0897 OK (Note 2)

Moment 437,213 565,316 431,513 565.316 1,610,000 0.3511 OK (Note 2)

Level C Axial F 2,030 2,625 2,030 2,625 29,270 0.0897 OK (Note 2)

0) Inboard 074 -

Moment 437,213 565.316 437,213 565,316 1,610,000 0.3511 OK (Note 2)

Level D Axial F 2,222 2,873 2,344 2,873 29,270 0.0982 OK (Note 2)

Moment 477,212 617,035 551,025 617,035 1,610,000 0.3833 OK (Note 2)

Level A is not affected by TSV Loads.

from GE Stress Report 23A6127 Rev. I dated 2/7/89 which is part of Calculation No. C-0121 Rev.7 Forces and Moments are in Lbs and in In-Lbs respectively.

Note 1: Allowable are taken from GE Document 213A8411 Rev..

Note 2: The EPU loads are greater than the existing enveloped loads. However, the EPU loads are lower than the allowable.

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Title:

MainSteam Une'B', MSRV Unes 8. F. K& P

Description:

Main Steam Piping-Line B Inboard MSIV Inlet and OuLeltConnection Loads Evaluation Loading Existing Maximum of Ratio--

nboard MSI, Node Pl ASME Load Loading Combination EPU Loads Enveloped EPU & Allowable Maximum/All Remark Leval withTSV

_________ ~Leads_____

Load Leas wamble

- -ialtress 7,557 10JrO3.7327 7770 0.583 (Note I)

LevefB Tor. Stress 613 793 613 793 17,700 0.045 (Note 1 Bend. stress 10,305 13,402 10,365 13,402 17,700 0.757 (Note )

Axia Stress 7987 10.327 ,, 7,87 10r327 17.700 0.583 (Note I)

I~ Inlet 070 Level C Tor. Stress 613 793 613 793 717.00 0.045 Note 1 Connection Bend. stress 10,305 13,402 10,365 13,402 17r700 0.757 (Note 1) 0)

AxialStress 7,992 10,334 7,992 17700 0.584 (Note I1 Level C Tor. Stress 6689 865 740 865 17,70D 0.049 (Note I)

!x B'end s"ess 10,848 14,026 10,848 14,002 17,7o0 0.792 (Note I Axial S-re- 8,204 10,908 8,204 17700 0.598 (Note )

Level B Tor. Stress 407 528 407 526 17.700 0.030 (Note 1)

Bend. stress 6,213 0,033 6.407 8,033 17.700 0.454 (Note )

AxialStress 8 204 108608 t 17,700 0.599 Note 1 Outlt 078 Level C Tor. Stress 407 529 407 529 1 0.030 (Note 1 coanectlon Bend, stress 6,213 8,033 6,407 8,033 17,700 0.454 (Note 1 AxialStress 8,210 10,16 8,210 10,616 17700 0.600 (Note I1 Level C Tor. Stress 456 590 456 590 1 0.033 (Note 11 Bend. stress 1 6549 1468 6749 17,700 0.478 (Note 1

• LevelA Is not affecled by TSV Loads.

• ffrom GE Stress Report 23A6t27 Rev. I dated 2f7189which is a pert of Calculation atC-0121 Rea. 7. 0 Stresses are Inpsi Note 1:The EPU loads are greater than the existing enveloped loads. However, the EPU loads are lower than the allowable.

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Hope Creek Extended Power Uprate (1)

Project, Calculation No.: C-0121, Rev. 7 *Page Nos.: 50 through53 Calculation

Title:

Main Steam Line' B' MSRV Lines B, F, K & P

Description:

Main Steam Piping - Line B MSIV Acceleration Loads Evaluation Loading Existing Maximum of Node Pt Description ASME Load Level Loading Direction Combination ith iSo EPU U Loads Load EePU &

Enveloped Qualified by Test R Loads 0W43S 0.564 0.436 0.564 (Note 1) OK L AH 0555 0.551 0.555 (Note 1) OK 072 Body AH 0,491 0.635 0.856 0.856 (Note 1) OK Level D AV 0.532 0.687 0.659 0.687 (Note 1) OK AH 1.721 2.225 1.795 2.225 (Note 1) OK Level B

0) 076 Operator 076 AV 1 623 2.098 1.623 2.098 (Note 1) OK CA) 2.926 (Note 1) OK Level D AH 1.768 2.286 2.926 AV 1.938 2.506 1.938 2.506 (Note 1) OK AH 0.352 0.455 0.493 0.493 2.0 (Note 1) OK Level B 074 Bonnet Flange AV 0.589 0.761 0.657 0.761 3.0 (Note 1) OK Level D AH 0.395 0.510 0.886 0.886 4.5 (Note 1) OK

___AV 0.716 0.926 0.774 0.925 5.0 (Note 11 OK

'from GE Stress Report 23A6127 Rev.1 dated 217/89 which is a part of Calculation No. C-0121 Rev.7

• Acceleration Loads are in G's Note 1: NEDC 31020 - Hope Creek Environmental Qualification Report MSIV Actuator 0

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=30 Project: Hope Creek Extended Power Uprate Calculation No.: C-0121, Rev. 7 Page Nos. 47, 48 and 49 Calculation

Title:

Main Steam System Line 'B', MSRV Lines B, F, K & P

Description:

Main Steam Piping - Line B RPV Nozzle Loads

-- = -Ratio= =

ASME Load Loading Loading EPU Loads Allowable Maximum/ Remark ComponentLevel (Note 1) (Note 2) Combination ( Note 3 A o 1ith TSV I Allowable Secondary HR 74,999 96,974 674,200 0.144 OK (Level B)

Secondary MR 2,934,963 3,794,907 14,384,000 0.264 OK (Level B)

RPV 001----

Primary HR 49,482 63,980 374,520 0.171

0) OK (LevelD)m I1 e Prima( MR 1,955,944 2,529,036 7,417,000 (Level D)I 0.341 OK from GE Stress Report 23A6127 Rev. 1 dated 2/7/89 which is a part of Calculation No. C-0121 Rev.7 Forces and Moments are in Lbs and in In-Lbs respectively.

Note 1 Levels A and C are not affected by the TSV loads Note 2 For Secondary load level, HR and MR are the envelope of case 1 and case 3 HR and MR are calculated at the RPV I NOZZLE junction using given FX, FY, FZ, MX, MY, MZ load components.

Note 3 For the EPU loads, only the load combination with TSV load is assumed to be impacted.

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• 3 CDP Project: Hope Creek Extended Power Uprate Calculation No.: C-0121 Rev. 7 "Attachmnent 9. Page Nos. 20 and 21 Calculation

Title:

Main Steam Une' B ', MSRV Lines B, F, K & P

Description:

Main Steam Piping - Line B Penetration Flued Head Evaluation" ASME - d -Loading Existing Maximum of Anchor Node Pl. Load EoLiating EPU & Allowable Ratio = Remark Number Load DiLoadin Combinetion .PU Loads En',loped Enveloped (Note 1) Max/Aflowable Level with TSV Load ______ Loads Axial (P) 51.165 66,156 51,165 66.156 79.000 0.837 OK

~1 OBE+ Shear(VM 3.968 5,131 5,381 5,381 28,300 0.190 OK TSVC -

Moment (Mt 644,316 833,101 644,316 833,101 2,212,000 0.377 OK at Cs) ANC 84 -- -

Axial (P) 109,442 141 509 109,442 141,509 564,300 0.251 OK 0)

Shear (V) 29,706 38,410 31,550 38,410 338,600 0.113 OK Level 0D___ ____ ___ ___

Moment 3,280,370 4,241,518 3,347,297 4.241,518 5,078,00 0.835 OK

• Level A and Level C are not affected by TSV Loads.

Forces (P), (VMand Moments (M) are in Lbs and in In-Lbs respectively.

Note 1: Penetration Flued Head Allowable are taken from Attachment 9, Pages 20 & 21 In Catc. No. C-0121 Rev. 7 0

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CD Project: Hope Creek Extended Power Uprate Calculation No.: C-0121, Rev. 7 *Page Nos.: 23, 39 & 536 Calculation

Title:

Main Steam Line' B', MSRV Lines B, F, K & P

Description:

SRV Line B - Maximum Stress Intensities I

I 0)

I'

• from GE Stress Report 23A6127 Rev. 1 dated 2/7189 which is a part of Calculation No. C-0121 Rev. 7 I-Note 1: Not affected by, Power Uprate. -0

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CD Project: Hope Creek Extended Power Uprate Calculation No.: C-0121, Rev. 7 *Page Nos.: 23, 41, 567 & 581 Calculation

Title:

Main Steam Line' B', MSRV Lines B, F, K &P

Description:

SRV Line F - Maximum Stress Intensities CRITERIA PER ASME Node Original Ext. Power Code Ratio:

CODE SECTION IIIND-3600 Number Maximum Uprate Allowable EPU Remark Stress* Stress Stress to Allowable (Psi) (Psi) (Psi)

Equation: 8 Sustained Loads 457 7,648 7,648 22,500 0.340 Note 1 Equation: 9 Occasional 431 15,303 16,175 27,000 0.599 (0 Level C 431 15,303 16,175 33,750 0.479 Ij Level D 431 25,897 25,897 45,000 0.575 Note 1 Equation :10 Thermal + 420N 9,848 9,848 22,500 0.438 Note 1 OBED Equation 11 Thermal + 420N 15,287 15,287 37,500 0.408 Note 1 Sustained I

• from GE Stress Report 23A6127 Rev. 1 dated 217/89 which is a part of Calculation No. C-0121 Rev. 7 Note 1: Not affected by Power Uprate.

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Project : Hope Creek Extended Power Uprate Calculation No.: C-0121, Rev. 7 *Page Nos.: 23, 43, & 665 Calculation

Title:

Main Steam Une' B', MSRV Lines B, F, K & P

Description:

SRV Line K - Maximum Stress Intensities CRITERIA PER ASME Node Original Ext. Power Code Ratio:

CODE SECTION IIIND-3600 Number Maximum Uprate Allowable EPU Remark Stress* Stress Stress to Allowable (Psi) (Psi) (Psi)

Equation: 8 Sustained Loads 287 7,648 7,648 22,500 0.340 Note 1 Equation: 9 Occasional 220F 19,590 20,707 27,000 0.767 0)

Level C 257 10,509 11,108 33,750 0.329 I Level D 220F 19,590 19,590 45,000 0.435 Note I Equation :10 Thermal + 275N 12,172 12,172 22,500 0.541 Note 1 OBED Equation 11 Thermal + 275N 17,695 17,695 37,500 0.472 Note 1 Sustained I II

'from GE Stress Report 23A6127 Rev. 1 dated 217/89 which is a part of Calculation No. C-0121 Rev. 7 Note 1: Not affected by Power Uprate.

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(D Project : Hope Creek Extended Power Uprate Calculation No.: C-0121, Rev. 7 *Page Nos.: 23, 45, & 703 Calculation

Title:

Main Steam Line' B', MSRV Lines B, F, K & P

==

Description:==

SRV Line P - Maximum Stress Intensities CRITERIA PER ASME Node Original Ext. Power Code Ratio:

CODE SECTION III ND-3600 Number Maximum Uprate Allowable EPU Remark Stress* Stress Stress to Allowable (Psi) (Psi)

(Psi)

Equation : 8 Sustained Loads 193 7,648 7,648 22,500 0.340 Note 1 Equation: 9 Occasional 140 18,243 19,283 27,000 0.714 Level C 140 12,641 13,362 33,750 0.396 Level D 140 18,243 18,243 45,000 0.405 Note 1 Equation :10 Thermal + 125N 11,306 11,306 22,500 0.502 Note 1 OBED Equation 11 Thermal + 125N 16,275 16,275 37,500 0.434 Note 1 Sustained II

• from GE Stress Report 23A6127 Rev. 1 dated 217/89 which is a part of Calculation No. C-0121 Rev. 7 Note 1: Not affected by Power Uprate.

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Title:

Main Steam Line'B ', MSRV Unes B, F, K &P

Description:

SRV Inlet & Outlet Flange Moments Loads Evaluation ASME Loading Existing Maximum Ri=

SRV Une Node PL Load Loading Combination EPU Loads Enveloped EPUed Allowable Ma&imum/ Remark Level with TSV Load Allowable Load, 2

SRVP 103 Level B MR 143,215 185.177 271,662 271f66 800,000 0.340 OK (Note I Inlet Level D 146025 188,81D 272.990 272,990 925,000 0.295 OK ( Note I I SRVP 105 Level B MR 151,858 196,352 209,389 209,389 600.000 0.349 OK (Note 1) 0)

Outlet M¶ *5,5 198,93 ae~lD 2 a 210,262 600,000 0.350 OK ( Note 1)

SRVK 203 Level B MR 134,241 173.574 335,586 335,586 800,000 0.419 OK (Note 1 Inlet Level D MR 135,237 174,861 335.872 335,872 925.000 0.363 OK I Note 11 SRV K 205 LevelB MR 137,663 177,998 233.726 233,728 600,000 0.390 OK (Note 1)

Outlet LevelD MR 136.632 176,924 234.059 234,059 600,000 0.390 OK (Note 11 SRV B 303 Level B MR 449,499 551.202 4751150 561,202 800,000 0.727 (Note 2)

Inlet Level D MR 449,499 551,202 475,150 581,202 925,000 0.628 (Note 2 )

SRV B 305 Level s MR 350 598 453 325 360.761 453,325 0000 0.75 No 2 Outlet OR 350!99 453!325 360,862 453,325 600,000 0,755 3.2 SRV F 403 Level B MR 107.288 138,723 169,383 169,383 800,000 0.212 OK (Note 1I LevelD Fnlet MR 113,765 147,098 171.891 171,891 925000 0.186 OK (Nota 1I SRV F 405 Lavel B MR 94.276 121,89 15137,345 600000 0229 Outlet teveD *.TM 99,097 128,132 139.595 139,595 600,000 0.233 O J, 0

tfromGE Stress Report 23A6127 Rev.1 dated 217/89 which isa patl ofCalculatlon No. C-0121 Rev.7 Forces and Moments are in Lbs and in In-Lbs respectively.

Note 1: EPU Loads are less than the Existing Enveloped Loads and are within allowable.

Note 2: The EPU loads are greater than the existing enveloped loads. However, the EPU loads are lower than the allowable.

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CD Project Hope Creek Extended Power Uprste Calculation No.: C-0121, Rev- 7 **Page Nos. 50,54 through 57 Calculation

Title:

Main Steam Line'B', MSRV Lines B. F, K,& P

Description:

Main Steam Piping - Line B SRV Valve Acceleration Loads Evaluation*

of Maximum Nods Pt. Description ASME Load Loading Loading Existing EPU &

Direc Combination EPU Loads Enveloped Enveloped Allowable Remarks Levei with TSV Load lopd Loads Levei B AH 0.294 0,380 1.854 1.854 8.00 OK (Note 1) 103 SRVUNEP - AV 0.374 0.484 1.234 1.234 6.00 OK (Note 1 LevelD AH 0.350 0.452 1.863 1.863 8.00 OK (Note I) r AV 0.431 0.557 1.252 1.252 6.00 OK (Note 1)

LevelB AH 0.336 0.434 1.547 1.547 8.00 OK (Note )

203 SRVLINEK - AV 0.396 0.512 1.508 1.508 6.00 OK (Note I LeveiD AH 0.400 0.518 1.562 1.562 8.00 OK (Note )

AV 0.459 0.593 1.626 1.526 6.00 OK (Note 1 Level B AH 0.426 0.551 2.189 2.189 8.00 OK (Note 1) 0)

303 SRV UNE B AV 0.332 0.429 2.218 2.218 6.00 OK (Note 1 Level D AH 0.514 0.665 2.207 2.207 8.00 OK (Note 1 )

I it AV 0.395 0.511 2.228 2.228 6.00 OK (Note 1 Level B AH 0.449 0.581 1,703 1.703 8.00 OK (Note 1) 43 SRVUNEF 0.236 0.305 1.416 1.416 6.00 OK (Note 1 LevelD AH 0.544 0704 1.731 1.731 8.00 OK (Note 1)

AV 0.280 1.424 10362 1.424 600 OK (Note1 Level A and Level C are not affected by TSV Loads.

• from GE Stress Report 23AS127 Rev.1 dated 2/7189 which is a part of Calculation No. C-0121 Rev, 7 Acceleration Loads are in G's Note 1: EPU loads are less than Existing Enveloped Loads. Therefore, the structural Integrity of the SRV valves are not affected.

