BSEP 03-0128, Response to Request for Additional Information Core Flow Operating Range Expansion

Response to Request for Additional Information Core Flow Operating Range Expansion

ML032530423
Person / Time
Site:	Brunswick
Issue date:	09/03/2003
From:	Keenan J Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 03-0128, TAC MB6692, TAC MB6693, TSC-2002-09
Download: ML032530423 (15)

Text

0071 ~~~~~~~~~~~~~~~~~~~~~~~~~~John s 3 Progress Energy ic S.Keenan Puesde lant Brunswick Nuclear Plant Progress Energy Carolinas, Inc.

SEP 0 3 2003 SERIAL: BSEP 03-0128 TSC-2002-09

/ Nuclear Regulatory Commission KTrN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Docket Nos. 50-325 and 50-324/License Nos. DPR-71 and DPR-62 Response to Request for Additional Information Core Flow Operating Range Expansion (NRC TAC No. MB6692 and MB6693)

Reference:

1. Letter from John S. Keenan to the U. S. Nuclear Regulatory Commission (Serial: BSEP 02-0169), "Request for license Amendments - Core Flow Operating Range Expansion," dated November 12, 2002

2. Letter from Brenda L. Mozafari, U. S. Nuclear Regulatory Commission, Senior Project Manager, Section 2, to John S. Keenan, "Request for Additional Information On Amendment Request to Expand Core Flow Operating Range (MELLLA+)," dated July 21, 2003 Ladies and Gentlemen:

On November 12, 2002, Progress Energy Carolinas, Inc. (PEC) requested a revision to the Technical Specifications (TSs) for the Brunswick Steam Electric Plant (BSEP), Units 1 and 2. The proposed license amendments revise TSs, as necessary, to support an expansion of the core flow operating range (i.e., Maximum Extended Load Line Limit Analysis Plus (MELLLA+)).

On July 21, 2003, the NRC provided a written request for additional information (RAI) concerning the anticipated transient without scram (ATWS) analysis performed in support of the amendment request. The RAI requested a response within 45 days, which corresponds to September 4, 2003. The response to this RAI is enclosed.

Enclosure 1 contains information that General Electric Company (GE) considers to be proprietary as defined by 10 CFR 2.790. GE, as the owner of the proprietary information, has executed the affidavit provided in Enclosure 2, which identifies that the enclosed PO. Box 10429 Southport, NC 28461 T> 910.457.2496 F> 910.457.2803

Document Control Desk BSEP 03-0128/ Page 2 proprietary information has been handled and classified as proprietary, is customarily held in confidence, and has been withheld from public disclosure. The proprietary information was provided to PEC in a transmittal that is referenced by the affidavit. The proprietary information has been faithfully reproduced such that the affidavit remains applicable. GE requests that the enclosed proprietary information be withheld from public disclosure in accordance with the provisions of 10 CFR 2.790 and 9.17. A non-proprietary (i.e., redacted) version of the response is provided in Enclosure 3.

Please refer any questions regarding this submittal to Mr. Edward T. ONeil, Manager - Support Services, at (910) 457-3512.

Sincerely, 0, ant~a MAT/mat

Enclosures:

1. Response to Request for Additional Information (RAI) 4 - Proprietary

2. General Electric Company Affidavit of Proprietary Information

3. Non-Proprietary Version of Response to Request forAdditional Information (RAI) 4 C. J. Gannon, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are officers, employees, and agents of Carolina Power & Light Company.

Notary (Seal)

My commission expires: 3 , LOO po

Document Control Desk BSEP 03-0128 / Page 3 cc: (with Enclosures except as noted)

U. S. Nuclear Regulatory Commission, Region II ATrN: Mr. Luis A. Reyes, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTN: NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)

ATN: Ms. Brenda L. Mozafari (Mail Stop OWFN 8G9) 11555 Rockville Pike Rockville, MD 20852-2738 Ms. Jo A. Sanford (w/o Enclosure 1)

Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 Ms. Beverly 0. Hall, Section Chief (w/o Enclosure 1)

Radiation Protection Section, Division of Radiation Protection North Carolina Department of Environment and Natural Resources 3825 Barrett Drive Raleigh, NC 27609-7221

BSEP 03-0128 Enclosure 2 General Nuclear Fuels Affidavit of Proprietary Information

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

(1) I 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 Attachment 2 to GE letter GE-KBO-AEP-439P, Carl Hinds (GE) to Bob Kitchen (Progress Energy), Brunswick AEP, MELLLA+ RAI 4, HC Temperature Limit (HCTL), dated August 29, 2003.

The proprietary information in Attachment 2, GE Response to NRC RAI 4 is identified by a double underline inside double square brackets. In each case, the superscript notation refers to Paragraph (3) of the enclosed 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 2.790(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 Project v. Nuclear Regulatory Commission.

