Text

Technical Specification Change Request (TSCR-059): Adoption of Generic Letter 96-03: Relocation of Pressure Temperature Limit Curves & Low Temperature Overpressure Protection System Limits
	ML030710261
Person / Time
Site:	Duane Arnold
Issue date:	02/28/2003
From:	Peifer M; Nuclear Management Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	GL-96-003, NG-03-0123, TAC MA9891, TAC MB0394
	Download: ML030710261 (61)
	v • d • e

NMc Committed to Nuclear Excellence February 28, 2003 NG-03-0123 Duane Arnold Energy Center Operated by Nuclear Management Company, LLC 10 CFR 50.90 Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station O-P 1-17 Washington, DC 20555-0001

Subject:

References:

File:

Duane Arnold Energy Center Docket No: 50-331 Op. License No: DPR-49 Technical Specification Change Request (TSCR-059): "Adoption of Generic Letter 96-03: 'Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits'."

1) Letter, B. Mozafari (USNRC) to G. VanMiddlesworth (NMC),

"DUANE ARNOLD ENERGY CENTER - ISSUANCE OF AMENDMENT RE: REVISED PRESSURE-TEMPERATURE CURVES (TAC NO. MB0394)," April 30, 2001.

2) Letter, S. A. Richards (USNRC) to J. F. Klapproth (GE), "SAFETY EVALUATION FOR NEDC-32983P, 'GENERAL ELECTRIC METHODOLOGY FOR REACTOR PRESSURE VESSEL FAST NEUTRON FLUX EVALUATION' (TAC NO. MA9891),"

September 14, 2001.

A-117 In accordance with the Code of Federal Regulations, Title 10, Sections 50.59 and 50.90, Nuclear Management Company, LLC (NMC) hereby requests revision to the Technical Specifications (TS) for the Duane Arnold Energy Center (DAEC). The proposed change relocates the vessel pressure and temperature (P/T) limit curves and other related numerical values from the TS to a licensee-controlled document, pursuant to NRC Generic Letter (GL) 96-03. The proposed changes also incorporate an NRC-approved generic change to the Improved Standard Technical Specifications (ISTS) for BWR/4, NUREG-1433, Revision 2; specifically, Technical Specification Task Force (TSTF) change TSTF-419, Revision 0.

In Reference 1, the Staff approved the existing P/T limit curves for the DAEC. However, as noted in the accompanying Staff's Safety Evaluation (SE), such approval was granted on an interim basis until such time as the Staff could complete its review and approval of the General Electric Company's (GE) topical report (NEDC-32983P) which describes their methodology for calculating the neutron fluence values used in generating the subject P/T limit curves. The Staff has subsequently completed its review and issued its SE approving NEDC-32983P (Reference 2).

3277 DAEC Road 0 Palo, Iowa 52324-9785 0,0 Telephone 319 851 7611

February 28, 2003 NG-03-0123 Page 2 One of the prerequisites for adopting GL 96-03 is that the P/T limits be calculated using "NRC-approved methods." As stated above, at the time of our last revision of the DAEC P/T limit curves, the GE methods used to calculate the neutron fluence were not yet approved by the Staff. The remaining methods used to generate the P/T limits were found in Reference 1 to meet the Staff's requirements under 10 CFR 50, Appendix G and associated guidance documents, such as Regulatory Guide 1.99, Rev. 2. Given that the Staff has now approved the GE fluence methods, the GL prerequisite is now satisfied. Consequently, NMC proposes to modify the DAEC TS pursuant to NRC Generic Letter 96-03 (as modified by TSTF-419, Rev. 0) and relocate the P/T limits to a licensee-controlled document, the Pressure and Temperature Limits Report (PTLR).

The proposed changes in the attached submittal incorporate the remaining GL prerequisites.

The TS are modified, pursuant to the GL (as revised by TSTF-419, Rev. 0), to incorporate the definition of the PTLR and the Administrative Controls Section has been modified to include the necessary restrictions and reporting requirements for the PTLR. The other affected TS line items are modified to show the relocation of the P/T limit numerical values and curves to the PTLR.

Also, as required by GL 96-03, a copy of our proposed PTLR is attached for the Staff's review. As noted in Reference 2, approval of the GE methodology was granted, provided that a specified bias is applied to the result of the calculations. GE has subsequently updated the P/T limit curves for the DAEC, applying this required bias to the previously-calculated fluence values. Thus, the revised curves comply with the approved methodology. The new, revised P/T limit curves are incorporated into the initial issuance of the DAEC PTLR.

This application has been reviewed by the DAEC Operations Committee and the Offsite Review Committee. A copy of this submittal, along with the 10 CFR 50.92 evaluation of "No Significant Hazards Consideration," is being forwarded to our appointed state official pursuant to 10 CFR Section 50.91.

To support implementation of this amendment, prior to the expiration date of the existing P-T limit curves on September 1, 2003, NMC requests that the NRC review and approve this license amendment request by August 15, 2003.

There are no new commitments being made in this letter.

February 28, 2003 NG-03-0123 Page 3 This letter is true and accurate to the best of my knowledge and belief.

Nuclea n

nt Company, LLC By k*,*

Mh$rk Peif6l DAEC Site Vice President State of Iowa (County) of Linn Signed and swom to before me on this c*

day of /'[1-ý rLtL

,2003, by UoL *. Pei-lr

.]

No(ary Public4in and for the State of Iowa MYFI FR10s "Commission Expires Attachments:

1. Evaluation of Change Pursuant to 10 CFR Section 50.92

2. Proposed Change TSCR-059 to the Duane Arnold Energy Center Technical Specifications

3. Safety Assessment

4. Environmental Consideration

5. Duane Arnold Energy Center Pressure And Temperature Limits Report, Revision 0 cc:

R. Browning (w/a)

D. Hood (NRC-NRR) (w/a)

J. Dyer (Region III) (w/a)

D. McGhee (State of Iowa) (w/a)

NRC Resident Office (w/a)

IRMS (w/a) to NG-03-0123 Page 1 of3 EVALUATION OF CHANGE PURSUANT TO 10 CFR SECTION 50.92

Background:

This proposed amendment request relocates the Reactor Coolant System (RCS) pressure and temperature (P/T) limits numerical values and associated figure from the Duane Arnold Energy Center (DAEC) Technical Specifications (TS) to a licensee-controlled document, the Pressure and Temperature Limits Report (PTLR). This relocation and creation of the associated TS administrative controls on the PTLR are in conformance to NRC guidance as proposed in Generic Letter 96-03, as modified by an approved generic change to the NRC Improved Standard TS, Technical Specification Task Force (TSTF) change package number 419, Revision 0. TSTF 419, Rev. 0 was approved by the Staff, as communicated in its letter, W. Beckner (USNRC) to A.

Pietrangelo (NEI), March 21, 2002.

Nuclear Management Company. LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa Date of Amendment Request: February 28, 2003 Description of Amendment Request:

The proposed amendment will revise the Technical Specifications (TS) to hdopt the Pressure and Temperature Limits Report (PTLR), as outlined in Generic Letter 96-03, as modified by Technical Specification Task Force (TSTF) change package number 419, Revision 0.

Specifically, a definition for the PTLR will be added to TS Section 1.0 (Definitions), and administrative controls for the generation and reporting requirements associated with the PTLR will be added to TS Section 5.6 (Reporting Requirements). In addition, the Limiting Condition for Operation (LCO) and Surveillance Requirements (SR) for Reactor Coolant System (RCS)

Pressure and Temperature Limits (LCO 3.4.9) are being modified to remove the numerical values and associated curve (Figure 3.4.9-1) for the various P/T limits, which have been updated using an NRC-approved methodology, and replace them with the reference to the PTLR.

Basis for proposed No Significant Hazards Consideration:

The Commission has provided standards (10 CFR Section 50.92(c)) for determining whether a significant hazards consideration exists. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

After reviewing this proposed amendment, NMC has concluded:

1) The proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.