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CD Project: Hope Creek Extended Power Uprate Calculation No.: C-0138, Rev. 9 *Page No: 38, 40 & 217 Calculation

Title:

Main Steam Line' C', MSRV Lines C, E, G & L

Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable Power Uprate Stress* Stress Stress to Allowable (PSI) (PSl) (PSI)

Equation :9 Normal/Upset 008 24,554 25,954 31,860 0.81 o) Emergency 008 23,343 23,343 39,825 0.59 (31 008 25,132 26,565 53,100 0.50 Faulted Equation :10 042F*** 55,862 57,482 53,100 1.08 Equation 12 042F17 35,644 35,644 53,100 0.67 Equation :13 042F*** 27,097 27,097 53,100 0.51 Equation: 14 (Fatigue) CUF 100 0.05 0.051 0.100 OK from GE Stress Report 23A6128 Rev. 1 dated 217/89 which is a part of Calculation No. C-0138 Rev. 9 I-Since equation 10 is not satisfied, the piping is qualified by equations 12 and 13. C)

• • • thick elbow assumption.

X J17 CD A-52 ,4 o

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0 CD Project: Hope Creek Extended Power Uprate Calculation No.: C-0138, Rev. 9 *Page No: 38, 40 & 217 Calculation

Title:

Main Steam Line' C', MSRV Lines C, E, G & L

Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable Power Uprate Stress* Stress Stress to Allowable

_PSI) (PSI) (PSI)

Equation

• 9 Normal/Upset 008 24,554 25,954 31,860 0.81 0)

Emergency 008 23,343 23,343 39,825 0.59 0)

Faulted 008 25,132 26,565 53,100 0.50 Equation: 10 052F 53,792 55,352 53,100 1.04 Equation: 12 052F 34,555 34,555 53,100 0.65 Equation: 13 052F 27,097 27,097 53,100 0.51 Equation : 14 (Fatigue) CUF 100 0.05 0.051 0.100 OK from GE Stress Report 23A6128 Rev. 1 dated 2/7/89 which is a part of Calculation No. C-0138 Rev. 9 Since equation 10 is not satisfied, the piping is qualified by equations 12 and 13.

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(D Project: Hope Creek Extended Power Uprate Calculation No.: C-0138, Rev. 9 *Page No: 38, 40 & 217 Calculation

Title:

Main Steam Line' C', MSRV Lines C, E, G & L

Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable Power Uprate Stress* Stress Stress to Allowable (PSI) (PSI) (PSI)

0) Equation : 9 Normal/Upset 008 24,554 25,954 31,860 0.81 o4 Emergency 008 23,343 23,343 39,825 0.59 Faulted 008 25,132 26,565 53,100 0.50 Equation :10 042F*** 50,139 51,593 53,100 0.97 Equation: 12 042F*** 26,536 26,536 53,100 0.50 Equation :13 042F*** 27,453 27,453 53,100 0.52 Equation :14 (Fatigue) CUF 100 0.05 0.051 0.100 OK r-from GE Stress Report 23A6128 Rev. I dated 2/7/89 which is a part of Calculation No. C-0138 Rev. 9 thin elbow assumption. C) 0 F-CD A-54

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CO3 Project: Hope Creek Extended Power Uprate Calculation No.: C-0138, Rev. 9 -Page Nos.: 105, 106 and 107 Calculation

Title:

Main Steam Line' C', MSRV Lines C, E, G & L

Description:

Main Steam Piping-Line C MSIV (Inboard) Bonnet Flange Loads Evaluation*

Loading Existing Maximum o w Ratio=

Inboard MSIV Node Pt. ASME Load Loading Combination EPU Loads Enveloped Enped (Note1 Maximum/ Remark Level with TSV Load veloped (Note 1) Alowable Loads Level B Axial F 2,134 2,759 2,237 2,759 29,270 0.0943 (Note 2)

Moment 462,299 597,753 462,299 597,753 1,618,000 0.3694 (Note 2)

OD Level C Axial F 2,134 2,759 2,237 2,759 29.270 0-0943 (Note 2)

Inboard 056 Io Moment 462,299 597,753 462,299 597,753 1,618,000 0.3694 (Note 2)

Level D Axial F 2,281 2,949 2,433 2,949 29,270 0.1008 (Note 2)

Moment 525,121 678.981 586,106 678,981 1,618,000 0.4196 (Note 2)

Level A is not affected by TSV Loads.

Forces and Moments are in Lbs and in In-Lbs respectively.

• • from GE Stress Report 23A6128 Rev. 1 dated 27/89 which Is a part of Calculation No. C-01 38 Rev. 9 Note 1: Allowable are taken from GE Document 213A8411 Rev. 0 Note 2: The EPU loads are greater than the existing enveloped loads. However, the EPU loads are lower than the allowable.

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0 (17 Project. Hope Creek Extended Power Uprate, Calculation No.: C-0138. Rev. 9 -Page Nos.105, 108 through II1 Calculation lte: Main Steam Une' C ', MSRV Lines C. E. G & L Descrption: Main Steam Piping-Line C Inboard MSIV Inlet and Outlet Conneotion Loads Evaeation*

Loadig ri porg Maximum of R ASMELoading EPU & o RalM Inboard MSIV Node Pt. ALeeLo Loading Combination EPU Loads Enveloped Enveloped Allowable Maximum/ Remark 5

with TSV Load Loads Allowable oat

- - - *,bes 71790 10,72 7,790 10,072 17,700 0.5691 (Note-1)

Level B Tor. Stress 566 732 566 732 17,700 0.0413 (Note 1)

Bend. stress 9!436 12,201 9,436 12,201 17700 0.6893 (Noti 11 Axial Stress 7.790 10,072 7,790 10,072 17,700 0.5691 (Note 1)

Inlat 054 LevetC Tor. Stress s98 732 566 732 17,700 0.0413 (Note 1)

Connection sbess 09435Bend. 120201 01 17,700 0.6893 (Note 1)

Axial Stress 7,823 10,115 7,824 10.115 17.700 0.5715 (Note 1) 0.0 Level D Tor. Stress Bend. stress 613 793 661 783 17.700 0.0449 (Note 11 9e821 12,099 9, 821 12,699 I 17.700 0.7174 (Note 1)

AxialStress 8,293 10,723 8.293 10.723 17.700 0.8058 (Note 1)

Level B To(. Stess 366 496 386 499 17,700 0.0282 (Note 1)

Bend. stress 6,864 6,875 7,079 8.875 17,700 0.5014 (Note 1)

Ax8.lStrss 8,293 10,723 8,293 1 , 1770 0.6058 (Note1)

Outlet 058 Level C Tor. Stress 386 499 386 499 17,700 0-0282 -Note 1) connection Bend. stress 8,84 75 7079 8,875 17,700 0.5014 (Note 1)

Axial Stress 8,298 10,729 8,298 10,729 17,700 0.6062 (Note11 Level D Tot. Soes 409 529 409 528 17,70 1 0.0209 (Note11

_Bend. stress 7,184 8.289 7,438 9,289 1717,00r 1 0.5248 1(Note 11 Level A is not affected by TSV Loads.

Stresses am in psi 0

-trom GE Stress Report 23A6128 Rev. 1 dated 2J7/89which is a part of Calculation No. C-0138 Rev.g Note 1: The EPU loads ere greater then the existing enveloped loads. However, the EPU loads ere lower than the allowable.

I Cr-A-56 (DOC

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Title:

Main Steam Line' C', MSRV Lines C, E, G & L

Description:

Main Steam Piping - Line C MSIV Acceleration Loads Evaluation-Maximum of ASME Load Loading Loading Existing EPU & Qualified Node Pt. Description Direction Combination EPU Loads Enveloped Enveloped by Test Remarks with TSV Load Loads 0.459 0.593 0.459 0.593 (Note 1) OK Level B AH 055 AV 0.508 0.656 0.670 0.670 (Note 1) OK Level AH 0'511 0.661 1.135 1.135 (Note 1) OK AV 0,581 0.751 0.738 0.751 (Note 1) OK AH 1.758 2.273 1.798 2.273 (Note 1) OK Level

0) Operator - AV 1.800 2.327 1.800 2.327 (Note 1) OK 057 2.073 2.680 3-010 3.010 (Note 1) OK I Level D AH AV 2.110 2.728 2.110 2.728 (Note 1) OK Level B AH 0.364 0.470 0.364 0.470 2.0 (Note 1) OK Bonnet AV 0.675 0.872 0.771 0.872 3.0 (Note 1) OK 056 Flange AH E Level 0 AH 0.407 0.527 0.955 0.955 4.5 (Note 1) OK AV 0.772 0.998 0.857 0.998 5.0 (Note 1) OK

• from GE Stress Report 23A6128 Rev. 1 dated 217189 which is a part of Calculation No. C-0138 Rev. 9

'Acceleration Loads are in G's Note 1: NEDC 31020 - Hope Creek Environmental Qualification Report MSIV Actuator.

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Hope Creek Extended Power Uprete Calculation No.: C-0138, Rev. 9 Page Nos. 52, 53 and 54 Calculation

Title:

Main Steam Line' C', MSRV Lines C, E, G& L

Description:

Main Steam Piping - Une C RPV Nozzle Loads Evaluation ASME Load Loading Loading EPU Loads Ratio=

Combination Allowable Maximum/ Remark Level (Note 1) (Note 2) with TSV (_Note_3_ Allowable Secondary HR 79,982 103,417 674,200 0.153 OK (Level B)

Secondary MR 3,869,725 5,003,554 14,384,000 0.348 OK RPV RPV 001 001- (Level 6) 0)

Primary (Level D) HR 48,788 63,083 374,520 0.168 OK Primary (Level P ) MR 2,315,204 2,993,559 7,417,000 0.404 OK Forces and Moments are in Lbs and in In-Lbs respectively.

• from GE Stress Report 23A6128 Rev. 1 dated 217189 which is a part of calculation No, C-0138 Rev.9 Note 1 Levels A and C are not affected by the TSV loads Note 2: For Secondary load level, HR and MR are the envelope of case I and case 3 HR and MR are calculated at the RPV / NOZZLE junction using given FX, FY, FZ, MX, MY, MZ load components.

Note 3: For the EPU loads, only the load combination with TSV load is assumed to be Impacted.

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Project: Hope Creek Extended Power Uprate Calculation No.: C-0138 Rev. 9 -Attachment 9, Page Nos. 28 and 29 Calculation

Title:

Main Steam Une' C', MSRV Lines C, E, G & L

Description:

Main Steam Piping - Line C Penetration Rued Head Evaluation*

Maximum of

• • Loading Existing EPU & Allowable Ratio = Remark ASME .. Loading Anchor Node PL Load Combination EPU Loads Enveloped Enveloped (Note 1) Max/Aliowable Number Level Direction with TSV Load load (

_LjadL_

Axial (P) 55.446 71,692 55,446 71,692 79,000 0.907 OK I, 6,256 6,810 28,300 0.241 OK OBE + Shear L 51267 6,810

0) TSVC -

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Moment (M) 739,722 956461 739,722 956,461 2,212,000 0.432 OK ANC 62 -

Axial iP) 55,101 71246 55.101 71,246 564,300 0.126 OK Shear MV) 5,547 7,172 8,620 8.620 338,600 0.025 OK Level D Moment Ml 1,216,796 1,573,317 1,216.796 1,573.317 5,078,000 0.310 OK Level A and Level C are not affected by TSV Loads.

Forces (P), (V) and Moments (M) are in Lbs and in In-Lbs respectively.

Note 1: Penetration Flued Head Allowable are taken from Attachment 9, Pages 28 & 29 in Calc. No. C-0138 Rev. 9 I-Cor--

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Title:

Main Steam Line 'C', MSRV Lines C, E, G & L

Description:

SRV Piping - Line C Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

CODE SECTION III ND-3600 Number Maximum Uprate Allowable Power Uprate Remark Stress* Stress Stress to Original Equation : 8 Sustained Loads 311F 7,974 7,974 22,500 0.354 Note 1

0) Equation : 9 Occasional 348F 19,772 20,899 27,000 0.774 Level C 335F 17,878 18,897 33,750 0.560 Level D 361N 20,971 20,971 45,000 0.466 Note 1 Equation :10 Thermal + 361F- 13,439 13,439 22,500 0.597 Note 1 OBED I Equation 11 Thermal+ 361F 18,906 18,906 37.500 0.504 Note2 Sustained J From GE Stress Report 23A6128 Rev. 1 dated 217/89 which is a part of Calculation No. C-0138 Rev. 9 Note 1: Not affected by Power Uprate.

Note 2: Equation 10 satisfies the required criteria. Equation 11 is not required.

• • Stresses at all other node points of SRV Line C is less than this value I-0 6

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0 CD W4 Project: Hope Creek Extended Power Uprate Calculation No.: C-0138, Rev. 9 *Page Nos.: 44 & 599 Calculation

Title:

Main Steam Line 'C', MSRV Lines C, E, G & L

Description:

SRV Piping - Line E Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

CODE SECTION III ND-3600 Number Maximum Uprate Allowable Power Uprate Remark Stress* Stress Stress to Original Equation 8 Sustained Loads 133N 10,626 10,626 22,500 0.472 Note 1

0) Equation :9 Occasional 11OF 24,589 25,991 27,000 0.963 c,1

(~7~ Level C 117N 19,130 20,220 33,760 0.599 Level D 11OF 25,204 25,204 45,000 0.560 Note 1 Equation :10 Thermal + 107N 15,221 15,221 22,500 0.676 Note 1 OBED Equation 11 Thermal + 107N 20,668 20,668 37,500 0.551 Note 2 Sustained

• From GE Stress Report 23A6128 Rev. 1 dated 2/7189 which is a part of Calculation No. C-0138 Rev. 9 Note 1: Not affected by Power Uprate.

Note 2: Since Equation 10 is satisfied, Equation 11 is not required. I--

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CD Project : Hope Creek Extended Power Uprate Calculation No.: C-0138, Rev. 9 *Page Nos.: 46 & 735 Calculation

Title:

Main Steam Line 'C', MSRV Lines C, E, G & L

Description:

SRV Piping - Line G Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

CODE SECTION III ND-3600 Number Maximum Uprate Allowable Power Uprate Remark Stress* Stress Stress to Original Equation : 8 Sustained Loads 424N 8,138 iI 8,138 22,500 0.362 Note 1 I

Equation: 9 Occasional 406 24,102 25,476 27,000 0.944

0) Level C 406 21,176 22,383 33,750 0,663 Level D 406 24,102 24,102 45,000 0.536 Note 1 Equation :10 Thermal + 466F 16,298 16,298 22,500 0.724 Note 1 OBED Equation 11 Thermal + 466F 21,140 21,140 37,500 0.564 Note 2 Sustained From GE Stress Report 23A6128 Rev. I dated 217/89 which is a part of Calculation No. C-0138 Rev. 9 Note 1: Not affected by Power Uprate.

Note 2: Since Equation 10 is satisfied, Equation 11 is not required.

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CD Project: Hope Creek Extended Power Uprate Calculation No.: C-0138, Rev. 9 *Page Nos.: 48 & 762 Calculation

Title:

Main Steam Line 'C ', MSRV Lines C, E, G & L

Description:

SRV Piping - Line L Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

CODE SECTION III ND-3600 Number Maximum Uprate Allowable Power Uprate Remark Stress* Stress Stress to Original Equation: 8 Sustained Loads 236F 7,845 7,845 22,500 0.349 Note 1

0) Equation : 9 Occasional 218N 21,240 22,451 27,000 0.832 C,,

-.4 Level C 213F 20,137 21,285 33,750 0.631 Level D 218N 23,066 23,066 45,000 0.513 Note 1 Equation :10 Thermal + 218N 15,613 15,613 22,500 0.694 Note 1 OBED Equation 11 Thermal + 218N 20,850 20,850 37,500 0.556 Note 2 Sustained

• From GE Stress Report 23A6128 Rev. 1 dated 2/7/89 which is a part of Calculation No. C-0138 Rev. 9 Note 1: Not affected by Power Uprate.