975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2dl280 (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;

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; GBS-03-08-af CP&L M+ RAI 8-29-03.doc Affidavit Page

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 2.790 (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 by the 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 results and conclusions from evaluations of the safety-significant changes necessary to demonstrate the regulatory acceptability for the expended power/flow range of MELLLA+ for a GE BWR, utilizing analytical models and methods, including computer codes, which GE has developed, obtained NRC approval of, and applied to perform evaluations of transient and accident events in the GE Boiling Water Reactor ("BWR"). The development and approval of these system, component, and thermal hydraulic models and 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-03-08-af CP&L M+ RAI 8-29-03.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 2i9 day of 2003.

. /1S GBS-03-08-af CP&L M+ RAI 8-29-03.doc Affidavit Page 3

BSEP 03-0128 Enclosure 3 Page 1 of 8 Non-Proprietary Version of Response to Request for Additional Information (RAI) 4 Backeround On November 12, 2002, Progress Energy Carolinas, Inc. requested a revision to the Technical Specifications (TSs) for the Brunswick Steam Electric Plant (BSEP), Units 1 and 2. The proposed license amendments revise TSs, as necessary, to support an expansion of the core flow operating range (i.e., Maximum Extended Load Line Limit Analysis Plus (MELLLA+)).

On July 21, 2003, the NRC provided a written request for additional information (RAI) concerning the anticipated transient without scram (ATWS) analysis performed in support of the amendment request. The response to this RAI follows.

NRC Question 4-1 On November 12, 2002, Carolina Power & Light Company (CP&L), the licensee for the Brunswick Steam Electric Plant (BSEP), Units 1 and 2, submitted a request to revise the Technical Specifications, as necessary, to support an expansion of the core flow operating range (i.e., Maximum Extended Load Line Limit Analysis Plus (MELLLA+) following the extended power uprate (EPU). The MELLLA+ application (NEDC-33063P) states that the anticipated transient without scram (ATWS) suppression pool temperature limit is 207.70 F (See table in Section 9.3.1 of NEDC-33063P). NEDC-33063P also indicates that the suppression pool design limit is 2200F. However, page 20 of the CP&L EOP document (OEOP-01-NL, Rev. 12) shows the Heat Capacity Temperature Limit (HCTL) plot as function of pressure. For a pressure of 1150 psig, the HCTL is between 120 0F and 1680 F, depending on the assumed suppression pool water level. In addition, page 109 of OEOP-01-NL, Rev. 12, states that the suppression pool temperature requiring reactor depressurization is 120'F. It appears that the safety limits stated by NEDC-33063P and OEOP-01-NL, Rev. 12, are contradictory. The following questions address capability of BSEP, Units 1 and 2 to maintain containment integrity during ATWS for the EPU/MELLLA+ operation.

Justify the use of the 207.7F ATWS suppression pool temperature limit for the EPUIMELLLA+

ATWS analysis. Specifically, justify why is the suppression pool temperature limit higher than the temperature limit requiring reactor depressurization.

Response to NRC Question 4-1 The ATWS analysis as described in NEDC-32424P-A (ELTR-1), February 1999, and NEDC-32523P-A (ELTR-2), February 2000, and its Supplement 1, Volumes I and II could be described as (( ] analysis. It was not intended that ATWS analysis for justification of a power uprate, MELLLA+, or similar plant design change be a [(

)) there is a relatively small impact on the

BSEP 03-0128 Enclosure 3 Page 2 of 8 ATWS response for these changes and these events are generally not limiting in the overall plant licensing basis. The small impact on plant response due to these changes is a result of the mitigating actions that are specified for the ATWS events. For example, there is initially a dramatic power reduction as a result of the flow runback, followed by power reductions as a result of reactor water level reduction and boron injection.

]1 and the fact that subsequent changes affecting the ATWS analysis (e.g., power uprates) have been reviewed and accepted by the NRC. Hence, this approach is appropriate since it is known that operation for current plant conditions is acceptable as the base case. The (( )) approach only makes sense if it is known that operation for current plant conditions is acceptable as the base case. The symptomatic Emergency Procedure Guidelines (EPG) instructions for ATWS include rapid reactor depressurization to maintain plant conditions within the specification of the heat capacity temperature limit (HCTL) that is not included in the EPU/MELLLA+ ATWS analysis. The Boiling Water Reactor Owners' Group (BWROG) and NRC agreed to a process by which differences between EPG actions and licensing basis assumptions could be reconciled. This is an inherent recognition that licensing basis calculations do not necessarily include all actions specified in the EPGs.