Attachment I to NG-03-0123 Page 2 of 3 The P/T limits are not derived from Design Basis Accident (DBA) analyses. They are prescribed by the ASME Code and 10 CFR 50 Appendix G and H as restrictions on normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause non-ductile failure of the reactor coolant pressure boundary. Thus, they ensure that an accident precursor is not likely. Hence, they are included in the TS as satisfying Criterion 2 of 10 CFR 50.36(c)(2)(ii). The relocation of the numerical value of these limits to a licensee-controlled document does not remove the existing TS requirement that the limits be met. The new TS administrative controls for the PTLR will ensure that only NRC-approved methods are used to calculate the actual limifs to be applied. Thus, this relocation will not increase the probability of any accident previously evaluated.

The proposed changes do not alter the design assumptions, conditions, or configuration of the facility or the manner in which the facility is operated or maintained. The proposed changes will not affect any other System, Structure or Component (SSC) designed for the mitigation of previously analyzed events. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. Thus, the proposed relocation of the existing numerical values and the updated figure for the RCS P/T limits based upon an NRC-approved methodology, to a licensee-controlled document (i.e., the PTLR), with all the requisite TS restrictions placed upon it by NRC Generic Letter 96-03, as modified by TSTF-419, Rev. 0, will not increase the consequences of any previously evaluated accident.

2) The proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements or eliminate any existing requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice. We are merely requesting to move the existing numerical values and the updated figure for the RCS P/T limits based upon an NRC-approved methodology, from the TS to a licensee controlled document (i.e., the PTLR), with all the requisite TS restrictions placed upon it by NRC Generic Letter 96-03, as modified by TSTF-419, Rev. 0.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3) The proposed amendment will not involve a significant reduction in a margin of safety.

The proposed changes do not alter the manner in which Safety Limits, Limiting Safety System Settings or Limiting Conditions for Operation are determined. The setpoints at which protective actions are initiated are not altered by the proposed changes. Sufficient equipment remains available to actuate upon demand for the purpose of mitigating an analyzed event. We are merely requesting to move the existing numerical values and the updated figure for the RCS P/T limits based upon an NRC-approved methodology, from the TS to a licensee-controlled document (i.e.,

to NG-03-0123 Page 3 of 3 the PTLR), with all the requisite TS restrictions placed upon it by NRC Generic Letter 96-03, as modified by TSTF-419, Rev. 0. Thus, the proposed changes will not significantly reduce any margin of safety that currently exists.

Based upon the above, NMC has determined that the proposed amendment will not involve a significant hazards consideration.

Attorney for Licensee: Jonathan Rogoff, Esquire, General Counsel, NMC, LLC, 700 First St.,

Hudson, WI, 54016.

to NG-03-0123 Page 1 of I Proposed Change TSCR-059 to the Duane Arnold Energy Center Technical Snecifications The holders of license DPR-49 for the Duane Arnold Energy Center propose to amend the Technical Specifications (TS) by deleting the referenced pages and replacing them with the enclosed new pages. Both "pen & ink" markups of the existing TS pages and corresponding clean, typed revisions are provided.

Page Description of Changes 1.1-5 Add a definition for the Pressure and Temperature Limits Report (PTLR).

3.4-20 Add reference to the PTLR in the Limiting Condition for Operation (LCO) statement of LCO 3.4.9.

3.4-21, Revise Surveillance Requirements 3.4.9.1 through 3.4.9.7 to remove the 3.4-22, and numerical values and references to Figure 3.4.9-1 and replace them with a 3.4-23 reference to the PTLR.

3.4-24 Delete Figure 3.4.9-1.

5.0-21 and Add new Section 5.6.7 for the administrative controls on the PTLR.

22 The "pen & ink" markups of the BASES pages are provided for "Information Only" and will be incorporated during implementation of this amendment under TS 5.5.10, BASES Control Program.

Definitions 1.1 1.1 Definitions (continued)

MINIMUM CRITICAL POWER RATIO (MCPR)

(continued)

MODE OPERABLE - OPERABILITY RATED THERMAL POWER (RTP)

REACTOR PROTECTION SYSTEM (RPS)

RESPONSE

TIME film boiling occur intermittently with neither type being completely stable.

A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 1912 MWt.

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

(continued)

Amendment.,4*

(fTcAL-.0~

DAEC 1.1-5 I

Technical Specification Mark-up Inserts For TSCR-059 (PTLR)

Insert to Definitions PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.

RCS P/T Limits 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 RCS Pressure and Temperature (P/T) Limits LCO 3.4.9 APPLICABILITY:

RCS pressure. RCS temperature. RCS heatup and cooldown rates, and the recirculation pump st.arting temperature requirements shall be maintained within limits/

At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. --------NOTE --------

A.1 Restore parameter(s) 30 minutes Required Action A.2 to within limits.

shall be completed if this Condition is AND entered.

A.2 Determine RCS is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months acceptable for Requirements of the continued operation.

iCO not met in MODE 1.

2. or 3.

B. Required Action and B.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND not met.

B.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued) 3.4-20 Amendment 12J C-rSCL-0 S-)

DAEC

RCS P/T Limits 3.4.9

)

SURVEILLANCE REQUIREMENTS

)

SR 3.4.9.1

NOTE -------------------

Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

RCS pressurev_.

RCS temperaturear~e withi th

-lmit RCS atup an oo own ates are

-I.

C in nq knitir nuam 'M Arliinn 30 minutes (continued)

Amendment

(-.sc-oA)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

NOTE-------- C.1 Initiate action to Immediately Required Action C.2 restore parameter(s) shall be completed if to within limits.

this Condition is entered.

AND C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 LCO not met in other operation.

or 3.

than MODES 1. 2.

and 3.

DAEC 3.4-21

RCS P/T Limits 3.4.9 r

3.4-22 Amendment C 75C.-IL-01)

)

DAEC SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.9.2 Verify RCS pressure and RCS temperature are Once within within the criticality limits specified it) 15 minutes

....e 8.4.-9t.

prior to control rod

-withdrawal for the purpose of achieving criticality SR 3.4.9.3 ------------------

NOTE-------------

Only required to be met in MODES 1. 2, 3.

and 4 during recirculation pump startup.

Verify the difference between the bottom Once within head coolant temperature and the Reactor 15 minutes Pressure Vessel (RPV) coolant temperature prior to each is -

.startup of a reci rcul ati on

____-____*e_

pump SR 3.4.9.4 -----------------

NOTE--------------

Only required to be met in MODES 1. 2. 3.

and 4 during recirculation pump startup.

Verify the difference between the reactor Once within coolant temperature in the reci rcul ation 15 minutes loop to be started and the RPV coolont, prior to each temperature is.-*

startup of a uocfk%

tke.., U.vt reci rcul ati on "r

ýý;e,.

t pump (continued)

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.4.9.5

NOTE--------------

Only required to be performed when tensioning the reactor vessel head bolting studs.

Verify temperatures at the reactor vessel head flange and the shell adjacent to the head flange are >

SR 3.4.9.6

NOTE--------------

Not required to be performed until 30 minutes after RCS temperature : 80°F in MODE 4.

Verify temperatures at the reactor vessel head flange and the shell adjacent to the head flange are >

Y FREQUENCY 4.

30 minutes 30 minutes SR 3.4.9.7 NOTE-------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS temperature 5 100°F in MODE 4.

Verify temperatures at the reactor vessel 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months head flange and the shell adjacent to the head flange are*

A-

L

& +/-k ?1Th Amendmentx

( -rc-p-- osi)

DAEC 3.4-23 r---

I

RCS P/T Limits 3.4.9 150 200 25 ssel Metal Temperature (F) 9,(page 1 of 1)

Valid to Thirty-two 2003)

o. 800 0

i-0

Reporting Requi rements 5.6 5.6 Reporting Requirements 5.6.5 5.0-21 Amendment ;2T C-Fsc j

CORE OPERATING LIMITS REPORT (COLR)

(continued)

c.