Note 2: Since Equation 10 is satisfied, Equation 11 is not required. C) 0lr-C-0 A-66 -4 CD C0

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0 CD Project, Hope Creek Extended Power Uprate Celculabon No.: C-O138, Rev. 9 Page Nos. 71, 88 through 104 Calculation

Title:

Main Steam Line C ', MSRV Lines C, E, G &L Deucription: SRV Inlet & Outlet Flange Moments LoWdeEvaluation ASME Loading Existing Maximum of Ratio- l Rmr SRV Line Node Pt. Load Loading Combination EPU Loeds Enveloped EnvlPUed loabeMxm Level with TSV Enveloped Allowable Loads 0

SRV E 103 Level B MR 253,041 327,182 385,513 385.513 800 00 0.482 OK I Note )

Inlet Level D MR 257.845 333,394 8 387.024 925,000 0.418 OK (Note 1 I 2

SRV E 105 Level B MR 247.310 31977 325,626 325.626 500.000 0.543 OK I Note 1 OUtlet Level I) M 250.337 323,886 327,426 327.428 600,000 0.540 OK ( Note 1 SRVL 203 Level B MR 242938 314,119 275.105 314,119 800.080 0.393 (Note2)

Inlet -Level D -R 242.938 314,119 275,105 314,l119 925,000 0.340 (Note 2) 0 0)

I SRVL 205 Level B MR 221,024 2385784 235,318 285,784 600,00 0.476 (Note 2)

Outlet LeTve MR 221,024 288,784 235,318 285,784 600,000 0.476 JNote 2)

SRV C 303 Level B MR 205,206 265,331 265,617 265,617 800,000 0.332 OK (Note 11)

Inlet Level i MR 205,206 265,331 265,617 2656,617 9250Q0 0.287 OK (Note I)

SRV C 305 Level B MR 197,291 255097 2401 15 255.097 800000 0.425 . Note 2 )

outlet Level 1 - UT 197.291 255!097 240!115 255,097 600!000 0.425- (Note 2 )

SRVG 403 Level 8 MR 168.364 217,695 231,486 231,486 800,000 0.289 OK (Note 1)

Inlet - Level D 168,364 217.690 231,4 231,486 825,000 0.250 OK (Note 1)

SRV B 405 Level B MR 177,420 2291404 206.545 229,404 600,000 0.382 Note 2 outlet LeveltD rtM 177,420 229.404 206,545 229,404 800 000 0.382 Note 2

-from GE Stress Report 3Ael28 Rev I dated 2 ta pert of Calculation No. C-0138 Rev.9 0m9twhich Forces end Moments am in Lbs and InIn-Lbs respectively.

Note 1: EPU Loads ere lles then the Existing Enveloped Loads and ere within allowable. I-Note 2: The EPU loads am greater than the existing enveloped loads. However, the EPU loads are lower than the allowable. C-)

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Title:

Main Steam Line' C %,MSRV Lines C, E, G &L

Description:

SRV Valve Acceleration Loads Evaluation

• AMLod Ldng Loading Existing Maximum of M~mmo Node Pt.NoeP.Dsrpin Descriptio Level ASME Load Direction Loading wtTVLod LCombination EPU Loads Enveloped Enveloped EPU & Allowable Remarks with TSV Load Loads Level AlAH 1.311 1.695 1.984 1.984 8.00 OK (Note 1) 103 SRV LINE E AV 0.540 0.699 1-105 1.105 6.00 OK (Note 1)

Level D AH 1.311 1.695 1.984 1.984 8.00 OK (Note I)

-AV 0.540 0.699 1.105 1.105 6.00 OK (Note 1I Level B AH 1,300 1.681 2.306 2.306 8.00 OK (Note 1) 203 SRV LINE L - AV 0.515 0.666 1.214 1.214 6.00 OK (Note 1)

0) LevelD AH 1.300 1.681 2.306 2.306 8.00 OK (Note I) 01 AV 0.515 0,666 1.214 1.214 6.00 OK (Note CI Level AlAH 1.213 1.568 2.051 2.051 8.00 OK (Note )

303 SRV LINE C AV 0.386 0.499 (Note I1) 1.484 1.484 6.00 OK LeLel

- AH 1.213 1.568 2.051 2.051 8.00 OK (Note Level D - __________

AV 0.404 0.523 1.489 1,489 6.00 OK (Note 1 LevelB AH 0.912 1179 2.008 2.008 8.00 OK ( Note 1) 403 SRV LINEG AV 0.247 0.320 1110 1110 6.00 OK (Note, Level 0 Al 0.912 1.179 2008 2008 8.00 OK (Note 1)

AV 0.288 0.372 1.119 1119 6.00 OK (Note1l "Level A and Level C are not affected by TSV Loads.

Acceleration Loads are in G's

-from GE Stress Report 23A6128 Rev. I dated 217189 which Is a part of Calculation No. C-0138 Rev. 9 Note 1: EPU loads are less than Existing Enveloped Loads and allowable. Therefore, the structural integrity of the SRVs is not affected.

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Calculation No.: C-0138, Rev. 9 *Page No: 50 & 819 Calculation

Title:

Main Steam Line ' C', MSRV Lines C, E, G & L

Description:

HPCI Piping at Main Steam Line C Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable Power Uprate Stress* Stress Stress to Allowable (PSI) (PSI) (PSI)

Equation : 9 Normal/Upset 504F 19,066 20,153 31,860 0.63 Emergency 529F 8,553 8,553 39,825 0.21

0) Faulted 502 21045 22,245 53,100 0.42 Equation :10 502 47,378 48,752 53,100 0.92 Equation:12 502 15,898 15,898 53,100 0.30 Equation:13 502 20,174 20,174 53,100 0.38 Equation: 14 (Fatigue) CUF 529N 0.02 0,020 0.100 OK from GE Stress Report 23A6128 Rev. 1 dated 2/7/89 which is a part of Calculation No. C-01 38 Rev. 9 Equation 10 is satisfied. Equations 12 and 13 are not required.

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(D Project: Hope Creek Extended Power Uprate Calculation No.: C-0139 Rev. 8 *Page No: 22, 33 through 36, 283, and 398 Calculation

Title:

Main Steam Line' D', MSRV Lines M, D & H

Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable Power Uprate Remark Stress* Stress Stress to Allowable (PSI) (PSI) (PSI)

Equation : 9 Normal/Upset 009 21,023 22,221 31,860 0.697 OK 0)

0) Emergency 009 20,014 20,014 39,825 0.503 OK 009 21,562 22,791 53,100 0.429 OK Fauled Equation :10 039F 70,347 72,387 53,100 1.363 Note 1 Equation 12 039F 51,854 51,854 53,100 0.977 Note 2 Equation 13 039F 22,756 22,756 53,100 0.429 Note 2 Equation :14 (Fatigue) CUF 300 0.096 0.097 0.100 0.974 OK

• from GE Stress Report 23A6129 Rev.1 dated 2/7/89 which is a part of Calculation No. C-01 39 Rev. 8 17-Note 1: Since equation 10 is not satisfied, the piping is qualified by equations 12 and 13.

Note 2: Pipe Breaks have been postulated at NP-39 per CaIc. No. SC-44 Rev.1

Attachment:

A1 Sheet 1 of 2 Cn ;

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(D Project: Hope Creek Extended Power Uprate Calculation No.: C-0139 Rev. 8 *Page No: 22, 33 through 36, 334 and 398 Calculation

Title:

Main Steam Line' D', MSRV Lines M, D & H

Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable Power Uprate Remark Stress* Stress Stress to Allowable

.... _(PSI) (PSI) (PSI)

Equation : 9 Normal/Upset 009 21,023 22,221 31,860 0.697 OK I)

0) Emergency 009 20,014 20,014 39,825 0.503 OK Faulted 009 21,562 22,791 53,100 0.429 OK Equation 10 051F 54,840 56,430 53,100 1.063 Note 1 Equation :12 051F 35,973 35,973 53,100 0.677 Equation :13 051F 22,351 22,351 53,100 0.421 Equation : 14 (Fatigue) CUF 300 0.096 0.097 0.100 0.974 OK

• from GE Stress Report 23A6129 Rev.1 dated 217/89 which is a part of Calculation No. C-0139 Rev. 8 Note 1: Since equation 10 is not satisfied, the piping is qualified by equations 12 and 13.

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=F Project: Hope Creek Extended Power Uprate Calculation No.: C-0139 Rev. 8 *Page No: 22, 33 through 36, 328 and 398 Calculation

Title:

Main Steam Line' D', MSRV Lines M, D & H

Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

SECTION IIINB-3600 Number Maximum Uprate Allowable Power Uprate Remark Stress* Stress Stress to Allowable (PSI) (PSI) (PSI)

Equation: 9 Normal/Upset 009 21,023 22,221 31,860 0.697 OK Emergency 009 20,014 20,014 39,825 0.503 OK 0)

Faulted 009 21,562 22,791 53,100 0.429 OK 0)

U1, Equation : 10 050 44,068 45,346 53,100 0.854 OK Equation: 12 050 25,012 25,012 53,100 0.471 Equation :13 050 25,888 25,888 53,100 0.458 Equation :14 (Fatigue) CUF 300 0.096 0.097 0.100 0.974 OK

• from GE Stress Report 23A6129 rev. I dated 217189 which is a part of Calculation No. C-0139 Rev. 8 r-C)

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Project: Hope Creek Extended Power Uprate Calculation No.: C-0139 Rev. 8 *Page No: 22, 33 through 36, 247 and 398 Calculation

Title:

Main Steam Line' D', MSRV Lines M; D & H

Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable Power Uprate Remark Stress* Stress Stress to Allowable (PSI) (PSI) (PSI)

Equation :9 Normal/Upset 009 21,023 22,221 31,860 0.697 OK Emergency 009 20,014 20,014 39,825 0.503 OK 0)

Faulted 009 21,562 22,791 53,100 0.429 OK 0)

Equation :10 021N 47,227 48,597 53,100 0.915 OK Equation :12 021N 26,305 26,305 53,100 0.495 Equation :13 021N 22,170 22,170 53,100 0.418 Equation: 14 (Fatigue) CUF 300 0.096 0.097 0.100 0.974 OK

• from GE Stress Report 23A6129 rev. 1 dated 2/7/89 which is a part of Calculation No. C-0139 Rev. 8 r-)

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Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable Power Uprate Remark Stress* Stress Stress to Allowable

_PSI) (PSI) (PSI) 0h Equation: 9 Normal/Upset 009 21,023 22,221 31,860 0.697 OK Emergency 009 20,014 20,014 39,825 0.503 OK 4)

Faulted 009 21,562 22,791 53,100 0.429 OK Equation :10 200 47,739 49,123 54,600 0.900 OK Equation :12 200 20,951 20,951 54,600 0.384 Equation :13 200 27,995 27,995 54,600 0.513 Equation: 14 (Fatigue) CUF 300 0.096 0.097 0.100 0.974 OK

• from GE Stress Report 23A6129 rev. 1 dated 2/7/89 which is a part of Calculation No. C-0139 Rev. 8

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Project: Hope Creek Extended Power Uprate Calculation No.: C-0139 Rev. 8 *Page No: 22, 33 through 36, 396 and 398 Calculation

Title:

Main Steam Line' D', MSRV Lines M, D & H

Description:

Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

SECTION III NB-3600 Number Maximum Uprate Allowable Power Uprate Remark Stress* Stress Stress to Allowable (PSI) (PSI) (PSI)

Equation : 9 Normal/Upset 009 21,023 22,221 31,860 0.697 OK Emergency 009 20,014 20,014 39,825 0.503 OK 0) 0I cI Faulted 009 21,562 22,791 53,100 0.429 OK Equation: 10 300 64,131 65,991 54,600 1.209 Note 1 Equation :12 300 28,701 28,701 54,600 0.526 Equation :13 300 26,990 26,990 54,600 0.494 Equation: 14 (Fatigue) CUF 300 0.096 0.097 0.100 0.974 OK

• from GE Stress Report 23A6129 rev. 1 dated 2/7/89 which is a part of Calculation No. C-0139 Rev. 8 Note 1: Since equation 10 is not satisfied, the piping is qualified by equations 12 and 13.

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=r Project: Hope Creek Extended Power Uprate Calculation No.: C-0139 Rev. 8 -Pages 85, 86 and 87 Calculation

Title:

Main Steam Une ' D', MSRV Unes M, 0 & H

Description:

Main Steam Piping -Line D MSIV (Inboard) Bonnet Flange Loads Evaluation" Maximume or ASME Load Loading Existing EPU & Ratio =

Inboard MSIV Node Pt. Load Loading Combination EPU Loads Enveloped EPU & Allowable Maximum/ Remark with TSV Load Enveloped Allowable

_____ Leads ______

Level B Axial F 2,209 2,856 2,209 2,856 29,000 0.0985 OK (Note 1) 0Y) Moment 246,125 318,240 246,125 318,240 1.610,000 0.1977 OK (Note 1)

(0 059 Level C Axial F 2,209 2,856 2,209 2,856 29,000 0.0985 OK (Note 1) i Inboard __________

Moment 246,125 318,240 246,125 318,240 1,610,000 0.1977 OK (Note 1)

Level D Axial F 2,377 3,073 3,040 3,073 29,000 0.1060 OK (Note 1)

Moment 291,610 377,052 325,731 377,052 1,610,000 02342 OK (Note 1)

Level A is not affected by TSV Loads.

Forces and Moments are InLbs and In In-Lbs respectively.

-from GE Stress Report 23A6129 Rev. I dated 2J7189 which Is a part of Calculation No. C-0139 Rev. 8 Note 1: The EPU load is greater than the existing enveloped loads. However, the EPU load Is lower than the allowable.

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Project: Hope Creek Extended Power Uprate Calculation No.: C-0139 Rev. 8 -Pages 85, 88 through 91 Calculation

Title:

Main Steam Line' D',MSRV Lines M, D &H

Description:

Main Steam Piping Line D Inboard MSIV Inlet and Outlet Connection Loads Evaluation*

Loading Existing Maximum of Ratio tnboard MSIV Node Pt ASME Load Loading Combination EPU Lads Ewable Maximum I Remark Leve with TSV Load Enveloped Allowable Loads

-altess 7,853 10,154 = 107154 , 7 0.5737 (Note I Level B Tor. Stress 661 855 661 855 17,700 0.0483 ( Note I Bend. stress 6,345 8,204 6,345 8,204 17,700 0.4635 (Note I Axial Stress 7,853 10,154 7,853 10,154 17,700 0.5737 (Note 1)

Inlet 055 Level C Tor. Stress 661 855 661 855 17,700 0.0483 Note I Connection Bend. stress 6,345 8,204 6,345 8,204 17,700 0.4635 ( Note 1)

I Axial Stress 7,884 10,194 7,884 10,194 17,700 0.5759 I Note 1) 04 Level D Tor. Stress 697 931 718 901 17,700 0.0509 (Note 1 Bend. stress 6,743 8.719 6,743 8,719 17,700 0.4925 (Note 1) 0 Axial Stress 7,953 10,283 7,953 10,283 17.700 0.5810 ( Note 1)

Level B Tor. Stress 66t 855 561 855 17,700 0.0483 (Note 1 )

Bend. stress 4,597 5,944 4,597 5.944 17,700 0.3358 (Note 1)

Axial Stress 7,953 10,283 7,953 10,283 17,700 0.5810 I Note 1)

Outlet 063 Level C Tor. Stress 661 855 501 855 17,700 0.0483 (Note 1)

Connection Bend. stress 4,597 5,944 4,597 5,944 17,700 0.3358 (Note 1)

Axial Stress 7.968 1003 7988 1303 17.,700 0.5821 1 Note 1)

Level D Tor. Stress 695 899 730 899 17,700 0.0508 (Note 1 Bend. stress 4,871 6298 6,298 17,700 0.3558 (Note 1 Level A Is not affected by TSV Loads.

Stresses are in psi I-

"from GE Stress Report 23A6129 Rev. I dated 217/89 which Is a part of Calculation No, C-0139 Rev. 8 0 Note 1: The EPU loads are greater than the existing enveloped loads. However, the EPU loads are lower than the allowable.

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CD CA3 Project: Hope Creek Extended Power Uprate Calculation No,: C-01 39 Rev. 8 Page Nos.: 46 through 49 Calculation

Title:

Main Steam Line' D', MSRV Lines M, D & H

Description:

Main Steam Piping. Une D MSIV Acceleration Loads Evaluation-Loading E Maximum of Node Pt. Description ASME Load Level Loading din CombinLoadg Combination EPU Loads En Envelopedg EPU &

Enveloped Qualified by Test Remarks with TSV* Load* ads AH 0.419 0.542 0.419 0.542 (Note 1) OK LevelB Body AV 0.191 0.247 0.249 0.249 (Note 1) OK

0) 057 AH 0.452 0.584 0.567 0.584 (Note 1) OK

-1 Level 0 I AV 0.265 0.343 0.580 0.580 (Note 1) OK LevelB AH 1.050 1.357 1.050 1.357 (Note 1) OK 061 Operator _ AV 0.577 0.746 0.577 0.746 (Note 1) OK AH 1.211 1.566 1.587 1.587 (Note 1) OK LevelD AV 0,824 1.066 0.862 1.066 (Note 1) OK Level B______ AH 0.404 0.523 0.494 0.523 2.0 (Note 1) OK 059 Bonnet AV 0.224 0.290 0.248 0.290 3.0 (Note 1) OK Flange Level D AH 0.483 0.625 0.955 0.955 4.5 (Note 1) OK AV 0.248 0.320 0.326 0.326 5.0 (Note 1) OK

'from GE Stress Report 23A6129 Rev. 1 dated 2/71O9 which is a part of Calculation No.: C-0139 Rev. 8 Acceteration Loads are in G's Note 1: NEDC 31020 - Hope Creek Environmental Qualification Report MSIVActuator. 0 or-A-84 CD 0:

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CD Project lope Creek, Extended Power Uprate Calculation No.: C-0139 Rev. 8 *Page Nos. 43, 44 and 45 Calculation

Title:

Main Steam Line' D', MSRV Lines M,D & H

Description:

Main Steam Line 0 RPV Nozzle Loads Evaluation Loading Loading EPU Loads Ratio = EPU ASME Load Note 2) mbination (Not Allowable Loads/ Remark (Note 1) (

ComontNwitLevel h TSV e 3) Allowable Secondary HR 85,012 109,921 674,200 0.163 OK (Level B)

Secondary MR 3,433,638 4,439,694 14,384,000 0.309 OK RPV(Level B) I01 IQ Primary HR 44,237 57,198 374,520 0.153 (Level D) OK I\ Primary (Level 0) MR 1,899,131 2,455,576 7.417,000 0.331 OK Forces and Moments are in Lbs and in In-Lbs respectively.

• from GE Stress Report 23A6129 Rev. 1 dated 217189 which is a part of Calculation No.: C-0139 Rev. 8 Note 1: Levels A and C are not affected by the TSV loads.

Note 2: For Secondary load level, HR and MR are the envelope of case 1 and case 3 HR and MR are calculated at the RPV I NOZZLE junction using given FX, FY, FZ, MX, MY, MZ load components at safe end.

Note 3: For the EPU loads, only the load combination with TSV load is assumed to be impacted.

_17 A-85 CD 0D

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Project Hope Creek Extended Power Uprata Calculation No.: C-0139 Rev. 8 -Attachment 9, Page Nos. 24 and 25 Calculation

Title:

Main Steam Line' D', MSRV Lines M, D & H

Description:

Main Steam Piping - Line D Penetration Flued Head Evaluation*

ASME _*Loading Existing Maximum of Anchor Node PL Load Loading Combination EPU Ld Enloped EPU & Allowable Ratio = Remark Number Level Direction with TSV Envelopead Enveloped (Note 1) Max/Allowable (Note 1) Loads Axial (P) 53,112 68,674 53,112 68,674 79,000 0.869 OK OBE + Shear(V) 3,762 4,864 3.762 4,864 28,300 0.172 OK TSVC

0) Moment(Ml 266,105 344,074 266,105 344,074 2,212,000 0.156 OK

-.4 IA ANC 67 Axial (P) 5 55,077 71,215 4 0.126 OK Shear (V) 12,647 16,353 12,647 16,353 338,600 0.048 OK Level l Moment (M) 770,208 995,879 770,208 995,879 5,076T000 0196 OK

• Level A and Level C are not affected by TSV Loads.

Forces (P), (V) and Moments (M)are In Lbs and in In-Lbs respectively.

Note 1: Penetration Flued Head Loads and Allowable are taken from Attachment 9, Pages 24 & 25 in Caic. No. C-0139 Rev. 8 C:)

(0 A-88 CD (DO

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0I 6-oz A-89 CD CD 6

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0 CD Project: Hope Creek Extended Power Uprate Calculation No.: C-0139 Rev. 8 *Page Nos.:23, 37 & 472 Calculation

Title:

Main Steam Line 'D', MSRV Lines M, D & H

Description:

SRV Piping - Line D Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

CODE SECTION IIIND-3600 Number Maximum Uprate Allowable Power Uprate Remark Stress* Stress Stress to Original

-A4 Equation: 8 Sustained Loads 282 7,727 7,727 22,500 0.343 Note 1 Equation: 9 Occasional 236F 20,909 22,101 27,000 0.819 41 Level C 236F 15,894 16,800 33,750 0.498 Level D 236F 20,309 20,309 45,000 0.451 Note 1 Equation :10 Thermal + 262N- 20,685 20,685 22,500 0.919 Note 1 OBED Equation 11 Thermal + 262N 25,495 25,495 37,500 0.680 Note 1 Sustained from GE Stress Report 23A6129 Rev. 1 dated 2/7/89 which is a part of Calculation No.: C-0139 Rev. 8 Note 1: Not affected by Power Uprate.

" Stresses at all other node points of SRV Line D are less than this value 017 1-.

A-90 CD C0 co

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=r Project : Hope Creek Extended Power Uprate Calculation No.: C-0139 Rev. 8 *Page Nos.:23, 37 & 472 Calculation

Title:

Main Steam Line 'D', MSRV Lines M, D & H

Description:

SRV Piping - Line D Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

CODE SECTION IIIND-3600 Number Maximum Uprate Allowable Power Uprate Remark Stress* Stress Stress to Original 0)

Equation : 8 Sustained Loads 282 7.727 7,727 22,500 0.343 Note 1 0)

Equation : 9 Occasional 236F 20,909 22,101 27,000 0.819 Level C 236F 15,894 16,800 33,750 0.498 Level D 236F 20,309 20,309 45,000 0.451 Note I Equation :10 Thermal + 262N** 20,685 20,685 22,500 0.919 Note 1 OBED Equation 11 Thermal + 262N 25,495 25,495 37,500 0.680 Note 1 Sustained , I I from GE Stress Report 23A6129 Rev. 1 dated 2/7M89 which is a part of Calculation No.: C-0139 Rev. 8 Note 1: Not affected by Power Uprate. C) -

Stresses at all other node points of SRV Line D are less than this value C~)

A-91 CD 0)

(0 CO

(D Project: Hope Creek Extended Power Uprate Calculation No.: C-0139 Rev. 8 *Page Nos.:23, 37 & 472 Calculation

Title:

Main Steam Line 'D ', MSRV Lines M, D & H

==

Description:==

SRV Piping - Line H Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

CODE SECTION III ND-3600 Number Maximum Uprate Allowable Power Uprate Remark Stress* Stress Stress to Original Equation : 8 Sustained Loads 376 7,799 7,799 22,500 0.347 Note I Equation : 9 Occasional 336 15,471 16,353 27,000 0.606 Level C 334 11,670 12,335 33,750 0.365 Level D 336 15,471 15,471 45,000 0.344 Note 1 Equation :10 Thermal + 358F** 17,588 17,588 22,500 0.782 Note 1 OBED .....

Equation 11 Thermal + 358F 23,504 23,504 37,500 0.627 Note 1 Sustained I I

• from GE Stress Report 23A6129 Rev. I dated 217189 which is a part of Calculation No.: C-01 39 Rev. 8 Note 1: Not affected by Power Uprate.

• • Stresses at all other node points of SRV Line H are less than this value C:),

F-A-92 CD 0co (0

CO

0) 2r 3D Project: Hope Creek Extended Power Uprate Calculation No.: C-0139 Rev. 8 *Page Nos.:23, 37 & 472 Calculation

Title:

Main Steam Line ' D', MSRV Lines M, D & H

Description:

SRV Piping - Line M Maximum Stress Intensities CRITERIA PER ASME Node Original Power Code Ratio:

CODE SECTION III ND-3600 Number Maximum Uprate Allowable Power Uprate Remark Stress* Stress Stress to Original Equation : 8 Sustained Loads 174 7.648 7,648 22,500 0.340 Note 1 Equation : 9 Occasional 134N 18,358 19,404 27,000 0.719 00

-I Level C 174 8,802 9,304 33,750 0.276 Level D 134N 18,358 18,358 45,000 0.408 Note 1 Equation :10 Thermal + 156F- 17,743 17,743 22,500 0.789 Note 1 OBED I Equation 11 Thermal + 156F 22,257 22,257 37,500 0.594 Note 1 Sustained from GE Stress Report 23A6129 Rev. I dated 2/7189 which is a part of Calculation No.: C-0139 Rev. 8 Note 1: Not affected by Power Uprate.

• • Stresses at all other node points of SRV Line M are less than this value CI-C) 4 A-93 CDO CD

0 C)

Project: Hope Creek Extended Power Uprate Calculation No.: C-01 39 Rev. 8 *Page Nos. 60 through 72 Calculation

Title:

Main Steam Une 'D', MSRV Lines M.D & H

Description:

SRV Inlet & Outlet Flange Moments Loads Evaluation ExsigMaxirnumoelt ASME Loading Existing Ma & Ratio=

SRV Line Node Pt. Load Loading Combination EPU Loads Enveloped Enveloped Allowable Remark Level with TSV Load Loads Alowable SRV D 206 Level B MR 305,453 394,051 3T7,749 394,951 800,000 0.494 (Note 2)

Inlet [evel D MR 305.453 394,951 377,749 394,951 926,000 0.427 (Note 21 SRVD 210 Level B MR 239033 309,070 289,652 309070 60000 0.515 (Note Outlet LTehveID MIr 20 3,7 289,652 309,070 600,000 0 55 -. NotIe2T)

SRV H 306 Level 8 MR '356,145 461,401 474,055 474,055 800,000 0.593 OK (Note 1

-,4 Inlet Level D MR 357,945 462,823 474,038 474,038 925.000 0.512 OK (Note 1 (0

SRV H 310 Level B MR 304,158 393,276 320,333 393,276 600,000 0.655 (Note 2)

Outlet Level D MR 304,158 393,276 321,366 393,276 600,000 0.655 (Note 21 SRV M 106 Level B MR 140,402 181,540 243,358 243.358 800,000 0.304 OK (Note 1I Inlet Level D MR 136,633 176,666 242,037 242,037 925,000 0.262 OK (Note I )

SRV M 110 Level B MR 123,316 159,448 183,583 183,583 600,000 0.30 K Note 121 Outlet Level -M*

123,316 159,44!8 18 83 _1E3,583 60 0.306 2.A.9 K

• from GE Stress Report 23A6129 Rev. I dated 217189 which is a part of Calculation No. C-0139 Rev. 8 Forces and Moments are in Lba and in In-Lbs respectively, Note 1: EPU Loads are less than the Existing Enveloped Loads and are within allowable.

Note 2: The EPU loads are greater then the existing enveloped loads. However, the EPU loads are lower than the allowable.

0 "l-A-94 CD

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0 CD Hope Creek Extended Power Uprate =30 Project:

Calculation No.: C-0139 Rev. 8 -Page Nos. 46, 50 through 52 Calculation Titde: Main Steam Une' D', MSRV Lines M, D & H

Description:

Main Steam Piping - Line D SRV Valve Acceleration Loads Evaluation*

Maximium of Loading Existing EPUM & Alowable Remarks ASME Load Loading NodE Loa DCombination EPU Loads Enveloped Enveloped Node Pt. Description Level Direction wipthTSV LoadALoadsble Rema withTSV oad Loads 0.390 0.505 2.199 2.199 8.00 OK (Note I Level B AN AV 0.199 0,257 0.529 0.529 6.00 OK (Note I 20 SRV LINED AN 0,468 0.605 2.214 2.214 8.00 OK (Note 1)

Level D AV 0.257 0.333 0.763 0.763 6.00 OK (Note 1I All 0.412 0.533 2.901 2.901 8.00 OK (Note )

LevelB AV 0.263 0.340 0.528 0.526 6.00 OK (Note I2 306 SRVLINEH AN 0.515 0,666 2.917 2.917 8.00 OK (Note 1)

LevelD AV 0.327 0422 0.705 0.705 6.00 OK (Note 1.

Level B AH 0.412 0.532 1.779 1,779 8.00 OK (Note 1I

0) 106 SRV LINE M33 AV 0.180 0232 0.464 21 0.464 6.00 OK (Note 12I 2 2 0 (I( Note 1)

C) 03 Level A 0.511 0.661 1.805 .805 8.00 OK I AV 0.232 0.300 0.815 0.815 6.00 OK (Note 1 Level A and Level C are not affected by TSV Loads.

Acceleration Loads are in G's

• • from GE Stress Report 23A6129 Rev. I dated 217189 which is a part of Calculation No. C-0139 Rev. 8 Note 1: EPU loads are less then Existing Enveloped Loads. Therefore, the structural integrity of the SRV valves are not affected.

17-

_17 A-95 01-0 CD 0D

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-CD

0 CD RCIC Line C)r-A-98 CD CD C0

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-CD LR-N07-0099 LCR H05-01, Rev. 1 Project Hope Croek Ettended Power Uprate Calculation Nos.: 1. SR-1300, Rev.l dated 10112J93 *Pages 11. 14, and 18

2. SR-13t3, Rev.1 dated S7194 *Attachnent 2 Pages 2,221. and 222 Calculation Titles: 1. ASME Section III Class I Analysis of The Reactor Core Isolation Cooling System

2. RCIC Line from MS Line A to P11 RCIC Line No.: 1-FC-003-DBA-4'

Description:

RCIC Piping - Highest Stress Summary Data Point *Calculated EPU *Allowable Ratio Remarks Stress Stress Stress EPU/

(PSI) (PSi) (Psi) Allowable Eq.9: Design 455 949 16.949 26.410 0.642

,_NU 455 16,648 17,597 31,690 0.555 Emergency 455 17.383 17.383 398610 0.439 Faulted 472 21.13t 22,528 582,812 0.427 Note 1 Eq.110 472 58,302 59,993 52,812 1.136 Eq.12 472 1,409 1.408 52,812 0.027 Eq. 13 472 28,976 28,976 52,812 0.549 Usage Factor 472 0.0166 0.0166e 0.1000 0.168

-Since equation 10 is not satisfied, the piping is qualified by equations 12 and 13.

Note 1: Enveloped Faulted Values of 28497 Psi in SR-1300, Page 11 was not used due to lack of data point information.

A-99

- 14.46-82 -

LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Calcultion Nos.: 1. SR-1300, Rev.1 dated 10/12/93 'Pages 11, 14. and 18

2. SR-1303, Ra.1 dated 9g/94 "Allchrnent 2 Pages 2, 221, an0 222 Calculation Titles: 1. ASME Section III Class I Analysis of The Reactor Core Isolation Cooling System

2. RCIC Linefrom MS Line A to Pl1 RCIC Line No.: I-FC-003-DBA-4"

Description:

RCIC Piping - Highest Stess Sumnmaly Data Point *Calculated EPU Allowable Ratio Remarks Strnss Stress Stress EPUI (Psi) (PSI) (PSI) Alpowable Eq. 9: Design 455 1949 16,949 26,410 0.642 NIU 455 16Wi4l 17.597 31,690 0.555 Emergency 455 17,383 17,383 39,610 0.439 Faulted 472 121313 22,528 52.812 0.427 Note I Eg:I.10" 405 53.966 55222 52.812 1.046 Eq.12 405 31603 3603 52,812 0.058 Eq. 13 405 19,342 19342 52,812 0.366 usage P=actor 405 0.0075 9.0079 0.1500 0.079

-Sinoe equatlon 10 is not satirfied, the piping Is quAtd by equatlons 12 and 13.

Note 1: Enveloped Faulted Values of 28497 Psi In SR-1 300. Page 11 was not used due to lack of data point information.

A-100

- 14.46-83 -

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3 (D

Project: Hope Creek Extended Power Uprate Calculation Nos.: 1. SR-1300, Rev.1 dated 10/12/93 *Pages 11, 14, and 18

2. SR-1303, Rev.1 dated 9/7/94 *Attachment 2 Pages 2, 221, and 222 Calculation Titles: 1. ASME Section IIIClass I Analysis of The Reactor Core Isolation Cooling System

2. RCIC Une from MS Line A to P11 RCIC Line No.: 1-FC-003-DBA-4"

Description:

RCIC Piping - Highest Stress Summary Data Point Calculated EPU *Allowable Ratio Remarks Stress Stress Stress EPU/

I(Psi) (Psi) (Psi) Allowable Eq. 9: Design 455 16,949 16,949 26,410 0.642 N/U 455 16,648 17,597 31,690 0.555 0) co Emergency 455 17,383 17,383 R39610 0-439 Faulted 472 21,313 22.528 52,812 0.427 Note 1 Eq.10 455 49,756 51,199 52,812 0.969 Eq.12 Note 2 Eq. 13 Note 2 Usage Factor 405 0.0075 0.0076 0.1000 0.076 Note 1: Enveloped Faulted Values of 28497 Psi in SR-1300, Page 11 was not used due to lack of data point information. r-Note 2: Equations 12 & 13 are not required since Equation 10 is satisfied. 0

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CD Project: Hope Creek Extended Power Uprate Calculation No.: SR-1100, Rev 1 dated 9/21/94 *Pages 12, 16 and 37 Calculation

Title:

ASME Section Il Class 1 Analysis of The Main Steam Isolation Valve (Inboard) Drain System Line No. IDs: AB-050-DBA-3"

Description:

Main Steam Drain Une Class I - Maximum Stress Intensity Summary Data Point *Catculated EPU *Alowable Ratio Remarks Stress Stress Stress EPU/

(Psi) (Psi) (Psi) Allowable Eq. 9: Design 47 8,918 8,918 26,550 0.336 G) N/U 47 8,568 9,056 31.860 0284

0) Emergency 1 47 9,209 9,209 39,820 0.231 I

Faulted 47 21,230 22,440 53,100 0.423 Eq,10 145 44,557 45,849 53,100 0.863 Eq.12 *_

Eq. 13 _

Usage Factor 130 0.0205 0.0208 0.100 0.208

• • • Not required since Equation (10) requirement is mel r1-C)

A-104 4.4 0

0 3) 3O Project: Hope Creek Extended Power Uprate Calculation No.: SR-1100, Rey 1 dated 9W21194 *Pages 13.16 and 37 Calculation

Title:

ASME Section ItI Class 1 Analysis of The Main Steam Isolation Valve (Inboard) Drain System Line No. IDs: AB-1S46-DBA-2"

Description:

Main Steam Drain Une Class I - Maximum Stress Intensity Summary Data Point *Calculated EPU WAowable Ratio Remarks Stress Stress Stress EPU/

(Psi) (Psi) (Psi) Allowable Eq. 9: Design 417 7,317 7*317 26550 0.276

0) N/U 417 7,005 7,404 31,860 0,232 4 Emergency 417 7,577 7,577 39820 0.190 Faulted 417 12,063 12,751 53 100 0.240 Eq.10, " 255 37,988 39,090 53,100 0.736 Eq.12 __

Eq. 13 ...

Usage Factor 262 0.0165 0017 0.100 0.17

-Not required since Equation (10) requirement is met. 0 I

A-105 -0

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LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Exlended Power Uprate CGalauuitNo.: Calculation No. C-136, Rev 8 Section 3.2, Sheet Nos, 85 and 98 Calculation

Title:

Main Steam Drain Inside Drwelt Support Io: ABi090.429 Type: RAD (Node No. 290)

Support Cola.No.: 1-P-AB-050-C19, Rev.2 Sheet No. 5 Drawing No.ý FSK-P-1AB-603, Rev.6

Description:

From F22A - Main Steam Drain Line Support (Inside Dryweml)

SRP EPU -Qualgired Ratio = Remarks LoadCase Load Load Load Load EPU/

Direction (Lts) (Lbs) (Lbs) Qualified MAXNOR FA 0 0 1.050 0.000 OK MINNOR FA -397 -513 1,050 -0.489 OK MAXUPS FA 0 0 1,050 0.000 OK MINUPS FA -442 -572 1,050 -0.544 OK MAXFLT FA 35 45 1,397 0.032 OK MINFLT FA -933 -689 1.397 -0.493 OK "Qualified Load are taken from PSE&Gs Comment No. 45 dated 7/9103on Draft Task Report T0308, Rev. 0, (See Attachment # 9 of Ref. 4.)

A-106

- 14.46 0

(D O33 RPV Head Vent Line II C) r 6%0r-A-]HO A-ll -4 CD -4 0

CD W4 Project: Hope Creek Extended Power Uprale Calculation Nos.: 1. SR-10855-2100, Rev.I dated 3/2/93 'Pages 13,16, 17, 51 and 52

2. SR-2101, Rev.2 dated 8/30/95 *Attachment 2 Pages 2 and 5.

Calculation Titles: 1.ASME Section IIIClass 1 Analysis of The Head Vent System

2. The Head Vent System Head Vent Line Nos,: BB-192-DBA - 2" and AB-O51-DBA - 2"

Description:

Head Vent Piping - Highest Stress Summary Data Point "Calculated EPU *Allowable Ratio Remarks Stress Stress Stress EPU/

(Psi) (Psi) (Psi) Allowable Eq. 9 : Design 665 20,071 20,071 26,550 0.756 N/U 665 19,761 20,887 31,86 0.656 C) Emergency 665 20,331 20,331 39,820 0.511 0)

(0)

Faulted 665 42.805 45,245 53,100 0.852 Eq, 10;* 585 57,112 58,768 53,100 1.107 Eq. 12 585 7,957 7,957 53,100 0.150 Eq. 13 585 13,524 13,524 53,100 0.255 Usage Factor 565 0.0413 0.0419 0.1000 0.419

-Since equation 10 is riot satisfied, the piping is qualified by equations 12 and 13.

0

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C.0 LR-N07-0099 LCR H05-01, Rev. 1 BALANCE OF PLANT PIPING

- 14.46 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Calculation No.: SR-2400, Revision 1 Calculation Tile: Main Steam System Component IDs: Nodes 179 & 180

Description:

ASME Class I Piping - Butt Weld / TTJ @ Valve End CLTP Design Pressure No change CLTP Operating Temperature No Change EPU Design Pressure No change EPU Operating Temperature No Change Data Calculated EPU "Allowable Ratio Remarks ASMZ CLASS 1 Point Stress Stress Stress EPU/

(Psi) (Psi) (Psi) Allowable Eq. 9 Design No change N/U 179 7.600 8,033 31 860 0252 Emergency 179 8.724 9221 39,820 0.232 Faulted ' No change EqglO ' 180 23,593 24,277 63,100 0.457 Eq.12 EQ.13 Usage Factor 180 0.0020 0.0030 1 1 0.0030 A-2

- 14.46 LR-N07-0099 LCR H05-01, Rev. 1 Project Hope Creek Extended Power Uprate Calculation No.: SR-2400, Revision 1 Calculation Titde: Main Steam System Component IDs: Nodes 9 & 10

Description:

ASME Class 1 Piping - Butt Weld / TTJ @ Valve End CLTP Design Pressure No change CLTP Operating Temperature No Change EPU Design Pressure No change EPU Operating Temperature No Change Data Calculated EPU 'Allowable Ratio Remarks ASME CLASS I Point Stress Stress Stress EPUI (Psi) (Psi) (Psi) Allowable Eq. 9 Design No change a 782_N/U 0 8,054 31,860 0.253 Emernency 9 8,702 9198 39.820 0.231 Faulted Nochange Eq.1 0 10 24,999 25,724 53,100 0.484 Eq.12 .,

Eq.13 Usage Factor I10i 0.0024 0.0036 1 1 0.0036 A-3

- 14.46 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Calculation No.: SR-2400, Revision 1 Calculation

Title:

Main Steam System Component IDs: Nodes 529 & 530

Description:

ASME Class 1 Piping - Butt Weld / TTJ @ Valve End CLTP Design Pressure No change CLTP Operating Temperature No Change EPU Design Pressure No change EPU Operating Temperature No Change Data Calculated EPU *Allowable Ratio Remarks ASME CLASS 1 Point Stress Stress Stress EPUI (Psi) (Psi) (Psi) Allowable Eq. 9 Design No change NIU 529 7,951 8,404 31,860 0.264 Emernency 629 9,18 1 9,710 39,820 0.244 Faulted No change Eq.10 530 25,166 25,896 53,100 0,488 Eq.12________

E .13 U~a] Fctor 530 0.0024 0.0036 1 0.0036 A-4

- 14.46-94 -

LR-N07-0099 LCR H05-01, Rev. 1 Project Hope Creek Extended Power Uprate Calculation No.: SR-2400, Revision 1 Calculation

Title:

Main Steam System Component IDs: Node 350

Description:

ASME Class I Piping - TTJ @ Valve End CLTP Design Pressure No change CLTP Operating Temperature No Change EPU Design Pressure No change EPU Operating Temperature No Change Data Calculated EPU Allowable Rato Remarks ASME CLASS 1 Point Stress Stress Stress EPUI (Ps) (Psi) (Psi) Atowable Eq. 9 Design No change NIU 360 8,574 9,063 31 860 0.284 Emergency 350 10,004 10,574 39,820 0.266 Faulted . . No change Eq.10 350 27,194 27,983 53,100 0.527 Break _ 27,983 42,480 0.659 No break zone criteriansatisfied Eq.12 Eg.13 .. ....

Usage Factor 350 0.0034 0.0051 1 0.0051 0.0051 No break zone criteria satioed A-5

- 14.46 0

Project: Hope Creek Extended Power Uprate :3 CD Calculation No.: Pipe Stress Analysis C-0010, Revision 8 Calculation

Title:

Main Steam System Component IDs: Node 565B Desciription: ASME Class 2&3 Piping, B31.1 Piping CLTP Design Pressure No change CLTP Operating Temperature No Change EPU Design Pressure No change EPU Operating Temperature No Change ASME CLASS 2&3 Data Calculated EPU Allowable Ratio Remarks Q PIPING Point Stress Stress Stress EPU/

(Psi) (Psi) (Psi) Allowable Eq. 8 Sustained No change Eq. 9 N/U 565B 15722 16619 21000 0.792 C.6 Emergency 565B 16445 17383 31500 0.552 0)

Faulted No change Eqg10 I No change Eq.11 Not required

__1~i~j~1~J__I__I____

A-7 C-0 CDCD

-co LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Calculation No.: Pipe Stress Analysis C-0010, Revision 8 Calculation

Title:

Main Steam System Component IDs: Node 590/E25

Description:

ASME Class 2&3 Piping, B31.1 Piping CLTP Design Pressure No change CLTP Operating Temperature No Change EPU Design Pressure No change EPU Operating Temperature No Change ASME CLASS 2&3 Data Calculated EPU Allowable Ratio Remarks QS PIPING Point Stress Stress Stress EPUI (Psi) (Psi) (Psi) Allowable Eq. 8 Sustained No change Eq. 9 NIU 590 14975 15829 21000 0.754 Emergency N/A - Not evaluated Faulted No change Eq.10 No change Eq.11 Not required A-8

- 14.46 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Calculation No.: Pipe Stress Analysis C-0010, Revision 8 Calculation

Title:

Main Steam System Component IDs: Node 660/NZ29

Description:

ASME Class 2&3 Piping, B31.1 Piping CLTP Design Pressure No change CLTP Operating Temperature No Change EPU Design Pressure No change EPU Operating Temperature No Change Data Calculated EPU Allowable Ratio Remarks B31.1 PIPEIG Point Stress Stress Stress EPUI (PSI) (Psi) (Psi) Allowable Eq. 11 Sustained No change Eq. 12 N/U 660 5972 6313 15240 0.415 Emermency_ N/A- Not evaluated Faulted No changa Eq.13/14 N/A A-9

- 14.46-98 -

LR-N07-0099 LCR H05-01, Rev. 1 Project Hope Creek Extended Power Uprate Calculation No.: Pipe Stress Analysis C-0019, Revision 7 Calculation

Title:

Main Steam System Component IDs: Node 660/NZ29

Description:

ASME Class 2&3 Piping, 831.1 Piping CLTP Design Pressure No change CLTP Operating Temperature No Change EPU Design Pressure No change EPU Operating Temperature No Change Data Calculated EPU AlIowable Ratio Remarks B31.1 PIPING Point Stress Stress Stress EPU/

(Psi) (Psi) (Psi) Allowable Eq.11 Sustained No change Eq. 12 N/U 660 5665 5988 18000 0.333 Emerqency , N/A- Not evaluated

_ Faulteda N/A Eq.1 3114 N/A A-10

- 14.46 LR-N07-0099 LCR H05-01, Rev. 1 Project Hope Creek Extended Power Uprate Calculation No.: Pipe Stress Analysis C-1 922, Revision 6 Calculation Titde: Main Steam System Component lDs: Node 85

Description:

ASME Class 2&3 Piping, B31.1 Piping CLTP Design Pressure No change CLTP Operating Temperature No Change EPU Design Pressure No change EPU Operating Temperature No Change ASME CLASS 2&3 Date Calculated EPU Allowable Ratio Remarks Q PIPING Point Stress Stress Stress EPUl (Psi) (PSi) (Psi) Alowable Eq. 8 Sustained No change Eq. 9 NUl 85 5385 5692 18000 0.317

"_ Emergency NIA - Not evaluated Faulted No change Eq.10 No change Eq.11 No change A-Il

- 14.46-100 -

LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Calculation No.: Pipe Stress Analysis C-1 922, Revision 6 Calculation

Title:

Main Steam System Component IDs: Node 180

Description:

ASME Class 2&3 Piping, B31.1 Piping CLTP Design Pressure No change CLTP Operating Temperature No Change EPU Design Pressure No change EPU Operating Temperature No Change Data Calulated EPU Allowable Ratio Remarks B31.1 PIPING Point Stress Stress Stress EPU/

.. _(Psi) (Psi) (Psi) Ajiwable Eq. 11 Sustained No change Eq. 12 N/U 10" 4657 4923 18000 0.273 Energency NIA- Not evaluated Faulted No change Eq.13/14 N/A A-12

- 14.46-101 -

LR-N07-0099 LCR H05-01, Rev. 1 MAIN STEAM LINE "A" Calc. no. C-0122 Rev. 10 SUPT DATA LOAD SER EPU design REMARKS SUPPORT HANGER NO. SUPPORT CALC. NO. CALC PT DIR VICE Load Load Ratio =

REV. LEVEL (Ibs) (Ibs) EPU/des B 87848 761630 0.115 1-P-AB-030-H12 1-P-AB-030-C004 3 47 D 673258 761630 0.884 Includes pipe break load (Z B 33248 80000 0.416 D 88187 570000 0.155 Includes pipe break load 1-P-AB-030-H13 1-P-AB-030-C009 3 57 B 51285 191350 0.268 (RH13X) D 248232 322690 0.769 Includes pipe break load B 82089 388380 0.211 1-P-AB-030-H14 1-P-AB-030-CO10 3 67 D 266519 388380 0.686 Includes pipe break load (RH14Y, RH14X)

B 81816 319570 0.256 D 321816 393620 0.818 Includes pipe break load 1-P-AB-030-H03 1-P-AB-030-C006 6 19 Lat B 12567 44000 0.286 D 12567 44000 0.286 1-P-AB-030-H02 1-P-AB-030-C002 6 17 Lat B 17136 35000 0.490 D 17874 35000 0.511 1-P-AB-030-H05 1-P-AB-030-C003 6 29 Lat B 15922 44900 0.355 Snubber Allow D 18368 58000 0.317 1-P-AB-030-H04 1-P-AB-030-CO01 8 23 Lat B 5976 44900 0.133 SnubberAllow D 5976 70000 0.085 1-P-AB-053-H16 1-P-AB-053-C015 7 316 Lat B 1172 36670 0.032 C 2281 36670 0.062 1-P-AB-061-H11 1-P-AB-061-C013 5 124 Lat B 5028 24693 0.204 C 5326 24693 0.216 1-P-AB-066-H 11 1-P-AB-066-CO05 7 226 Lat B 13549 31400 0.431 C 15286 34100 0.448 14.46-102 LR-N07-0099 LCR H05-01, Rev. 1 MAIN STEAM LINE "B" Calc. no. C-0121 Rev. 8 SUPT DATA LOAD SER EPU desi2n REMARKS SUPPORT SUPPORT CALC.

HANGER NO. NO. CALC PT DIR VICE Load Load Ratio =

REV. LEVEL (Ibs) (Ibs) EPU/des 1-P-AB-031-H02 1-P-AB-031-C004 6 19 Lat B 14701 44900 0.327 Snubber Allow D 18421 70000 0.263 1-P-AB-031-H03 1-P-AB-031-C003 4 21 Lat B 16698 44900 0.372 Snubber Allow D 20045 70000 0.286 1-P-AB-031-H04 1-P-AB-031-C005 7 29 Lat B 11201 40000 0.280 D 13816 40000 0.345 Includes pipe break load 1-P-AB-031-HO5 1-P-AB-031-C002 5 33 Lat B 7213 44900 0.161 Snubber Allow D 10193 70000 0.146 1-P-AB-031-H07 1-P-AB-031-CO01 5 45 Lat B 24634 44900 0.549 Snubber Allow

_D 27453 50000 0.549 1-P-AB-031-H13 1-P-AB-031-CO11 2 51 Lat B 62966 175330 0.359 D 792534 831300 0.953 Includes pipe break load 1-P-AB-031-H 12 1-P-AB-031-C006 2 53 Lat B 91769 400000 0.229 D 607315 660000 0.920 Includes pipe break load 1-P-AB-031-H14 1-P-AB-031-C012 2 61 Lat B 19861 250000 0.079 D 271748 275000 0.988 Includes pipe break load 1-P-AB-054-H 13 1-P-AB-054-C013 7 316 Lat B 1137 15000 0.076 Snubber Allow C 3778 15650 0.241 1-P-AB-058-H12 1-P-AB-058-C012 7 418 Lat B 347 15000 0.023 Snubber Allow C 765 15460 0.049 1-P-AB-062-H 17 1-P-AB-062-C016 5 248 Lat B 9516 24200 0.393 C 11369 24200 0.470 1-P-AB-065-H 15 1-P-AB-065-C017 5 132 Lat B 2746 40473 0.068 Snubber Allow C 4154 44530 0.093 14.46-103 LR-N07-0099 LCR H05-01, Rev. 1 MAIN STEAM LINE "C" C-0138 Rev. 10 SUPT SER EPU design REMARKS SUPPORT SUPPORT CALC DATA LOAD VICE Load Load Ratio =

HANGER NO. CALC. NO. REV. PT DIR LEVEL (Ibs) (Ibs) EPU/des 1-P-AB-032- 1-P-AB-032-H02 cool 5 14 Lat B 34759 44900 0.774 snubber capacity = design load D 34759 44900 0.774 Support is acceptable 1-P-AB-032- 1-P-AB-032-H03 C002 5 17 Lat B 22613 25000 0.905 D 22613 25000 0.905 Support is acceptable 1-P-AB-032- 1-P-AB-032-H04 C003 5 22 Lat B 12868 25688 0.501 D 15409 25688 0.600 Support is acceptable 1-P-AB-032- 1-P-AB-032-H07 C008 6 37 Fvert B 26864 44900 0.598 D 28692 44900 0.639 Support is acceptable 1-P-AB-055- 1-P-AB-055-H18 C022 6 319 Fvert B 606 15589 0.039 C 2368 15589 0.152 Support is acceptable 1-P-AB-057- 1-P-AB-057-Hll C008 6 119 Lat B 4075 15000 0.272 C 4985 15000 0.332 Support is acceptable 1-P-AB-059- 1-P-AB-059-H05 C003 5 421 Lat B 1499 15296 0.098 C 2656 15296 0.174 Support is acceptable 1-P-AB-063- 1-P-AB-063-H08 C008 5 215 Fvert B 809 38000 0.021 C 2705 38000 0.071 Support is acceptable 1-P-AB-032- 1-P-AB-032-H12 C006 2 45 Lat B 91769 400000 0.229 similar to 1-P-AB-031-H12, calc. C-0121 14.46-104 LR-N07-0099 LCR H05-01, Rev. 1 MAIN STEAM LINE "C" C-0138 Rev. 10 (Cont.)

SUPT EPU design REMARKS SER SUPPORT SUPPORT CALC DATA LOAD VICE Load Load Ratio =

HANGER NO. CALC. NO. REV. PT DIR LEVEL (Ibs) (Ibs) EPU/des (PR52) D 607315 660000 0.920 Includes pipe break loads. support is acceptable 1-P-AB-032- 1-P-AB-032-H13 C010 2 44 Lat B 62966 175330 0.359 similar to 1-P-AB-031-H13, calc. C-0121 D 792534 831300 0.953 Includes pipe break loads, support is acceptable 1-P-AB-032- 1-P-AB-032-H14 C011 2 49 Lat B 27288 175190 0.156 (PR57) D 190487 250000 0.762 Includes pipe break load, support is acceptable 1-P-FD-001- 1-P-FD-001-H01 cool 5 532 Lat B 6936 8000 0.867 (RH01) D 6936 8000 0.867 Support is acceptable 1-P-FD-001- 1-P-FD-001-H05 C005 5 531 Lat B 1968 35484 0.055 (SH05) D 2596 35484 0.073 Support is acceptable 1-P-FD-001- 1-P-FD-001-H06 C006 4 523 Lat B 6355 82800 0.077 (RH06) D 6355 82800 0.077 Support is acceptable 1-P-FD-001- 1-P-FD-001-H07 C007 5 527 Lat B 3428 93947 0.036 (RH07) D 3428 93947 0.036 Support is acceptable 14.46-105 LR-N07-0099 LCR H05-01, Rev. 1 MAIN STEAM LINE "D" C-0139 Rev. 9 SUPT EPU design REMARKS SUPPORT SUPPORT CALC DATA LOAD SER VICE Load Load Ratio=

HANGER NO. CALC. NO. REV. PT DIR LEVEL (Ibs) (Ibs) EPU/des 1-P-AB-033-H02 1-P-AB-033-CO01 6 17 Lat B 10745 44900 0.239 snubber capacity = design load D 12573 44900 0.280 Support is acceptable 1-P-AB-033-H03 1-P-AB-033-C006 8 19 Lat B 13951 44900 0.311 snubber capacity = design load D 13951 44900 0.311 Support is acceptable 1-P-AB-033-H05 1-P-AB-033-C003 7 27 Lat B 10534 14362 0.733 D 11560 14362 0.805 Support is acceptable 1-P-AB-033-H04 1-P-AB-033-C008 6 29 Lat B 3249 8314 0.391 D 3557 8314 0.428 Support is acceptable 1-P-AB-033-H 12 1-P-AB-033-C002 3 40 X B 28004 60000 0.467 Support is acceptable D 60009 60000 1.000 Includes pipe break load 41 Lat B 88406 570000 0.155 D 525488 570000 0.922 Includes pipe break load 1-P-AB-033-H 13 1-P-AB-033-CO11 3 46 X B 73629 200000 0.368 Support is acceptable D 282116 322690 0.874 Includes pipe break load 1-P-AB-056-H19 1-P-AB-056-C018 7 222 Fvert B 269 10819 0.025 C 1840 10819 0.170 Support is acceptable 1-P-AB-064-H10 1-P-AB-064-C007 6 126 Flat B 1688 24455 0.069 C 1897 24455 0.078 Support is acceptable 1-P-AB-060-H06 1-P-AB-060-C008 6 330 Flat B 4713 26499 0.178 I C 8123 26499 0.307 Support is acceptable 14.46-106 LR-N07-0099 LCR H05-01, Rev. 1 RCIC LINE from MS Line "A" to Pll Calc. no. C-0123 SUPPORT SUPPORT CALC. SUPT DATA LOAD SER EPU design REMARKS HANGER NO. NO. CALC PT DIR VICE Load Load Ratio =

REV. LEVEL (Ibs) (Ibs) EPU/des 1-P-FC-003- 1-P-FC-003- 7 416 Lat N/U 1271 1339 0.949 1 H08 H08 C006 C006 FLTD 1891 2120 0.892 MAIN STEAM Drain Inside the Drywell Calc. no. C-0136 SUPPORT SUPPORT CALC.SE SUPT DATA LOAD SER EPU desizn REMARKS HANGER NO. NO. CALC PT DIR VICE Load Load Ratio =

REV. LEVEL (Ibs) (Ibs) EPU/des 1-P-AB-050- 1-P-AB-050- 3 290 Y N/U 548 1050 0.522 H020 0019 FLTD 660 1050 0.629 1-P-AB-050- 1-P-AB-050- 1 231 Y N/U 204 235 0.868 H030 C028 FLTD 269 294 0.916 1-P-AB-050- 1-P-AB-050- 2 352 N/U 307 650 0.472 H029 C020 FLTD 416 650 0.640 1-P-AB-050- 1-P-AB-050- 1 410 Y N/U 269 1150 0.234 H021 C027 FLTD 385 1150 0.335 RPV VENT LINE CaIc. no. C-1842 SUPPORT SUPPORT CALC. SUPT DATA LOAD SER EPU design REMARKS HANGER NO. NO. CALC PT DIR VICE Load Load Ratio =

REV. LEVEL (Ibs) (Ibs) EPU/des 1-P-BB-192- 1-P-BB-192- 4 290 Lat N/U 836 1090 0.767 H009 _____ C008 ____

_______ ____ ____FLTD 1090 1363 0.800 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

14.46-107 LR-N07-0099 LCR H05-01, Rev. 1 Calc. No. C-0010 Exsig CPPU Design Ratio= Rmrs Support Mk No. NP Direction Existing Ld Load R Load Load "B" A/B 1-P-AB-001- 230 Y-Snub 9392 12072 33505 0.360 H017 1-P-AB-001- 278 Z-Snub 24095 33002 47908 0.689 H002 1-P-AB-001- 27A X-Snub 56695 77619 79000 0.98 H003 1-P-AB-001- 272 Y-Stop 17564 18223 23289 0.782 H004 1-P-AB-001- 250 Y-Stop 22206 23535 27558 0.854 H008 1-P-AB-001- 230 X-Stop 7604 9818 31280 0.314 H018 1-P-AB-001- 280 Y-Stop 9859 10162 32648 0.311 H001 1-P-AB-001- 265 Y-Stop 14657 15404 25107 0.614 H005 1-P-AB-001- 260 Y-Stop 19550 20665 28901 0.715 H007 1-P-AB-001- 242 Y-Stop 73599 77043 73599 1.047 Max. stress ratio = 1.05(0.28/0.5) = 0.59 < 1.0 OK H01 1 1-P-AB-001- 238 Y-Stop 37022 38607 38238 1.01 Max stress ratio = 1.01 (0.233/0.375) = 0.628 < 1.0 ok H012 1-P-AB-O01- 234 Y-Stop 21860 23293 28158 0.827 H015 1-P-AB-002- 111 Z-Snub 25662 35155 59640 0.589 H001 1-P-AB-002- 74 Y-Snub 57109 76821 106672 0.720 H008 1-P-AB-002- 102 X-Snub 58427 80031 58427 1.370 Max. stress ratio = 1.37(1/2.257) = 0.61< 1 .0 OK H003

• Note: If the Stress Ratio was greater than one, the existing calculations were reviewed in detail. A load comparison was performed for the CPPU load versus the design load. The critical margin was identified, conservatism removed, and a new Stress Ratio was calculated as documented in the Remarks Column.

14.46-108 LR-N07-0099 LCR H05-01, Rev. 1 Calc. No. C-0010 Support Mk No. NP Direction Existing CPPU Design Ratio=Remarks*

Load LoA" "BL A/B 1-P-AB-002- 110 Y-Stop 14402 14742 39024 0.378 H002 1-P-AB-002- 95 Y-Stop 16224 17110 26797 0.639 H004 1-P-AB-002- 90 Y-Stop 17685 18380 31605 0.582 H005 1-P-AB-002- 80 Y-Stop 21075 22027 22335 0.986 H007 1-P-AB-002- 72 Y-Stop 36749 38026 36749 1.03 Max. stress ratio = 1.03(0.309/0.375) = 0.85 < 1 .0 OK H009 1-P-AB-002- 68 Y-Stop 33803 35387 33804 1.05 Max. stress ratio = 1.05(0.406/0.5) = 0.85 < 1 .0 OK H011 1-P-AB-002- 67 X-Stop 30997 35340 30997 1.14 Max. (N/U) stress ratio = 16374/20365 = 0.80 < 1.0 OK H012 1-P-AB-002- 64 Y-Stop 20708 21848 20708 1.06 Max. (N/U) stress ratio = 1.06(0.353/0.5) = 0.75 < 1 .0 OK H014 1-P-AB-002- 60 Y-Snub 8989 11561 12000 0.963 H016 1-P-AB-002- 60 X-Stop 9645 10791 15660 0.689 H017 1-P-AB-003- 62A X-Snub 53493 73203 53493 1.368 Max. stress ratio = (1.343/1.356) = 0.99 < 1 .0 OK H017 1-P-AB-003- 622 Y-Stop 11849 12260 11849 1.035 Max. stress ratio = (12260/11849)(1/1.827) = 0.57 < 1.0 OK H004 1-P-AB-003- 632 Y-Stop 15933 16454 37323 0.441 H002 1-P-AB-003- S33 Z-Snub 19965 27331 19965 1.369 Max. stress ratio = (27331/19965 )(1/1.835) = 0.75 < 1.0 OK H001 1-P-AB-003- Max. stress ratio = (13651/13102)(0.278/0.375) = 0.77 < 1.0 H006 615 Y-Stop 13102 13651 13102 1.042 OK 1-P-AB-003- 610 Y-Stop 12322 13121 22611 0.580 H007 1-P-AB-003- 605 Y-Stop 24883 26410 51605 0.512 H008 14.46-109 LR-N07-0099 LCR H05-01, Rev. 1 Calc. No. C-0010 Support Mk No. NP Direction Existing CPPU Design Ratio=Remarks*

Load Load "A. Load

..B A/B 1-P-AB-003- 588 Y-Stop 29640 31537 42883 0.735 H010 1-P-AB-003- 587 X-Stop 30431 36711 37000 0.99 H011 1-P-AB-003- 586 X-Stop 13096 15529 29804 0.521 H012 1-P-AB-003- 584 Y-Stop 18085 19670 18085 1.088 Max. stress ratio = (19670/18085)(0.34/0.5) = 0.74 < 1.0 H013 1-P-AB-003- 580 Y-Snub 12030 16033 24524 0.654 H015 1-P-AB-003- Max. Upset ratio = (13931/12841)(1765/23900) = 0.08 < 1.0 H016 580 X-Stop 12617 13931 12841 1.085 OK 1-P-AB-004- 452 Y-Stop 13940 14478 27710 0.522 H002 1-P-AB-004- 453 Z-Snub 30355 41565 44000 0.94 H001 1-P-AB-004- 417 Z-Snub 66809 90963 91000 1.00 H008 1-P-AB-004- 407 X-Stop 33071 39504 39600 1.00 HOll 1-P-AB-004- 400 X-Stop 7803 9230 34720 0.266 H016 1-P-AB-004- 408 Y-Stop 38216 39468 45172 0.874 H010 1-P-AB-004- 400 Y-Snub 11424 14971 11424 1.310 Max. stress ratio = (14971/11424)(1/1.357) =0.97 < 1.0 OK H015 1-P-AB-004- 435 Y-Stop 16093 16675 16093 1.036 Max. stress ratio = (16675/16093)(0.26/0.375) = 0.72 < 1.0 OK H005 1-P-AB-004- 430 Y-Stop 13187 13775 13187 1.045 Max. stress ratio = (13775/13187)(0.24/0.375) = 0.67 < 1.0 OK H006 1-P-AB-004- 420 Y-Stop 14972 15544 14973 1.038 Max. stress ratio = (15544/14973)(0.274/0.375) = 0.76 < 1.0 OK H007 1-P-AB-004- 412 YStop 34171 35322 34171 1.034 Max. stress ratio = (35322/34171)(0.266/0.375) = 0.73 < 1.0 OK H009 14.46-110 LR-N07-0099 LCR H05-01, Rev. 1 Calc. No. C-0010 Existing CPPU Design Ratio=

Support Mk No. NP Direction Load a Load Remarks*

Load A.. .B" O 1-P-AB-004- 406 X-Stop 15322 17905 15322 1.169 Max. stress ratio (N/U) = (1.17)(1/2.165) = 0.54 < 1.0 OK H012 1-P-AB-004- 404 Y-Stop 21052 22374 29650 0.755 H013 1-P-AB-005- 741 X-Snub 12117 16599 12117 1.370 Max. stress ratio (N/U) = (1.37)(1/1.776) = 0.77 < 1.0 OK H005 1-P-AB-005- 720 Y-Stop 3659 3987 3658 1.090 Max. stress ratio = (3987/3658)(0.14/0.3125) = 0.49 < 1.0 OK H001 1-P-AB-005- 736 Y-Stop 7217 7479 8767 0.853 H003 1-P-AB-005- 743 Y-Stop 3454 3688 3453 1.068 Max stress ratio = 1.068(0.153/0.25) = 0.65 < 1.0 OK H006 1-P-AB-005- 835 Y-Stop 2396 2508 4471 0.561 H007 15107 14673 1.030 27280 26950 1.012 1-P-AB-005- 757 ANC 62264 60337 1.032 H009 61044 60849 1.003 EPU Margin factor = 1.107/1.04 =1.06 > 1.0 ok 69039 66334 1.041 144239 143504 1.005 1-P-AB-006- 910 Y-Stop 2973 3180 2973 1.070 H001 1-P-AB-006- 775 Y-Stop 0 8734 9011 0.969 H014 1-P-AB-006- 774 X-Stop 0 32115 32491 0.988 H036 1-P-AB-006- 860 Y-Stop 2857 3010 2875 1.047 Max. stress ratio = (3010/2857)(1/2.1) = 0.50 < 1 .0 OK H008 1-P-AB-006- 810 Y-Stop 19636 20356 21149 0.963 H010 1-P-AB-006- 800 Y-Stop 7649 7753 9482 0.818 H01 1 1-P-AB-006- 950 Y-Stop 13594 13961 14003 0.997 14.46-111 LR-N07-0099 LCR H05-01, Rev. 1 Calc. No. C-0010 Support Mk No. NP Direction Existing CPPU Design Ratio=Remarks*

Load LA" "B" A/B H002 1-P-AB-006- 790 Y-Stop 6555 6817 10049 0.678 H013 1-P-AB-138- E38 Y-Stop 29114 29939 29115 1.028 Max. stress ratio = (29939/29115)(1/1.188) = 0.87 < 1 .0 OK H005 1-P-AB-138- E15 Y-Stop 23709 25024 23708 1.056 Max. stress ratio = (25024/23708)(1/1.522) = 0.69 < 1.0 OK H003 1-P-AB-138- E45 Y-Stop 50981 52719 63400 0.832 H006 1-P-AB-138- E7 X-Snub 30886 42217 42300 1.00 H001 1-P-AB-138- E55 Y-Stop 52220 54231 66900 0.811 H007 1-P-AC-006- 666 Snub- 6780 9277 42913 0.216 H001 LAT 1-P-AC-006- 664 Snub- 9179 12570 37811 0.332 H002 LAT 1-P-AC-006- 675 Y-Snub 1848 2529 43188 0.059 H003 1-P-AC-006- 690 X-Snub 1369 1876 1369 1.370 Max. stress ratio = (1876/1369 )(0.165/0.375) =0.60 < 1.0 OK H004 1-P-AC-007- 486 Snub- 7260 9935 42913 0.232 H001 LAT 1-P-AC-007- 484 Snub- 7249 9926 37811 0.263 H002 LAT 1-P-AC-007- 495 Y-Snub 1450 1984 30510 0.065 H003 1-P-AC-007- 512 Z-Snub 28339 28654 28339 1.011 Max stress ratio 1.01(1/2.319)

H005 = = 0.44 < 1.0 O' 1-P-AC-008- 316 Snub- 7584 10380 42913 0.242 H001 LAT 1-P-AC-008- 314 Snub- 44144 8695 37811 0.230 H002 LAT 1-P-AC-008- 322 Y-Snub 1496 2048 15953 0.128 14.46-112 LR-N07-0099 LCR H05-01, Rev. 1 Calc. No. C-0010 Existing CPPU Design Ratio=

Support Mk No. NP Direction Load Load Load Remarks*

Load A .. .B" /

H003 1-P-AC-009- 146 Snub- 7662 10487 42913 0.244 H001 LAT 1-P-AC-009- 144 Snub- 7158 9798 38169 0.257 H002 LAT 1-P-AC-009- 153 Y-Snub 1306 1788 37616 0.048 H003 1-P-AC-009- 167 Z-Snub 195615 195957 195615 1.002 Max stress ratio = 1.002(1/1.003) = 0.999 < 1.0 ok H005 664 703 7045 7094 1-P-AD-001- A90 ANC 2886 3006 This support was designed for Plastic Load H003 12360 13105 438 559 6504 6720 2727 2769 2289 2606 1-P-CA-001- 940 ANC 7178 7528 This support was designed for Plastic Load H007 14274 14934 5396 5461 5882 6071 1-P-CA-001- 930 Y-Stop 854 954 2321 0.411 H004 1-P-CA-001- 931 X-Stop 3509 3721 3510 1.060 Max stress ratio = (3721/3510)(1/1.13) = 0.94 < 1.0 OK H005 1 P-CV-000- 127 X-Snub 34483 47220 158857 0.297 H001 1-P-CV-000- SVA Z-Snub 21117 28909 63400 0.456 H002 1-P-CV-000-H003 (MST- SVB Z-Snub 18459 25288 63400 0.399

3) 1 1 1 14.46-113 LR-N07-0099 LCR H05-01, Rev. 1 Calc. No. C-0010 Existing CPPU Design Ratio=

Support Mk No. NP Direction Load Load Load A/B Remarks*

Lod "A" "B"B 2112 2304 3882 0.594 13064 13436 17650 0.761 1-P-FW-001- 2738 2891 5301 0.545 H016 7282 7539 10472 0.720 18242 19236 28514 0.675 5498 5655 7432 0.761 1-P-FW-001- 985 Y-Stop 16364 16876 16364 1.031 Max stress ratio = 1.03(0.303/0.3125) = 0.998 < 1.0 OK H001 1-P-AB-001- 244 Z-Snub 103441 140549 148000 0.95 H009 1-P-AB-001- 237 X-Stop 32986 37718 42000 0.90 H013 1-P-AB-001- 236 X-Stop 16790 18561 19500 0.95 H014 1-P-AB-002- 66 X-Stop 15096 16282 16300 1.00 H013 1-P-AB-003- 600 Z-Snub 71675 98019 98100 1.00 H009 1-P-AB-004- 442 X-Snub 54913 75007 82500 0.91 H004 14.46-114 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-008-H02 Support Calc. No.: 1-P-AB-008-C 11

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable MAXNOR N/A MINNOR N/A MAXUPS FZ 514 551 1,700 0.324 MINUPS FZ 514 551 1,700 0.324 14.46-115 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-008-H03 Support Calc. No.: 1-P-AB-008-C3

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 900 1,233 1,700 0.725 MINUPS FY 900 1,233 1,700 0.725 14.46-116 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-008-H05 Support Calc. No.: 1-P-AB-008-C5

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 3,040 3,077 3,077 1.000 FZ 493 530 1,700 0.312 MINUPS FY 3,040 3,077 3,077 1.000 FZ 493 530 1,700 0.312 14.46-117 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-008-H06 Support Calc. No.: 1-P-AB-008-C6

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 869 906 1,700 0.533 MINUPS FY 869 906 1,700 0.533 4 + 4 F 4 + + 4 F 14.46-118 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-008-H07 Support Calc. No.: 1-P-AB-008-C7

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) CLTP "Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 1,024 1,061 1,700 0.624 FZ 649 731 1,700 0.430 MINUPS FY 1,024 1,061 1,700 0.624 FZ 649 731 1,700 0.430 t + 4 + 4 14.46-119 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass AB-008-H08 (NP Support ID: 52)

Support Calc. No.: 1-P-AB-008-C2

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR MINNOR MAXUPS FX 3,380 4,490 22,730 0.198 MINUPS FX 3,380 4,490 22,730 0.198 4 4 4-14.46-120 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-008-H09 Support Calc. No.: 1-P-AB-008-C8

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 1,340 1,377 1,700 0.810 MINUPS FY 1,340 1,377 1,700 0.810

_______ ______ ______ ______ -I- -I-I +/- +/- F 14.46-121 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-008-H 10 Support Calc. No.: 1-P-AB-008-C9

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 1,346 1,383 1,700 0.814 MINUPS FY 1,346 1,383 1,700 0.814 14.46-122 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-009-H02 Support Calc. No.: I -P-AB-009-C6

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FZ 503 540 1,700 0.318 MINUPS FZ 503 540 1,700 0.318 14.46-123 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-009-H03 Support Calc. No.: 1-P-AB-009-C3

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 900 1,233 1,700 0.725 MINUPS FY 900 1,233 1,700 0.725 14.46-124 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No. Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-009-H05 Support Calc. No.: 1-P-AB-009-C8

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 3,060 3,097 3,100 0.999 FZ 532 569 1,700 0.335 MINUPS FY 3,060 3,097 3,100 0.999 FZ 532 569 1,700 0.335 14.46-125 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-009-H06 Support Calc. No.: 1-P-AB-009-C4

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 816 853 1,700 0.502 MINUPS FY 816 853 1,700 0.502 14.46-126 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-009-H07 Support Calc. No.: 1-P-AB-007-C9

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 1,122 1,159 1,700 0.682 FZ 951 1,025 1,700 0.603 MINUPS FY 1,122 1,159 1,700 0.682 FZ 951 1,025 1,700 0.603 14.46-127 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-009-H08 (NP 152)

Support Calc. No.: 1-P-AB-009-C2 Rev. 3

Description:

8" Main Steam Bypass Remark CLTP EPU Design Ratio s Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FX 3,134 4,170 22,712 0.184 MINUPS FX 3,134 4,170 22,712 0.184 14.46-128 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-009-H09 Support Calc. No.: 1-P-AB-009-C10

Description:

8" Main Steam Bypass Remark CLTP EPU Design Ratio s Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 1,111 1,148 1,700 0.675 MINUPS FY 1,111 1,148 1,700 0.675 14.46-129 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-009-H 10 Support Calc. No.: 1-P-AB-009-C 11

Description:

8" Main Steam Bypass Remark CLTP EPU Design Ratio s Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 1,008 1,045 1,500 0.697 Strut assembly weak link allowable =

1500#.

MINUPS FY 1,008 1,045 1,500 0.697 14.46-130 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-010-H02 Support Calc. No.: 1-P-AB-010-C6

Description:

8" Main Steam Bypass Remark CLTP EPU Design Ratio s Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FZ 494 531 1,700 0.312 MINUPS FZ 494 531 1,700 0.312 14.46-131 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-010-H03 Support CaIc. No.: 1-P-AB-010-C3

Description:

8" Main Steam Bypass Remark CLTP EPU Design Ratio s Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 900 1,233 1,700 0.725 MINUPS FY 900 1,233 1,700 0.725 14.46-132 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No. Pipe Stress Calculation C-0019, Rev. 7 Main Steam Calculation

Title:

Bypass Support ID: AB-010-H05 Support Calc. No.: 1-P-AB-010-C8

Description:

8" Main Steam Bypass Remark CLTP EPU Design Ratio s Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR MINNOR MAXUPS FY 3,010 3,047 3,047 1.000 FZ 570 607 1,700 0.357 MINUPS FY 3,010 3,047 3,047 1.000 FZ 570 607 1,700 0.357 14.46-133 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No.: Pipe Stress Calculation C-0019, Rev.7 Main Steam Calculation

Title:

Bypass Support ID: AB-010-H06 Support Calc. No.: 1-P-AB-010-C9

Description:

8" Main Steam Bypass Remark CLTP EPU Design Ratio s Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR MINNOR MAXUPS FY 873 910 1,700 0.535 MINUPS FY 873 910 1,700 0.535 14.46-134 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Analysis No. : Pipe Stress Calculation C-0019, Rev.7 Main Steam Calculation

Title:

Bypass Support ID: AB-0110-H07 Support Calc. No.: 1-P-AB-010-Cl0

Description:

8" Main Steam Bypass Remark CLTP EPU Design Ratio s Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 702 739 1,700 0.435 FZ 1,032 1,069 1,700 0.629 MINUPS FY 702 739 1,700 0.435 FZ 1,032 1,069 1,700 0.629 common with support AB-008-H07.

14.46-135 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-010-H08 (NP 252)

Support Calc. No.: 1-P-AB-010-C4 Rev. 3

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FX 2,969 3,968 22,674 0.175 MINUPS FX 2,969 3,968 22,674 0.175 14.46-136 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Calculation

Title:

Main Steam Outside Bypass Support ID: AB-010-Ho9 Support Calc. No.: 1-P-AB-010-C 11

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 1,926 1,963 1,926 1.019 See below MINUPS FY 1,926 1,963 1,926 1.019 See below Existing Minimum Margin Factor = 2.27 New Minimum Margin Factor = 2.27/1.019 = 2.23 > 1.0 OK 14.46-137 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-010-HlO Support Calc. No.: 1-P-AB-010-C2

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 1,210 1,247 1,523 0.819 MINUPS FY 1,210 1,247 1,523 0.819 14.46-138 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-011-H02 Support Calc. No.: 1-P-AB-01 1-C3

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FZ 236 273 1,700 0.161 MINUPS FZ 236 273 1,700 0.161

-I- +

4- + +

14.46-139 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-011-H03 Support Calc. No.: 1-P-AB-011-C4

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 900 1,233 1,700 0.725 MINUPS FY 900 1,233 1,700 0.725 14.46-140 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-011-H05 Support Calc. No.: 1-P-AB-010-C6

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 2,399 2,436 2,436 1.000 FZ 708 745 1,700 0.438 MINUPS FY 2,399 2,436 2,436 1.000 FZ 708 745 1,700 0.438

____ I

___ I F I I I 14.46-141 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Analysis No.: Rev.7 Main Steam Calculation

Title:

Bypass Support ID: AB-011-H06 (NP 342)

Support Calc. No.: 1-P-AB-011-C7 (Rev. 4)

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable MAXNOR N/A MINNOR N/A MAXUPS FX 2,500 3,203 22,880 0.140 MINUPS FX 2,500 3,203 22,880 0.140 Similar support AB-01O0-H08 lug weld margin factor 1.358 &

pipe clamp allowable = 30,000#.

_ _ _ I _ _ _ 1 _ _ _ I___

14.46-142 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-011-H07 Support Calc. No.: 1-P-AB-01 1-C8

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 1,089 1,126 1,700 0.662 MINUPS FY 1,089 1,126 1,700 0.662 F + 4 4 14.46-143 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-011-H08 Support Calc. No.: 1-P-AB-0 11 -C9

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 1,077 1,144 1,700 0.673 MINUPS FY 1,077 1,144 1,700 0.673 14.46-144 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-012-H02 Support Calc. No.: 1-P-AB-012-C3

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FZ 327 364 1,700 0.214 MINUPS FZ 327 364 1,700 0.214 F + 4 + +

F F + 4 + t 14.46-145 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-012-H03 Support Calc. No.: 1-P-AB-012-C4

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 900 1,233 1,700 0.725 MINUPS FY 900 1,233 1,700 0.725 14.46-146 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-012-H05 Support Calc. No.: 1-P-AB-012-C6

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 2,247 2,284 2,284 1.000 FZ 937 974 1,700 0.573 MINUPS FY 2,247 2,284 2,284 1.000 FZ 937 974 1,700 0.573 14.46-147 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-012-H06 (NP 442)

Support Calc. No.: 1-P-AB-012-C7 Rev. 3

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FX 2,461 3,126 4,000 0.782 MINUPS FX 2,461 3,126 4,000 0.782 Common w/ support AB-01 0-H08 14.46-148 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-012-H07 Support Calc. No.: 1-P-AB-012-C8

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 956 993 1,700 0.584 MINUPS FY 956 993 1,700 0.584 14.46-149 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-012-H08 Support Calc. No.: 1-P-AB-012-C8

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 808 845 1,700 0.497 MINUPS FY 808 845 1,700 0.497 1 4 4 4 4 + 4 4 F 4 + 4 F 4 14.46-150 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-013-H02 Support Calc. No.: 1-P-AB-013-C3

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FZ 413 450 1,700 0.265 MINUPS FZ 413 450 1,700 0.265 14.46-151 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-013-H03 Support Calc. No.: 1-P-AB-013-C4

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FZ 900 1,233 1,700 0.725 MINUPS FZ 900 1,233 1,700 0.725 14.46-152 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-013-H05 Support Calc. No.: 1-P-AB-013-C6

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR N/A MINNOR N/A See calc 1-P-SB-008-C005 MAXUPS FY 2,217 2,254 2,254 1.000 Rev. 2 FZ 1,242 1,279 1,700 0.752 MINUPS FY 2,217 2,254 2,254 1.000 FZ 1,242 1,279 1,700 0.752 14.46-153 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-013-H06 (NP 542)

Support Calc. No.: 1-P-AB-013-C7 Rev. 3

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FX 2,467 3,096 4,000 0.774 pipe clamp allowable = 4,000 #

MINUPS FX 2,467 3,096 4,000 0.774 F

[

F

[

I I

F

[

I

]

t

[

___ [

___ [ I

__ ___ [ ] [

14.46-154 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Analysis No.: Rev.7 Main Steam Calculation

Title:

Bypass Support ID: AB-013-H07 Support Calc. No.: 1-P-AB-013-C8

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 1,710 1,747 1,710 1.022 See below MINUPS FY 1,710 1,747 1,710 1.022 Minimum Margin Factor for EPU = 1.63 > 1.022 ok 14.46-155 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-013-H08 Support Calc. No.: 1-P-AB-013-C8

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FA 2,317 2,354 2,755 0.854 MINUPS FA 2,317 2,354 2,755 0.854 14.46-156 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No. : 7 Main Steam Calculation

Title:

Bypass Support ID: AB-014-H02 Support Calc. No.: 1-P-AB-014-C4

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FZ 1,121 1,158 1,700 0.681 MINUPS FZ 1,121 1,158 1,700 0.681 1

__________ _________ I- 4 -I 14.46-157 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No. : 7 Main Steam Calculation

Title:

Bypass Support ID: AB-014-H03 Support Calc. No.: 1-P-AB-014-C5

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 900 1,233 1,700 0.725 MINUPS FY 900 1,233 1,700 0.725 14.46-158 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-014-H04 Support Calc. No.: 1-P-AB-014-C6

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FX 3,254 3,698 24,000 0.154 MINUPS FX 3,254 3,698 24,000 0.154 Common with supports AB-015-H04 and AB-016-H04.