)) This is based on plant compliance with the requirements of the ATWS Rule (10 CFR 50.62) and the BWROG analysis that was performed to support approval of the EPGs for ATWS (OEI Document 9402-3, Revision 1,The Management of ATWS by Boron Injection and Water Level Control, June 1994). General Electric (GE) methods were not used directly when the symptomatic EPGs were expanded to include ATWS conditions, thus ((

)) Therefore, the ((

)) approach was developed for ATWS and was applied to EPU/MELLLA+ plant design changes. The applicability of EPGs to EPU was confirmed by an evaluation performed for the BWROG Emergency Procedures Committee (KLR Services document KLR-1024-02, Power Uprate and Fuel Reload Effects on EOPs and SAGs, May 2002).

1[

)) The AJTWS event, methods, assumptions, and conditions are specified commensurate with the peak suppression pool temperature acceptance criterion. [

1] Please note that, in the EPGs, the HCTL only places a limit on suppression pool temperature when the reactor is pressurized. When the reactor is depressurized below a certain level, the HCTL places no limit on the maximum suppression pool temperature. Therefore, even if depressurization at the HCTL were included in the

i BSEP 03-0128 Enclosure 3 Page 3 of 8 EPU/MELLLA+ analysis for ATWS, a peak suppression pool temperature criterion would still need to be specified.

In summary, the specified power uprate ATWS analysis [1

• ]

• The small impact of changes such as EPU/MELLLA+ on the ATWS response,

• 1[1

• Conservative models and assumptions used in the analysis, and

• Specification of a conservative acceptance criterion commensurate with the specified model.

In conclusion, since reactor depressurization at the HCTL is not part of the ATWS calculations for EPU/MELLLA+, it is acceptable to have a suppression pool temperature licensing analysis limit that is higher than where the HCTL specifies that reactor depressurization must be initiated.

NRC Ouestion 4-2 Provide the basis for the ATWS suppression pool design limit of 220 0F quoted in NEDC-33063P.

Response to NRC Question 4-2 Reviews of the suppression chamber design indicate that 2200 F is the design temperature for the suppression chamber. The July 31, 1968, Preliminary Safety Analysis Report Table V-2-1, Primary Containment System Principal Design Parameters and Characteristics, lists the design temperature of pressure-suppression chamber as 220'F. The original suppression chamber fabrication drawings for BSEP Unit I and BSEP Unit 2 specify a design temperature of 2200F.

The original containment structural design analysis was based on a pressure of 62 psig and temperatures of 300OF and 2200 F for the drywell and suppression chamber respectively. The drywell and suppression chamber steel containment pressure vessels and connecting vent system were procured and fabricated to those design values. Subsequent to the original analysis, a more detailed analysis was performed in BSEP Design Report No. 7, dated December 22, 1970, in response to NRC questions and comments. This analysis showed that the American Society of Mechanical Engineers (ASME) Code allowables would not be exceeded at a temperature of less than 220'F.

BSEP 03-0128 Enclosure 3 Page 4 of 8 The suppression chamber design temperature of 220F was most recently approved by the NRC, on May 27, 1998, in amendments 195 and 225 to the BSEP Units 1 and 2 technical specifications, respectively.

NRC Ouestion 4-3 The peak suppression pool temperature for EPU/MELLLA+ reported in NEDC-33063P is 197.70F. While this number is below the 207.70 F LOCA limit, the reactor is still at full pressure.

Thus, the reported 197.7 0 F is not the peak temperature, but the initial condition prior to the reactor depressurization. Following a depressurization (which is required by the EOP for temperature of 197.70 F), the suppression pool temperature would be greater than 207.7°F.

Please provide the actual peak suppression pool temperature when the ATWS transient is followed to completion according to the EOPs.

Response to NRC Question 4-3 As described in the response to RAI-1, the ATWS analysis is ((

] Based on previous studies sponsored by the BWROG, it is believed that the reactor only produces decay heat during the reactor depressurization, and returns to a lower generated power level, given the same reactor water level and boron concentration, following the depressurization. This can be explained in first principles terms by the differences between hfg and vg from rated reactor pressure to a depressurized state. In addition, the depressurization would provide mixing for any boron that had been injected into the reactor and provide a further power reduction.

The EPG-licensing basis reconciliation process, simply stated is: the licensing basis methodology is acceptable if it is more conservative that a realistic analysis of the licensing basis event with all EPG specified operator actions. It is believed that realistic modeling of the ATWS response, including realistic boron mixing assumptions, in conjunction with all EPG specified operator actions, would produce a lower peak suppression pool temperature than the ATWS licensing basis analysis specified to justify acceptability of EPU/MELLLA+ and similar plant changes.

NRC Ouestion 44 Provide the assumptions used in the ATWS analysis for the EPU/MELLLAh specific calculations reported in NEDC-33063P. Specifically, identify which ATWS transient is limiting in terms of each ATWS acceptance limit. Describe the initial conditions, including power, flow, suppression pool level used. Identify the operator actions that are assumed. Provide the ATWS mitigation actions that are implemented during the transient. Provide the values that are used for the EOP variables (e.g., HCTL, HSBW, etc).