The core operating limits shall be determined such that all applicable limits (e.g.. fuel thermal mechanical limits.

core thermal hydraulic limits. Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM. transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR. including any midcycle revisions or supplements.

shall be provided upon issuance for each reload cycle to the NRC.

PAM Report When a report is required by Condition B or F of LCO 3.3.3.1.

"Post Accident Monitoring (PAM)

Instrumentation," a report shall be submitted within the following 14 days.

The report shall outline the preplanned alternate method(s) of monitoring, describe the degree to which the alternate method(s) are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, the cause of the inoperability. and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.6 DAEC

Technical Specification Mark-up Inserts For TSCR-059 (PTLR)

Insert to Reporting Requirements 5.6.7 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.9, "RCS Pressure and Temperature (P/T) Limits."

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

NEDC-32983-P-A, "General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation";

GE-NE-A22-00100-08-01, "Pressure-Temperature Curves for Alliant Energy Duane Arnold Energy Center";

NEDC-32399-P, "Basis for GE RTNDT Estimation Method"; and, NEDO-32205-A, "10CFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 through BWR/6 Vessels."

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

Definitions 1.1 1.1 Definitions (continued)

MINIMUM CRITICAL POWER RATIO (MCPR)

(continued)

MODE OPERABLE -

OPERABILITY PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

RATED THERMAL POWER (RTP)

REACTOR PROTECTION SYSTEM (RPS)

RESPONSE

TIME film boiling occur intermittently with neither type being completely stable.

A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period.

These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 1912 MWt.

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

(continued)

TSCR-059 DAEC 1.1-5

RCS P/T Limits 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 RCS Pressure and Temperature (P/T) Limits LCO

3.4.9 APPLICABILITY

RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within the limits specified in the PTLR.

At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. --------- NOTE ---------

A.1 Restore parameter(s) 30 minutes Required Action A.2 to within limits.

shall be completed if this Condition is AND entered.

A.2 Determine RCS is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months acceptable for Requirements of the continued operation.

LCO not met in MODE 1, 2, or 3.

B.

Required Action and B.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND not met.

B.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

TSCR-059 DAEC 3.4-20

RCS P/T Limits 3.4.9 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. --------- NOTE--------- C.1 Initiate action to Immediately Required Action C.2 restore parameter(s) shall be completed if to within limits.

this Condition is entered.

AND C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 LCO not met in other operation.

or 3.

than MODES 1, 2, and 3.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 NOTE--------------

Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify RCS pressure, RCS temperature, and 30 minutes RCS heatup and cooldown rates are within the limits specified in the PTLR.

(continued)

TSCR-059 DAEC 3.4-21

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.9.2 Verify RCS pressure and RCS temperature are Once within within the criticality limits specified in 15 minutes the PTLR.

prior to control rod withdrawal for the purpose of achieving criticality SR 3.4.9.3

NOTE--------------

Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.

Verify the difference between the bottom Once within head coolant temperature and the Reactor 15 minutes Pressure Vessel (RPV) coolant temperature prior to each is within the limit specified in the PTLR.

startup of a reci rcul ati on pump SR 3.4.9.4 NOTE---------------

Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.

Verify the difference between the reactor Once within coolant temperature in the recirculation 15 minutes loop to be started and the RPV coolant prior to each temperature is within the limit specified in startup of a the PTLR.

reci rcul ation pump (continued)

TSCR-059 3.4-22 DAEC

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.4.9.5 NOTE---------------

Only required to be performed when tensioning the reactor vessel head bolting studs.

Verify temperatures at the reactor vessel head flange and the shell adjacent to the head flange are within the limit specified in the PTLR.

SR 3.4.9.6 NOTE--------------

Not required to be performed until 30 minutes after RCS temperature

80°F in MODE 4.

Verify temperatures at the reactor vessel head flange and the shell adjacent to the head flange are within the limit specified in the PTLR.

SR 3.4.9.7 NOTE---------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS temperature

100OF in MODE 4.

Verify temperatures at the reactor vessel head flange and the shell adjacent to the head flange are within the limit specified in the PTLR.

FREQUENCY

1*

30 minutes 30 minutes 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months TSCR-059 I

DAEC 3.4-23

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

(continued)

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 PAM Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM)

Instrumentation," a report shall be submitted within the following 14 days.

The report shall outline the preplanned alternate method(s) of monitoring, describe the degree to which the alternate method(s) are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Reactor Coolant System (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a.

RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.9, "RCS Pressure and Temperature (P/T) Limits."

b.

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

NEDC-32983-P-A, "General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation";

(continued)

TSCR-059 DAEC 5.0-21

Reporting Requirements 5.6 5.6 Reporting Requirements Reactor Coolant System (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

(continued)

GE-NE-A22-O0100-08-01, "Pressure-Temperature Curves for Alliant Energy Duane Arnold Energy Center";

NEDC-32399-P, "Basis for GE RTNDT Estimation Method"; and, NEDO-32205-A, "10CFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 through BWR/6 Vessels."

c.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

TSCR-059 5.6.7 DAEC 5.0-22

RCS P/T Limits B 3.4.9 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.9 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes.

These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips.

This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

ie*

,,L*,

contains P/T limit curves for heatup, cooldown, and inservice leakage and hydrostatic testing, and data for the maximum rate of change of reactor coolant temperature.

The heatup curve provides limits for both heatup and criticality.

Each P/T limit curve defines an acceptable region for normal operation.

The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the Reactor Coolant Pressure Boundary (RCPB).

The vessel is the component most subject to brittle failure.

Therefore, the LCO limits apply mainly to the vessel.

10 CFR 50, Appendix G (Ref. 1). requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials.

Reference 1 requires an adequate margin to brittle failure during normal operation.

anticipated operational occurrences, and system hydrostatic tests.

It mandates the use of the ASME Code,Section III, Appendix G (Ref. 2).

The actual shift in the RTT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and Appendix H of 10 CFR 50 (Ref. 4). _The operating P/T limit curves will be adjusted, (continued)

-Amendment 223 B 3.4-49 DAEC

RCS P/T Limits B 3.

4.9 BACKGROUND

(continued)

APPLICABLE SAFETY ANALYSES DAEC as necessary, based on the evaluation findings and the recommendations of Reference 5.

The P/T limit curves are composite.curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive.

At any specific pressure. temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit.

Across the span of the.P/T limit curves, different locations are more restrictive, and, thus. the curves are composites of the most restrictive regions.

The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed.

The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.

The criticality limits include the Reference 1 requirement that they be at least 40°F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic testing.

The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB. possibly leading to a nonisolable leak or loss of coolant accident.

In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components.

ASME Code,Section XI, Appendix E (Ref. 6). provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.

The P/T limits are not derived from Design Basis Accident (DBA) analyses.

They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed.

Reference 7 curves and limits specified in this section.

Since the P/T limits are not derived from any DBA, there are no acceptance low*- =*

o*

ao+44A e (continued)

B 3.4-50 1-s C'

ýL-0 5

)

BASES

RCS P/T Limits B 3.4.9 BASES APPLICABLE limits related to the P/T limits.

Rather, the P/T limits SAFETY ANALYSES are acceptance limits themselves since they preclude (continued) operation in an unanalyzed condition.

RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c) (2) (ii).

LCO The elements of tJýs LCO are:

0.4 't¶9A4

.rc*lt-I

a.

RCS pressureW.ar temperature are within the limits.-e#

]

/

l j.t h e,.-,

i p al,.

e,

..,.' y,,-, s 4f P i a n h'

~z~r~

?TLRcooldown. and :g 200 F/hr during prcnuea--

testifag-(c.g.

l

'5 INUte:

The IP/

limlits and corresponding heatup/cooldown rates of either Curve A or B may be applied while achieving or recovering from

""er conditions.

Curve A applies during pressure testing and when the limits of Curve B cannot be maintained:

b.

The temperature difference between the reactor vessel bottom head coolant and the Reactor Pressure Vessel (RPV) coolant is *-14 during recirculation pump G :"

e,, *e.*startup;

c.

The temperature difference between the reactor coolant in the respective recirculation loop and in the

_-rL reactor vessel is <

0 during recirculation pump L

~sta rtup:

d.

RCS pressure and temperature are within the criticality limits specified in,*.

rior to achieving criticality; and le.

TLR

e.

The temperatures at the reactor vessel head flange and the shell adjacent to the head flange are when tensioning the reactor vessel head bolting studs.

These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.

(continued)

TSCR-DaI-T DAEC B 3.4-51

RCS P/T LimiLts B 3.4.9 BASES SURVEILLANCE SR 3.4.9.1 (continued)

l')

REQUIREMENTS of 4ure 3.4.9 -

aremet when operation is to the right of the applicable limit curve.

Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be initiated and discontinued when the criteria given in the relevant plant procedure for starting and ending the activity are satisfied. During heatups and cooldowns, the temperatures at the reactor vessel shell adjacent to the shell flange, the reactor vessel bottom drain, recirculation loops A and B. and the reactor vessel bottom head shall be monitored.

During inservice hydrostatic or leak testing, the reactor vessel metal temperatures at the outside surface of the bottom head in the vicinity of the control rod drive housing and reactor vessel shell adjacent to the shell flange shall be monitored.

This SR has been modified with a Note that requires this Surveillance to be performed only during system heatup and cooldown operations and inservice leakage and hydrostatic testing.

SR 3.4.9.2 A separate limit is used when the reactor is approaching criticality.

Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor "L

\\

critical.

The limits of are met when p-p) operation is to the right of the applicable limit curve.

Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.

SR 3.4.9.3 and SR 3.4.9.4 Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances.

In addition, compliance with these limits (continued)

DAEC B 3.4-55 TSCR-.-

RCS P/T Limits B 3.4.9 BASES (continued)

REFERENCES

1.

2.

3.

4.

5.

6.

10 CFR 50, Appendix G, December 1995 ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.

ASTM E 185-82, July 1982.

10 CFR 50, Appendix H.

Regulatory Guide 1.99, Revision 2, May 1988.

ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.

7. B. ozafar (NRC) to G. anMiddles rth (NM S

ndment o. 238 to F cility Ope ating Licere No.

PR-49, ated April 3, 2001.

8.

UFSAR, Section 15.4.5.

"GE -RE-A2z-oo010-o I Rv i

-0 "UCes4

w 0 e4e o2.

TSCR-Pa*

0 DAEC B 3.4-58

RCS P/T Limits B 3.4.9 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.9 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes.

These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips.

This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

The PTLR contains P/T limit curves for heatup, cooldown, and inservice leakage and hydrostatic testing, and data for the maximum rate of change of reactor coolant temperature.

The heatup curve provides limits for both heatup and criticality.

Each P/T limit curve defines an acceptable region for normal operation.

The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the Reactor Coolant Pressure Boundary (RCPB).

The vessel is the component most subject to brittle failure.

Therefore, the LCO limits apply mainly to the vessel.

10 CFR 50, Appendix G (Ref.

1), requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials.

Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests.

It mandates the use of the ASME Code,Section III, Appendix G (Ref. 2).

The actual shift in the RT.

of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref.

3) and Appendix H of 10 CFR 50 (Ref. 4).

The operating P/T limit curves will be adjusted, (continued)

TSCR-059 DAEC B 3.4-49

RCS P/T Limits B 3.4.9 BASES BACKGROUND (continued)

APPLICABLE SAFETY ANALYSES as necessary, based on the evaluation findings and the recormmendations of Reference 5.

The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive.

At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit.

Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions.

The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed.

The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.

The criticality limits include the Reference I requirement that they be at least 40°F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic testing.

The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident.

In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components.

ASME Code,Section XI, Appendix E (Ref. 6), provides a recomTmended methodology for evaluating an operating event that causes an excursion outside the limits.

The P/T limits are not derived from Design Basis Accident (DBA) analyses.

They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed.

Reference 7 contains the methodology and plant-specific determination of the curves and limits specified in this section.

Since the P/T limits are not derived from any DBA, there are no acceptance (continued)

B 3.4-50 DAEC TSCR-059

RCS P/T Limits B 3.4.9 BASES APPLICABLE limits related to the P/T limits.

Rather, the P/T limits SAFETY ANALYSES are acceptance limits themselves since they preclude (continued) operation in an unanalyzed condition.

RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The elements of this LCO are:

a.

RCS pressure, temperature, and heatup or cooldown rate are within the limits specified in the PTLR. Note:

The P/T limits and corresponding heatup/cooldown rates of either Curve A or B specified in the PTLR may be applied while achieving or recovering from test conditions.

Curve A applies during pressure testing and when the limits of Curve B cannot be maintained;

b.

The temperature difference between the reactor vessel bottom head coolant and the Reactor Pressure Vessel (RPV) coolant is within the limit of the PTLR during recirculation pump startup;

c.

The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel is within the limit of the PTLR during recirculation pump startup;

d.

RCS pressure and temperature are within the criticality limits specified in the PTLR prior to achieving criticality; and

e.

The temperatures at the reactor vessel head flange and the shell adjacent to the head flange are within the limit of the PTLR when tensioning the reactor vessel head bolting studs.

These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.

(continued)

DAEC B 3.4-51 TSCR-059

RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.1 (continued)

REQUIREMENTS of the PTLR are met when operation is to the right of the applicable limit curve.

Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be initiated and discontinued when the criteria given in the relevant plant procedure for starting and ending the activity are satisfied. During heatups and cooldowns, the temperatures at the reactor vessel shell adjacent to the shell flange, the reactor vessel bottom drain, recirculation loops A and B, and the reactor vessel bottom head shall be monitored.

During inservice hydrostatic or leak testing, the reactor vessel metal temperatures at the outside surface of the bottom head in the vicinity of the control rod drive housing and reactor vessel shell adjacent to the shell flange shall be monitored.

This SR has been modified with a Note that requires this Surveillance to be performed only during system heatup and cooldown operations and inservice leakage and hydrostatic testing.

SR 3.4.9.2 A separate limit is used when the reactor is approaching criticality.

Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical.

The limits of the PTLR are met when operation is to the right of the applicable limit curve.

Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.

SR 3.4.9.3 and SR 3.4.9.4 Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances.

In addition, compliance with these limits (continued)

B 3.4-55 TSCR-059 DAEC

RCS P/T Limits B 3.4.9 BASES (continued)

REFERENCES B 3.4-58 DAEC

1.

10 CFR 50, Appendix G, December 1995

2.

ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.

3.

ASTM E 185-82, July 1982.

4.

10 CFR 50, Appendix H.

5.

Regulatory Guide 1.99, Revision 2, May 1988.

6.

ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.

7.

GE-NE-A22-00100-08-01, Rev. 1, "Pressure-Temperature Curves for Alliant Energy Duane Arnold Energy Center,"

September 2002.

8.

UFSAR, Section 15.4.5.

TSCR-059 to NG-03-0123 Page 1 of 4 SAFETY ASSESSMENT

1.0 INTRODUCTION

By letter dated February 28, 2003, Nuclear Management Company, LLC, (NMC), proposed changes to the Technical Specifications (TS) for the Duane Arnold Energy Center (DAEC). The requested changes are the relocation of the pressure temperature (P/T) limits numerical values and curves to the Pressure and Temperature Limits Report (PTLR), a licensee-controlled document, and the referencing of that report in the affected TS Limiting Conditions For Operation and Surveillance Requirements. Conforming changes to the TS BASES will be made pursuant to the TS BASES Control Program (TS 5.5.10). The proposed changes also include the addition of the PTLR to the Definitions section of the TS and the addition of a new section to the reporting requirements in the Administrative Controls section of the TS delineating the necessary requirements. Guidance on the proposed changes was developed by the Nuclear Regulatory Commission (NRC) on the basis of a proposal by the Owners Groups during the development of the Improved Standard Technical Specifications (ISTS). This guidance was provided to all power reactor licensees and applicants by Generic Letter 96-03, dated January 31, 1996. This guidance was subsequently modified by a generic change to the ISTS by Technical Specification Task Force (TSTF) change number TSTF 419, Rev. 0. The NRC has approved TSTF-419, Rev. 0 for use by licensees (

Reference:

W. Beckner (USNRC) to A. Pietrangelo (NEI), March 21, 2002).

2.0 BACKGROUND

Section 182a of the Atomic Energy Act (the Act) requires applicants for nuclear power plant operating licenses to include TS as part of the license. The Commission's regulatory requirements related to the content of TS are set forth in 10 CFR 50.36. That regulation requires that the TS include items in five specific categories: (1) safety limits, limiting safety system settings and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls, and states also that the Commission may include such additional TS as it finds to be appropriate. However, the regulation does not specify the particular requirements to be included in a plant's TS.

The Commission has provided guidance for the contents of TS in its "Final Rule on Technical Specification Improvements," 60 FR 36959 (July 19, 1995), in which the Commission indicated that compliance with the Final Rule satisfies Section 182a of the Act. In particular, the Commission indicated that certain items could be relocated from the TS to licensee-controlled documents, consistent with the standard enunciated in Portland General Electric Co. (Trojan Nuclear Plant),

ASLAB-531, 9 NRC 263, 273 (1979). In that case, the Atomic Safety and Licensing Appeal Board indicated that "technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety."

Consistent with this approach, the Final Rule identified four criteria to be used in determining whether a particular matter is required to be included in the TS, as follows: (1) installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary; (2) a process variable, design feature, or to NG-03-0123 Page 2 of 4 operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; (3) a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; (4) a structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety. As a result, existing TS requirements that fall within or satisfy any of the criteria in the Final Rule must be retained in the TS, while those TS requirements that do not fall within or satisfy these criteria may be relocated to other licensee-controlled documents.

One such TS requirement that was identified for relocation was the Pressure-Temperature (P/T)

Limits for the Reactor Coolant System (RCS). In Generic Letter 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," dated January 31, 1996, the Staff articulated the requirements for relocation of the P-T Limits to a licensee-controlled document. These requirements were incorporated into the Improved Standard Technical Specifications (ISTS) for Boiling Water Reactors (BWRs), NUREG - 1433, Revision 1.

Subsequently, these TSs were modified by a generic change submitted by the Technical Specification Task Force (TSTF) in traveler number TSTF-419, Revision 0. The approval of TSTF 419, Rev. 0 was communicated by the Staff in its letter, W. Beckner (USNRC) to A. Pietrangelo (NEI), March 21, 2002.

3.0 EVALUATION All components of the RCS are designed to withstand the effects of cyclic loads resulting from system pressure and temperature changes. These loads are introduced by normal heatup and cooldown operations, power transients, and reactor trips. In accordance with Appendix G to 10 CFR Part 50, TS limit the pressure and temperature changes during RCS heatup and cooldown within the design assumptions and the stress limits for cyclic operation. These limits are defined by P/T limit curves for heatup, cooldown, and in-service leak and hydrostatic testing. Each curve defines an acceptable region for normal operation. The curves are used for operational guidance during heatup and cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region. In addition, the rate of change of RCS temperature is controlled during normal heatup and cooldown operations, i.e., degrees Fahrenheit per hour.

NMC proposes changes to the DAEC TS that are in accordance with the guidance in Generic Letter 96-03, as modified by TSTF-419, Rev. 0, which are germane to BWRs. Specifically, BWRs do not have requirements for Low Temperature Overpressure Protection (LTOP), as discussed in GL 96 03; hence, they are not included in these changes. The specific changes being made are as follows:

(1) The definitions section of the TS was modified to include a definition of the "PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)" to which the updated figures, values, and parameters for P/T limits will be relocated on a unit-specific basis in accordance with a methodology approved by the NRC that maintains the acceptance limits and the limits of the safety analysis.

to NG-03-0123 Page 3 of 4 (2) The following specification was revised to replace the specific P/T limits with a reference to the PTLR that provides the actual limits:

Limiting Condition for Operation (LCO) 3.4.9, "RCS Pressure and Temperature (P/T) Limits."

(3) Specification 5.6.7, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," was added to the reporting requirements of the administrative controls section of the TS. This specification requires that the PTLR be submitted, upon issuance, to the NRC. The report provides the explanations, figures, values, and parameters of the P/T limits for the applicable effective fluence period. Furthermore, this specification requires that the figures, values, and parameters be established using the methodologies approved by the NRC for this purpose in the following approved topical reports:

NEDC-32983-P-A, "General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation;"

GE-NE-A22-00100-08-01, "Pressure-Temperature Curves for Alliant Energy Duane Arnold Energy Center;"

NEDC-32399-P, "Basis for GE RTNDT Estimation Method;" and, NEDO-32205-A, "10CFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 through BWR/6 Vessels."

Relocation of the P/T curves and heatup and cooldown rate limits does not eliminate the requirement to operate in accordance with the limits specified in Appendix G to 10 CFR Part 50.

The requirement to operate within the limits in the PTLR is specified in and controlled by the TS.

Only the figures, values, and parameters associated with the P/T limits are to be relocated to the PTLR. In order for the curves and heatup and cooldown rates to be relocated to a PTLR, a methodology for their development must be reviewed and approved in advance by the NRC. The methodology to be approved by the NRC is to be developed in accordance with GL 96-03. Generic Letter 96-03 delineates the requirements for both the methodology and the PTLR including, but not limited to, the requirements of Appendix G to 10 CFR Part 50. The PTLR review process requires that the NRC approve changes to the methodology. Further, when future changes are made to the figures, values, and parameters contained in the PTLR, the PTLR is to be updated and submitted to the NRC upon issuance.

On this basis, NMC concludes that the proposed changes provided an acceptable means of establishing and maintaining the detailed values of the P/T limit curves and heatup and cooldown rate limits. Further, because plant operation continues to be limited in accordance with the requirements of Appendix G to 10 CFR Part 50, and the updated P/T limits referenced in the TS are established using a methodology approved by the NRC, these changes will not impact plant safety.

NMC also concludes that the relocated requirements discussed above relating to the P/T limits are not required to be in the TS under 10 CFR 50.36 or Section 182a of the Atomic Energy Act, and are not required to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety. Additionally, they do not fall within any of the four criteria set to NG-03-0123 Page 4 of 4 forth in the Commission's Final Rule, discussed above. Accordingly, NMC concludes that the proposed changes are acceptable and that these requirements may be relocated from the TS to the PTLR, a licensee-controlled document.

to NG-03-0123 Page 1 of 1 ENVIRONMENTAL CONSIDERATION 10 CFR Section 51.22(c)(9) identifies certain licensing and regulatory actions which are eligible for categorical exclusion from the requirement to perform an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; and (3) result in a significant increase in individual or cumulative occupational radiation exposure. Nuclear Management Company, LLC has reviewed this request and determined that the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR Section 51.22(c)(9). Pursuant to 10 CFR Section 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the amendment. The basis for this determination follows:

Basis The change meets the eligibility criteria for categorical exclusion set forth in 10 CFR Section 51.22(c)(9) for the following reasons:

1. As demonstrated in Attachment 1 to this letter, the proposed amendment does not involve a significant hazards consideration.

2. The proposed change involves relocation of the existing numerical values and updated curves of the reactor vessel pressure and temperature (P/T) limits, using NRC-approved methods, to a licensee-controlled document, the Pressure and Temperature Limits Report, in accordance with NRC Generic Letter 96-03, as modified by an NRC-approved generic change (TSTF 490, Revision 0). The requirement to comply with these limits will be retained in the Technical Specifications.

The proposed change does not involve modifications to the radioactive waste processing systems or to radioactive waste effluent monitors. Accordingly, the changes do not require the radioactive waste processing systems to perform any different function than they are designed to perform nor do they change the operation or testing of any such system.

Therefore, this change will not result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

3. The proposed change will not appreciably change the way the plant or its systems are operated. Only the numerical values and curves of existing P/T limits are being relocated to a licensee-controlled document, but the requirement to comply with these limits will be retained in the Technical Specifications. Thus, there will be no significant increase in either individual or cumulative occupational radiation exposure.

Therefore, this change will not result in a significant increase in individual or cumulative occupational radiation exposure.

to NG-03-0123 Duane Arnold Energy Center Pressure And Temperature Limits Report Revision 0

IMMU Committed to Nuclear Excellence Duane Arnold Energy Center (DAEC)

Pressure And Temperature Limits Report (PTLR) up to 32 Effective Full-Power Years (EFPY)

Revision 0 Prepared by:

Reviewed by Approved b3ý-,-

Directe Engineering Concurred by:

a4nager Iegulatory Affairs Date: 2-1316 3 Date:

l6J '°3 Date:

Date:_____

DAEC PTLR Rev. 0 February, 2003 Table of Contents Section Page 1.0 Purpose 1

2.0 Applicability 1

3.0 Methodology 1

4.0 Operating Limits 2

5.0 Discussion 3

6.0 References 5

Figure 1 Composite P-T Curves Effective to 32 EFPY 6

Table I Data Table for Composite P-T Curve - 25 EFPY 7

Table 2 Data Table for Composite P-T Curve - 32 EFPY 13 App. A Reactor Vessel Material Surveillance Program 20 i

DAEC PTLR Rev. 0 February, 2003 1.0 Purpose The purpose of the Duane Arnold Energy Center (DAEC) Pressure and Temperature Limits Report (PTLR) is to present operating limits relating to:

1) Reactor Coolant System (RCS) Pressure versus Temperature limits during Heatup, Cooldown and Hydrostatic/Class 1 Leak Testing;

2) RCS Heatup and Cooldown rates;

3) Reactor Pressure Vessel (RPV) to RCS coolant AT requirements during Recirculation Pump startups;

4) RPV bottom head coolant temperature to RPV coolant temperature AT requirements during Recirculation Pump startups;

5) RPV head flange boltup temperature limits.

This report has been prepared in accordance with the requirements of Technical Specification (TS) 5.6.7, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)."

2.0 Applicability This report is applicable to the DAEC RPV up to 32 Effective Full-Power Years (EFPY).

The following TS is affected by the information contained in this report:

TS 3.4.9 RCS Pressure and Temperature (P/T) Limits; 3.0 Methodology The limits in this report were derived from the NRC-approved methods listed in TS 5.6.7, using the specific revisions listed below:

1) The neutron fluence was calculated per General Electric (GE) topical report NEDC-32983-P-A, "General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation," Rev. 0, August 2000, approved in Reference 6.1.

I

DAEC PTLR Rev. 0 February, 2003 3.0 Methodology (cont.)

2) The pressure and temperature limits were calculated per GE topical report, GE-NE-A22-00100-08-01, Rev. 1, "Pressure-Temperature Curves for Alliant Energy Duane Arnold Energy Center, September, 2002," which incorporated ASME Code Case N-640 and the increased fluence as a result of Extended Power Uprate (EPU). The methodology used was previously approved in Reference 6.2.

a. This methodology utilizes the following additional NRC-approved methods:

i. GE topical report, NEDC-32399-P, "Basis for GE RTNDT Estimation Method," Rev. 0, September 1994, approved in Reference 6.3.

ii. GE topical report, NEDO-32205-A, "10CFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 through BWR/6 Vessels," Rev. 1, February 1994, approved in Reference 6.4.

Changes to the curves, limits, or parameters within this PTLR, based upon new irradiation fluence data of the RPV, or other plant design assumptions in the Updated Final Safety Analysis Report (UFSAR), can be made pursuant to 10 CFR 50.59, provided the above methodologies are utilized. The revised PTLR shall be submitted to the NRC upon issuance.

Changes to the curves, limits, or parameters within this PTLR, based upon new surveillance capsule data of the RPV, cannot be made without prior NRC approval. Such analysis and revisions shall be submitted to the NRC for review prior to incorporation into the PTLR.

4.0 Operating Limits The pressure-temperature (P-T) curves included in this report represent steam dome pressure versus minimum vessel metal temperature and incorporate the appropriate non-beltline limits and irradiation embrittlement effects in the beltline region.

Complete P-T curves were developed for 25 and 32 EFPY. The P-T curves are provided in Figure 1 and a tabulation of the curves is included in Table 1 (25 EFPY) and Table 2 (32 EFPY).

The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level physics tests, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

2

DAEC PTLR Rev. 0 February, 2003 4.0 Operating Limits (cont.)

Normal Operating Heatup and Cooldown rate limit (Figure 1: Curve B - non nuclear heating and Curve C - nuclear heating): < 100 oF/hour.

Heatup and Cooldown rate limit during Hydrostatic and Class 1 Leak Testing (Figure 1: Curve A): < 20 oF/hour.

RPV bottom head coolant temperature to RPV coolant temperature AT limit during Recirculation Pump startup: _< 145 'F.

Recirculation loop coolant temperature to RPV coolant temperature AT limit during Recirculation Pump startup:

50 'F.

RPV flange and adjacent shell temperature limit: > 74 'F.

5.0 Discussion The adjusted reference temperature (ART) of the limiting beltline material is used to adjust the beltline P-T curves to account for irradiation effects. Regulatory Guide 1.99, Revision 2 (RG 1.99) provides the methods for determining the ART.

The RG 1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section.

The vessel beltline copper and nickel values (except for the N2 and N1 6 nozzles) were obtained from the evaluation of DAEC Surveillance Capsules (Reference 6.5). For the N2 and N16 nozzles, a bounding value of 0.18% was assumed for copper, and the nickel values for N16 and N2 of 0.85% and 0.84%, respectively, were obtained from a Certified Material Test Report. The copper (Cu) and nickel (Ni) values were used with Tables 1 and 2 of RG 1.99, to determine a chemistry factor (CF) per Paragraph 1.1 of RG 1.99 for welds and plates, respectively.

The P-T curves for the non-beltline region were conservatively developed for a boiling-water reactor product line 6 (BWR/6) with nominal inside diameter of 251 inches. The analysis is considered appropriate for DAEC, since the plant specific geometric values are bounded by the generic analysis for large BWR/6. The generic value was adapted to the conditions at DAEC using plant specific RTNDT values for the reactor pressure vessel.

3

DAEC PTLR Rev. 0 February, 2003 5.0 Discussion (cont.)

The peak RPV ID fluence used in the P-T curve evaluation for 32 EFPY is 4.17el 8 n/cm2, which was calculated using methods that comply with the guidelines of RG 1.190. This fluence applies to the lower-intermediate plates and longitudinal welds. The fluence is adjusted for the lower plates and longitudinal welds and the girth weld based upon an attenuation factor of 1.18; hence, the peak ID surface fluence for these components is 3.55e1 8 n/cm2. Similarly, the fluence is adjusted for the N2 nozzle based upon an attenuation factor of 3.7; hence the peak ID surface fluence used for this component is 7.64e17 n/cm2. The same method is applied to the N16 nozzle, which has an attenuation factor of 5.46, resulting in a peak ID surface fluence of 1.13e18 n/cm 2.

The P-T curves for the heatup and cooldown operating condition at a given EFPY apply for both the 1/4T and 3/4T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 3/4T location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cooldown and is in the outer wall during heatup. However, as a conservative simplification, the thermal gradient stress at the 1/4T location is assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stress at the 1/4T location. This approach is conservative because irradiation effects cause the allowable toughness, KIr, at 1/4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, well above the heatup/cooldown curve limits.

For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves specify a coolant heatup and cooldown temperature rate of< I 00°Fihr for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram and the nozzle thermal cycle diagrams. For the hydrostatic pressure and leak test curve (Curve A), a coolant heatup and cooldown temperature rate of*< 2 0 °F/hr must be maintained. The P/T limits and corresponding heatup/cooldown rates of either Curve A or B may be applied while achieving or recovering from test conditions. Curve A applies during pressure testing and when the limits of Curve B cannot be maintained.

For Duane Arnold, the N2 Recirculation Inlet nozzle is the limiting material for the beltline region for 32 EFPY. The beltline pressure test P-T curves provided in this report are calculated in the same manner as the Feedwater Nozzle pressure test P-T curves, using the N2-specific geometry. The initial RTNDT for the N2 Recirculation Inlet nozzle materials is 40TF. The generic pressure test P-T curve is applied to the Duane Arnold N2 Nozzle curve by shifting the P vs. (T - RTNDT) 4

DAEC PTLR Rev. 0 February, 2003 5.0 Discussion (cont.)

values to reflect the ART value of 119.2'F. Similarly, the generic pressure test P T curve is applied to the Duane Arnold N2 Nozzle curve by shifting the P vs. (T RTNDT) values to reflect the 25 EFPY ART value of 113.6°F. Using the fluence discussed above, the P-T curves are beltline (N2 Recirculation Inlet nozzle) limited above 240 and 230 psig for Curve A for 25 and 32 EFPY, respectively, and above 30 psig for Curve B for both 25 and 32 EFPY.

6.0 References 6.1 Letter, S.A. Richards, (USNRC) to J.F. Klapproth, (GE-NE), "Safety Evaluation for NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation (TAC No. MA9891)", MFN 0 1-050, September 14, 2001.

6.2 Letter, B. Mozafari (USNRC) to G. VanMiddlesworth (NMC), "DUANE ARNOLD ENERGY CENTER - ISSUANCE OF AMENDMENT RE: REVISED PRESSURE-TEMPERATURE CURVES (TAC NO. MB0394)," April 30, 2001.

6.3 Letter from B. Sheron (USNRC) to R.A. Pinelli (BWROG), "Safety Assessment of Report NEDC-32399-P, Basis for GE RTNDT Estimation Method, September 1994, "December 16, 1994.

6.4 J. T. Wiggins (USNRC) to L.A. England (BWROG), "ACCEPTANCE FOR REFERENCING OF TOPICAL REPORT NEDO-32205, Rev. 1, '1 OCFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 Through BWR/6 Vessels'," December 8, 1993.

6.5 L. J. Tilly, "Duane Arnold RPV Surveillance Materials Testing and Analysis" GE-NE, San Jose, CA, July 1997, (GE-NE-B 1100716-01, Revision 0).

5

DAEC PTLR Rev. 0 February, 2003 Figure 1 Curve A (EFPY) 25 32 B

C 50.0 100.0 150.0 200.0 250.0 Minimum Reactor Vessel Metal Temperature (F)

Pressure Versus Minimum Temperature Valid to Thirty-two Full Power Years, per Appendix G of 1 OCFR50 6

1400 1200 "R1000 CO CI 0 800 U,

0 600 0

C.,

CU 400 E

10..

0~

A - System Hydrotest Limit with F el in Vessel (20' /hr heatuplcool lown rate)

B - Non-Nu lear Heating Limit, Valid to 32 EFPY (100' F/hr

-heatup/cool Jown-rate)

C - Nuclear (Core Critical" Limit, Valid to 32 EFPY (100' F/hr

-...eatup/cooldown rate Bottom He d Curve A Bottom Head CurwB 0.0 300.0

DAEC PTLR Rev. 0 February, 2003 Table 1 BOTTOM U1JPSIRftP V&

DOTTOM1UPE b

rn bVa.

AN HEFAD S9ELTLIJNE AT HEAD QELT LINE AT BELTLIJFE AT 25EFPY 25EFPY 25 FFPY PRESSURE CURVE A CURVEA

CURVES, CURVEB CRV 10 6S.0 74,0 68,0 74,0 74.0 20 69,0 74,0 68,0 74,0 74,0 30 62S0 740i 68,0 74,0 100,3 40 6S,0 74,0 68,0 82,7 122,7 50 68.0 74,0 68,0 97.2 137,2 60 6s,0 74,0 68,0 1OsO 148,0 70 62O 74,0 68,0 116,6 156,6 s0 6S.0 74,0 68,0 123.6 163,6 90(

68,0 74.0 68,0 129,5 169'5 1 (R 6S,0 74,0(

6S,0 134,7 174,7 110 68,0 74,0 68,0 139,3 179,3 120 68,0 74,0 68,0 14-3,6 193,6 130 68,0 74,0 68,0 147,5 187,5 140 68,0 74,0 68,0 151,1 191,1 150 68,0 74,0 68,0 154,2 194,2 160 68,0 74,0 68,0 157,2 197,2 170 68,0 74,0 68,0 160A1 200,1 180 68,0 74,0 68,0 162,7 202,7 190 68,0 74.0 68,0 165,2 205,2 200 68,0 74,0 68,0 167,5 207,5 210 68.0 74.0 68.0 169,7 209j7 7

DAEC PTLR Rev. 0 February, 2003 Table I (cont.)

BOTTOM UPPE=R 1W') & BeOTTOM UPPE=R 1W') &

AND HEAD StLTLJNE AT HEAD OELWLINE AT BEETLINE AT 25 26tPry 25 rFPY 5

W C

220 230 240 250) 260 270 280 290 3so 310 312,5 312,5 320 330 340 350 360 370 380 390 400 410 420 430 68,0 68,0 68.0 68,0 168,0 68,0 68,0 6S,O 6s.0 68,0 68,0 63,0 68,0 68.0 68,0 68,0 68,0 68,0 68,0 68,0 68,0 68,0 68,0 74,0 74,0 74,0 78-5 93,6 993 92.6 96.5 100,2 103,6 104,4 104,4 106,8 109,8 112,6 115,2 117.8 120.2 122,5 124,7 126.8 128,8 130,8 132,7 68,0 68,0 69,0 68,0 69,0 68.0 68,0 68,0 68,0 68,0 68,0 6S,0 68,0 68,0 68,0 68,0 68.0 68,0 68,0 68,0 68,0 68,0 68,0 68A0 171.9 173,9 175,9 177,7 179,4 181.I 182,8 194A4 185.9 187,3 187.7 197,7 198,9 190,2 191,5 192,8 194,1 195,3 196,5 197,6 198,'

199,9 201,0 202,O 211.9 213,9 215.8 2 137,7 21%71 219A 221,1 222,8 224.4 225,9 227,3 227,7 227,7 228,8 230,2 231,5 232,8 234,1 235,3 236,5 237,6 238,8 239,9 241,0 242.0 8

DAEC PTLR Rev. 0 February, 2003 Table 1 (cont.)

no~t~td)t t

-RPV& 13bflM UP&~ahE v

Stb HIEAD BamTIrE AT HEAD SELTLINE AT BEtTLtt4E ATI PRESUE U VEA GURVEA CURVE 8 CURVE B, CURVE 440 450 460 470 480 490 500 510 520 530 540 550 560 570 580 590 60LI 610 620 630 640 650 660 670 68,0 68.0 68,0 68.0 68,0 68,0 68.0 68,0 68.0 68,0 618,0 69,0 168.0 68,0 68.0 6G,O 68,0 68,0 68,0 68,0 68.0 68.0 68,0 134.5 13 6,2 137,9 139.5 141.1 142,6 144,1 145,5 147,0 149.3 149,6 150,9 152,2 153A4 154,6 155.8 156,9 158,1 159,1 160,2 161.3 162,3 163,3 164,3 68,0 68,0 68,0 68.0 68,0 68,0 68,0 68.0 68.2 70,2 72,1 73,9 75,7 77A4 79,0

¶0.6 82.2 83,7 85,1 86.5 87,9 89.2 90's 91,8 203 1 204.1 205,0 206,0 207,0 207.9 208.8 209,7 210.6 211.4 212"3 213,1 213,9 214,7 215,5 216.2 217,0 217.7 218.5 219,2 219,9 220,6 221.3 222,0 243,1 24431 245,0 246,0 247,0 247,9 248,A 249,7 250.6 251,4 252,3 253,1 253,9 254,7 255,5 256.2 257.0 257,7 258,5 259,2 259,9 260,6 261,3 262,0 9

DAEC PTLR Rev. 0 February, 2003 Table 1 (cont.)

96ttU(ý rP'Okfl&V torSOTOr Uflftim KP HEAD 1t:LtINE AT HEAD Sr=LTLINE AT 8EtTL1NE AT

'25 tfpy 26FlPY 2 ~P 11PRESSURE CU R'VE A CUJRVE A CURVE B CURVE B CURVE C 680 690 700 710 720 730 740 750 760 770 780 79f0 800 910 820 830 840 850 860 970 890 900 910 68,0 68,0 69.2 70,7 72,1 73,5 74,9 76,1 77.4 78,6 79,'

81,0 82,2 93,3 84,4 85,5 96,5 87.6 88,6 89,6 90,5 91.5 92,4 93A 165,2 166,2 167,1 169,0 169,9 169,'

170,7 171,5 172,4 173,2 174,0 174,0 175,6 176.4 177.1 177.9 178,6 179,3 190.1 180,8 181,5 182,2 182,8 183,5 93,1 94,3 95,4 96.6 97,7 98,8 99,9 101.0 102.0 103.0 104.0 105.0 106.9 107.8 108,7 109,6 110.4 111.3 112,1 112.0 113.8 114.6 115A 222,6 223.3 224,0 22-4,6 225,2 225,8 226,5 227,1 227,7 228,3 228,9 229.4 230,0 230,6 231,1 231.7 232,2 232,8 233,3 233,8 234,3 234,9 235,4 235,9 262,6 263,3 264,0 264.6 265,2 265,'

266,5 267,1 267,7 26823 268,9 269A 270,0 270,6 271,1 271,7 272,2 272,8 273,3 273.8 274.3 274,9 27534 275,9 10

DAEC PTLR Rev. 0 February, 2003 Table 1 (cont.)

HEAD tELILINS AT

HEAD, S=LTUINE At OEL.TUINrATj

ýPRESURE CURVE A CUVA CURVES8 CURVESB CURVEt C 920 94,3 184.2 116.1 236,4 276,4 930 95,1 194.:

1169 236,9 276,9 940 96,0 185,5 11-77 237.3 277,3 950 96,9 186,1 lISA 237,8 277,9 960 97,7 186.7 119.1 238,3 278,3 970 98,6 187,3 119.9 238,8 278,8 980 99,4 188.0 120.6 239,2 279,2 990 100.2 188,6 121.3 239,7 279,7 1000 101.0 189.2 122A0 240,2 280.2 1010 101.7 189.7 122.6 240,6 280.6 1020 102-5 190,3 123.3 241,1 281,1 1030 10323 190.9 124.0 241,5 291,5 1040 104,0 191,5 124.6 241,9 281,9 1050 104.7 192.0 125.3 242.4 282,4 1060 105,4 192.6 125,9 242,8 282,9 1070 1062 193.1 126.5 243.2 283,2 1080 106.9 193.7 127.2 243,7 23-,7 1090 107.6 194,2 127.8 244,1 284,1 1100 108,2 1940 128A 244,5 294,5 1110 108.9 195,3 129.0 244,9 284,9 1120 109.6 195.8 129.6 245,3 285,3 1130 110.2 196,3 130.2 245,7 285,7 1140 110,9 196,8 130.7 246,1 286.1 1150 111.5 197,3 1313 246,5 286,5 11

DAEC PTLR Rev. 0 February, 2003 Table 1 (cont.)

HEAD 3ELTUNE AT HEAD OELTNE AT, 8ELTUIN AT

~256r=PP 25 r.F.

25$Efl'

,PRESSURE CURVE A CURVE A CURVE 13 CURV VB CURVE C 1160 1170 1180 1190 1200 1210 1220 123 0 1240 1250 1260 1270 1280 1290 1300 1310 1320 1330 1340 1350 1360 1370 1380 1390 1400 112.1 112,8 113A 114.0 114.6 115.2 115.8 11623 116.9 117.5 118.0 118.6 119,1 119.7 120.2 120.7 1213 121.8 1223 122.8 1233 123,8 1243 124.8 125.3 197.9 198,3 198,s 199,3 199.8 200,3 200,g 201,2 201,7 202,2 202,6 203.,1 203,5 2041) 204,4 204,&

205,3 205.7 206.1 206,5 207.0 207,4 207,8 208,2 208.6 131. 9 132.4 133.0 133.5 134.1 134.6 135.2 135.7 136.2 136.7 137.2 137.7 138.2 138.7 139,2 139.7 140.2 140.6 141.1 141,6 142.0 142,5 142.9 143A 143.8 246,9 247,3 247,7 248,0 248,3 248.5 248,g 249,1 249,3 249,6 249,9 250,1 250,3 250,6 250,8 231,1 251,3 251,6 251,8 252,0 252,3 25235 252,8 253,0 253.2 286,9 287,3 287,7 288,0 288,3 2988,5 298,8 289.1 289,3 289,6 289,8 290,1 290M3 290,6 290,8 29101 291,3 291,6 29108 29210 292.3 292,5 292,8 292,0 293.2 12

DAEC PTLR Rev. 0 February, 2003 Table 2 T ~

of~fUotp PR P&~ doTTom f1JORP'V 46WEI-CT Ei 32 EF~PY 32 LFY 32 5trPY PRESSURE CURVE A, OURVFA tCUVE B U~=13 GIV (PSIG)

(",F)

('FC)

(T

~)

(T),(T 0

68,0 74,0 68.0 74.0 74.0 10 68,0 74,0 68.0 74,0 74,0 20 68,0 74,0 68.0 74,0 74,0 30 6s,0 74,0 68.0 74,0 103.9 40 68,0 74,0 68.0 88,3 129.3 50 68,0 74,0 68,0 102,9 142.8 60 68,0 74,0 6S.0 113.6 15*.6 70 68,0 74,0 68.0 122.2 162.2 80 6S,O 74,0 68.0 129.2 169-2 90 68,0 74,0 68.0 135.1 175.1 1 H 68,0 74,0 68.0 140.3 1803 110 6,,0 74,0 68.0 144.9 184.9 120 68,0 74,0 68,0 149.2 189.2 130 68.,

74,0 68.0 153,1 193.1 140 68,0 74,0 68.0 156.7 196.7 150 68,0 74,0 68,0 159'8 199.8 160 68,0 74,0 68,0 162.8 202.8 170 68,0 74,0 68.0 165.7 205.7 180 68,0 74,0 68.0 168,3 2083 190 68,0 74,0 68.0 170.9 210.8 200 68.0 74.0 68.0 173.1 213.1 13