14.46-159 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-014-H06 Support Calc. No.: 1-P-AB-014-C8

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 1,808 1,882 1,767 1.065 See below Strut Allowable = 11630 MF = 11630/1882 = 6.18 > 1.0 MINUPS FY 1,808 1,882 1,767 1.065 OK

_ _ I _ I I I _ _ I _ _ I ____

Existing Minimum Margin Factor = 6.94 New Minimum Margin Factor = 6.94/1.065 = 6.49 14.46-160 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-015-H02 Support Calc. No.: 1-P-AB-015-C4

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FZ 1,685 1,722 1,700 1.026 See below MINUPS FZ 1,685 1,722 1,700 1.026 Existing Minimum Margin Factor = 8.79 New Minimum Margin Factor = 8.79/ 1.026 = 8.70 14.46-161 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-015-H03 Support Calc. No.: 1-P-AB-015-C5

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 1,000 1,370 1,700 0.806 MINUPS FY 1,000 1,370 1,700 0.806 14.46-162 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Analysis No.: Rev.7 Main Steam Calculation

Title:

Bypass Support ID: AB-015-H04 Support Calc. No.: 1-P-AB-015-C6

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FX 3,179 3,549 24,100 0.147 MINUPS FX 3,179 3,549 23,100 0.147 No material furnished with this support, plus common support AB-014-H04 14.46-163 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-015-H05 Support Calc. No.: 1-P-AB-015-C7

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 2,216 2,327 2,895 0.804 MINUPS FY 2,216 2,327 2,895 0.804 14.46-164 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-016-H02 Support Calc. No.: 1-P-AB-016-C4

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FZ 100 137 1,700 0.081 MINUPS FZ 100 137 1,700 0.081 14.46-165 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-016-H03 Support Calc. No.: 1-P-AB-016-C5

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 900 1,233 1,700 0.725 MINUPS FY 900 1,233 1,700 0.725 14.46-166 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-0019, Rev.

Analysis No.: 7 Main Steam Calculation

Title:

Bypass Support ID: AB-016-H04 Support Calc. No.: 1-P-AB-016-C6

Description:

8" Main Steam Bypass CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable MAXNOR N/A MINNOR N/A MAXUPS FX 2,056 2,356 23,400 0.101 MINUPS FX 2,056 2,356 23,400 0.101 No material furnished common with support AB-014-H04.

14.46-167 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Calculation No.: Pipe Stress Calculation C-1922, Rev.6 Calculation

Title:

Main Steam Outside Containment Support ID: AB-006-H16 (NP 139)

Support Calc. No.: 1-P-AB-006-C8 Rev. 3

Description:

Supply to Air Ejector and Steam Seal Evaporator CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 633 761 904 0.842 MINUPS FZ 606 745 775 0.961

+ 4 -I- 4 i 4 +i 14.46-168 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Calculation No.: Pipe Stress Calculation C-1922, Rev.6 Calculation

Title:

Main Steam Outside Containment Support ID: AB-006-H17 NP 137 Support CaIc. No.: 1-P-AB-006-C32 Rev. 1

Description:

Supply to Air Ejector and Steam Seal Evaporator CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FZ 862 1,063 1,500 0.709 MINUPS FZ 862 1,063 1,500 0.709 i i +

+ + +

14.46-169 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Calculation No.: Pipe Stress Calculation C-1922, Rev.6 Calculation

Title:

Main Steam Outside Containment Support ID: AB-006-H18 NP 132)

Support Calc. No.: 1-P-AB-006-C26 Rev. 2

Description:

Supply to Air Ejector and Steam Seal Evaporator CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 443 549 1,500 0.366 MINUPS FY 443 549 1,500 0.366 14.46-170 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Calculation No.: Pipe Stress Calculation C-1922, Rev.6 Calculation

Title:

Main Steam Outside Containment Support ID: AB-006-H19 (NP 120)

Support Calc. No.: 1-P-AB-006-C29 Rev. 1

Description:

Supply to Air Ejector and Steam Seal Evaporator CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 354 437 588 0.743 MINUPS FY 354 437 588 0.743 14.46-171 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Calculation No.: Pipe Stress Calculation C-1922, Rev.6 Calculation

Title:

Main Steam Outside Containment Support ID: AB-006-H20 (NP 112)

Support Calc. No.: 1-P-AB-006-C028 Rev. 2

Description:

Supply to Air Ejector and Steam Seal Evaporator CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FZ 1,065 1,323 1,250 1.058 MINUPS FZ 1,065 1,323 1,250 1.058 4 4 .4. 4 From calc. Minimum Margin Factor = 1.507 New Minimum Margin Factor = 1.507/1.058 = 1.42 > 1.0 OK 14.46-172 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Calculation No.: Pipe Stress Calculation C-1922, Rev.6 Calculation

Title:

Main Steam Outside Containment Support ID: AB-006-H21 (NP 109)

Support Calc. No.: 1-P-AB-006-C9 Rev. 3

Description:

Supply to Air Ejector and Steam Seal Evaporator CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FX 225* 89 225 0.396 MINUPS FX 225* 89 225 0.396

• Minimum design load. This is still greater than the EPU load.

14.46-173 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Calculation No.: Pipe Stress Calculation C-1922, Rev.6 Calculation

Title:

Main Steam Outside Containment Support ID: AB-006-H23 (NP 75B)

Support Calc. No.: 1-P-AB-006-C30 Rev. 2

Description:

Supply to Air Ejector and Steam Seal Evaporator CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 461 524 706 0.742 MINUPS FY 461 524 706 0.742 I I I I- -I I 14.46-174 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Calculation No.: Pipe Stress Calculation C-1922, Rev.6 Calculation

Title:

Main Steam Outside Containment Support ID: AB-006-H24 (NP 75A)

Support Calc. No.: 1-P-AB-006-C33 Rev. 3

Description:

Supply to Air Ejector and Steam Seal Evaporator CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable" MAXNOR N/A MINNOR N/A MAXUPS Hori.Skew 225* 124 260 0.477 MINUPS Hori.Skew 225* 124 260 0.477

• Minimum design load. This is still greater than the EPU load.

14.46-175 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-1922, Calculation No.: Rev.6 Calculation

Title:

Main Steam Outside Containment AB-006-H25 (NP Support ID: 75)

Support Calc. No.: 1-P-AB-006-C31 Rev. 2

Description:

Supply to Air Ejector and Steam Seal Evaporator CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS Hori.Skew 225* 125 260 0.481 MINUPS Hori.Skew 225* 125 260 0.481 1 1 1

• Minimum design load. This is still greater than the EPU loads.

14.46-176 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-1922, Calculation No.: Rev.6 Calculation

Title:

Main Steam Outside Containment Support ID: AB-006-H26 (NP 65A)

Support Calc. No.: 1-P-AB-006-C16 Rev. 3

Description:

Supply to Air Ejector and Steam Seal Evaporator CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable MAXNOR N/A MINNOR N/A MAXUPS FX 225* 160 259 0.618 MINUPS FX 225* 160 259 0.618 4 4 +

+ 4 4 +

4 + + + +

• Minimum design load. This is still greater than the EPU loads.

14.46-177 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-1922, Calculation No.: Rev.6 Calculation

Title:

Main Steam Outside Containment AB-006-H28 (NP Support ID: 58)

Support Calc. No.: 1-P-AB-006-C16 Rev. 3

Description:

Supply to Air Ejector and Steam Seal Evaporator CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FZ 225* 98 259 0.378 MINUPS FZ 225* 98 259 0.378

• Minimum Design load. This is still greater than the EPU load.

14.46-178 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-1922, Calculation No.: Rev.6 Calculation

Title:

Main Steam Outside Containment AB-006-H29 (NP Support ID: 48)

Support Calc. No.: 1-P-AB-006-C35 Rev. 2

Description:

Supply to Air Ejector and Steam Seal Evaporator CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FX 661 843 820 1.028 See below MINUPS FZ 270 336 400 0.840 f +/- 4 F Existing Minimum Margin Factor = 1.244 New Minimum Margin Factor = 1.244/1.028 = 1.21 > 1.0 OK 14.46-179 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-1922, Calculation No.: Rev.6 Calculation

Title:

Main Steam Outside Containment AB-006-H31 (NP Support ID: 38)

Support Calc. No.: 1-P-AB-006-C27 Rev. 3)

Description:

Supply to Air Ejector and Steam Seal Evaporator CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable MAXNOR N/A MINNOR _N/A MAXUPS FZ 225* 94 259 0.363 MINUPS FZ 225* 94 259 0.363 I- I- -I 4

• Minimum design load. This is still greater than the EPU load.

14.46-180 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-1922, Calculation No.: Rev.6 Calculation

Title:

Main Steam Outside Containment AB-006-H32 (NP Support ID: 16)

Support Calc. No.: 1-P-AB-006-C34 Rev. 2

Description:

Supply to Air Ejector and Steam Seal Evaporator CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FZ 225* 255 413 0.617 MINUPS FZ 225* 255 413 0.617

• Minimum design load.

14.46-181 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-1922, Calculation No.: Rev.6 Calculation

Title:

Main Steam Outside Containment AB-006-H33 (NP Support ID: 15)

Support Calc. No.: 1-P-AB-006-C13 Rev. 2

Description:

Supply to Air Ejector and Steam Seal Evaporator CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS FY 567 696 718 0.969 MINUPS FY 567 696 718 0.969 14.46-182 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-1922, Calculation No.: Rev.6 Calculation

Title:

Main Steam Outside Containment AB-006-H35 (NP Support ID: 47)

Support Calc. No.: 1-P-AB-006-C6 Rev. 3

Description:

Supply to Air Ejector and Steam Seal Evaporator CLTP EPU Design Ratio Remarks Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable MAXNOR N/A MINNOR N/A MAXUPS Hori.Skew 225* 212 260 0.815 MINUPS Hori.Skew 225* 212 260 0.815

• Minimum design load. This is still greater than the EPU load.

14.46-183 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-1922, Calculation No.: Rev.6 Calculation

Title:

Main Steam Outside Containment Support ID: AB-127-H03 (NP 95C)

Support Calc. No.: 1-P-AB-127-C5 Rev. 1

Description:

Supply to Air Ejector and Steam Seal Evaporator Remark CLTP EPU Design Ratio s Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FZ 479 590 551 1.071 See below MINUPS FZ 479 590 551 1.071 F I + +

Existing Minimum Margin Factor = 31.25 New Minimum Margin Factor = 31.25/1.071 = 29.18 > 1.0 OK 14.46-184 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-1922, Calculation No.: Rev.6 Calculation

Title:

Main Steam Outside Containment Support ID: AB-127-H04 (NP 95B)

Support Calc. No.: 1-P-AB-127-C6 Rev. 2

Description:

Supply to Air Ejector and Steam Seal Evaporator Remark CLTP EPU Design Ratio s Load Case Load Load Load Load EPUI/

Direction (Lbs) (Lbs) (Lbs) Design Acceptable" MAXNOR N/A MINNOR N/A MAXUPS FY 225* 175 259 0.676 MINUPS FY 225* 175 259 0.676

-I I- -I I-

• 1 I. -I +

• Minimum design load. This is still greater than the EPU loads.

14.46-185 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-1922, Calculation No.: Rev.6 Calculation

Title:

Main Steam Outside Containment Support ID: AB-127-H05 (NP 100)

Support Caic. No.: 1-P-AB-127-C3 Rev. 2

Description:

Supply to Air Ejector and Steam Seal Evaporator Remark CLTP EPU Design Ratio s Load Case Load Load Load Load EPU/ UNITS: lbs and Ft-lbs.

Direction (Lbs) (Lbs) (Lbs) Design UPSET FX 761* 981

• Acceptable" Min/Max FY 259* 313

• FZ 127* 152

• 242*

MX ft.# 279

• 360*

MY ft.# 426

• 1210*

MZ ft.# 1,457

• Note:

• Symbolizes actual loads per latest analysis, but not used for design.

This support is evaluated for plastic forces = 10108 # and plastic moments = 6802 ft-lbs.

14.46-186 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-1922, Calculation No.: Rev.6 Calculation

Title:

Main Steam Outside Containment Support ID: AB-128-H05 (NP 180)

Support Calc. No.: 1-P-AB-128-C2 Rev. 2)

Description:

Supply to Air Ejector and Steam Seal Evaporator Remark CLTP EPU Design Ratio s Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design UPSET FX 479* 585 * "Acceptable" Min/Max FY 130* 161

• FZ 363* 449

• MX 56* ft.# 67

• 1336*

MY ft.# 1,657

• 186*

MZ ft.# 228 *

-4 4 Note:

• Symbolizes actual loads per latest analysis, but not used for design.

This support is evaluated for plastic forces = 10108 # and plastic moments = 6802 ft-lbs.

14.46-187 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-1922, Calculation No.: Rev.6 Calculation

Title:

Main Steam Outside Containment Support ID: AB-128-H07 (NP 160)

Support Calc. No.: 1-P-AB-128-C4 Rev. 1

Description:

Supply to Air Ejector and Steam Seal Evaporator Remark CLTP EPU Design Ratio s Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable MAXNOR N/A MINNOR N/A Vert.

MAXUPS Skew 447 549 488 1.125 See below Vert.

MINUPS Skew 447 549 488 1.125 Existing Minimum Margin Factor = 49.3 New Minimum Margin Factor = 49.3/1.125 = 43.8 > 1.0 OK 14.46-188 LR-N07-0099 LCR H05-01, Rev. 1 Project: Hope Creek Extended Power Uprate Pipe Stress Calculation C-1922, Calculation No.: Rev.6 Calculation

Title:

Main Steam Outside Containment Support ID: AB-128-H08 (NP 161)

Support Calc. No.: 1-P-AB-128-C5 Rev. 2

Description:

Supply to Air Ejector and Steam Seal Evaporator Remark CLTP EPU Design Ratio s Load Case Load Load Load Load EPU/

Direction (Lbs) (Lbs) (Lbs) Design "Acceptable MAXNOR N/A MINNOR N/A Vert.

MAXUPS Skew 487 599 534 1.122 See below Vert.

MINUPS Skew 487 599 534 1.122

-I. 4 4 4 4 Existing Minimum Margin Factor = 1.6 New Minimum Margin Factor = 1.6/1.122 = 1.43 > 1.0 OK 14.46-189 LR-N07-0099 LCR H05-01, Rev. 1 The following supports were modified as shown in remarks column:

Remarks Support Mk No.

1-P-AB-001-H009 Upgrade weld size New Strut, Clamp & Rear Bracket, 1-P-AB-001-H013 Upgrade weld New Strut, Clamp & Rear Bracket, 1-P-AB-001-H014 Upgrade weld 1-P-AB-002-H013 Add stiffener plates & new welds 1-P-AB-003-H009 Upgrade weld 1-P-AB-004-H004 Upgrade weld and add additional weld 14.46-190