BSEP 03-0128 Enclosure 3 Page 5 of 8 Response to NRC Question 44 The limiting ATWS events evaluated for the Brunswick MELLLA+ submittal are Closure of All MSIVs (MSIVC) and Pressure Regulator Failure Open - Maximum Steam Demand (PRFO).

The results of the MELLLA+ analysis show that the PRFO is the limiting event for the vessel overpressure and peak cladding temperature criteria, and the MSIVC is the limiting event for suppression pool temperature and containment pressure criteria. The initial conditions assumed in the analysis are provided in Table 1.

The assumed operator actions for ATWS mitigation include lowering of water level, initiation of the standby liquid control system, and increasing of water level after injection of hot shutdown boron weight. The hot shutdown boron weight is based on the generic 522 ppm. The NRC approval of the ODYN code application to ATWS analysis is provided in "Qualification of the One-Dimensional Core Transient Model (ODYN) for Boiling Water Reactors," NEDC-24154P-A, Supplement 1, Volume 4, February 2000. ((

1]:

lble 1- Initial Conditions forATWS Analysis _____I_-^:

Ke UParaeter Unit M+Value Dome Pressure psig 1030 Rated Core Flow Mlbmlhr 77.0 Minimum Core Flow at Rated Thermal Power (RTP)  % rated 85 Minimum Core Flow at RTP Mlbm 65.5 RTP MWt 2923 Rated Steam Flow Mlbm/hr 12.77 Feedwater Temperature OF 431 Initial Suppression Pool Liquid Volume ft3 86,450 Initial Suppression Pool Temperature OF 95 Initial Suppression Pool Mass (Note 1) Mlbm 5.365

BSEP 03-0128 Enclosure 3 Page 6 of 8 Table 1- Initial Conditions for ATWS Analysis Key Pramer Unit M+ Value Initial Inventory in CST Mlbm 0.821 Initial Inventory in CondenserlHotwell Mlbm 0.525 Note 1: The pool mass corresponds to low water level condition.

NRC Question 4-5 Provide a plot of suppression pool temperature versus time (short-term and long-term).

Response to NRC Question 4-5 This information has been provided to the NRC by GE letter, MFN 03-056, "MELLLA Plus LTR RAI ATWS and Containment Data, Parts 1, 3 and 4 (TAC No. MB6157)." See ADAMS Accession Numbers ML032130614, ML032130611, and ML032130649.

NRC Question 4-6 Provide a plot of reactor pressure and power versus time during the event.

Response to NRC Question 4-6 This information has been provided to the NRC by GE letter, MFN 03-056, MELLLA Plus LTR RAI ATWS and Containment Data, Parts 1, 3 and 4 (TAC No. MB6157)." See ADAMS Accession Numbers ML032130614, ML032130611, and ML032130649.

NRC Ouestion 4-7 Provide a table showing the sequence of events (system and component actuations and operator actions) during the pressure regulator failure to open (PRFO) and Main Steam Isolation Valve Closure (MSIVC) ATWS events analyzed. Show the safety/relief valve actuation setpoints and indicate the time each safety/relief valve group first actuates.

Response to NRC Question 4-7 Tables 2 and 3 provide the event sequence for an MSIVC event and a PRFO event respectively.

The assumed safety/relief valve (SRV) opening setpoints are 1174 psig, 1184 psig, and 1194 psig for the three valve groups. Table 4 provides the SRV opening timing for an MSIVC event and a PRFO event. It is noted that the difference in opening times among the valves in the same setpoint group is due to the implementation of statistical spread of the opening setpoints.

BSEP 03-0128 Enclosure 3 Page 7 of 8

-Table 2-EetSqnc fo a MSVC Event item Repne-M+ Event Time (sec) 1 [ ___________________

2 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

3 ____________________________________________

4 _____________________________________

5 _________________________________________

6 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

7 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

89 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

9 1_ _ _ _ _ _ _ _ _ _ _

1 0 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

14 Tbe 3' Event Sequencefo a PRFO Event-:

Item R~~~esponse -W EetTme (sec) 12 I_ _ _ _ _ _ _ _ _ _ _ _ _ _

23 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

34 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

4 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

9 1 0 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

11 1 2 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

1 3 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

1 4 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

BSEP 03-0128 Enclosure 3 Page 8 of 8 Table 3- Event Sequence for a PRFO Event Item -vent I mCe (sec) 15 16 _ ]

Tb 4 I-nitial -ale Opening Timing for an, MsIVC Eventand a PRFO Event Valve MSIVC- Event ~PRFO Event 13__________________ ____________________

24_ _ _ _ _ _ _ _ _ _ _ _ _

3 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

6 _ _ _ _ _ _ _ _ _ _ _ _ _

7 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

9 _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _

to_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

S _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _