Text

License Amendment Request 02-04, Revision to TS Associated with Pressure/Temperature Curves & Low Temperature Overpressure Protection Limits
	ML022940024
Person / Time
Site:	Seabrook
Issue date:	10/11/2002
From:	Vargas J; North Atlantic Energy Service Corp
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NYN-02093
	Download: ML022940024 (123)
	v • d • e

North Atlantic Energy Service Corporation SNorth P.O. Box 300 Seabrook, NI1 03874 Atlantic 474-9521 The Northeast Utihties System October 11, 2002 Docket No. 50-443 NYN-02093 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 Seabrook Station License Amendment Request 02-04 "Revision To Technical Specifications Associated With Pressure/Temperature Curves and Low Temperature Overpressure Protection Limits" North Atlantic Energy Service Corporation (North Atlantic) has enclosed herein License Amendment Request (LAR) 02-04. License Amendment Request 02-04 is submitted pursuant to the requirements of 10 CFR 50.90 and 10 CFR 50.4.

Provided in Enclosure 1, LAR 02-04 proposes changes to the Seabrook Station Technical Specifications 3.4.9.1, "Reactor Coolant System - Pressure/Temperature Limits" and 3.4.9.3, "Reactor Coolant System - Overpressure Protection Systems". Specifically, the proposed changes will replace Technical Specification Figure 3.4-2, "Reactor Coolant System Heatup Limitations", Figure 3.4-3, "Reactor Coolant System Cooldown Limitations" and Figure 3.4-4, "RCS Cold Overpressure Protection Setpoints" to allow operation to 20 Effective Full Power Years.

A similar submittal was approved for Arkansas Nuclear One, Unit 2, on April 15, 2002 (TAC NOS. MB3301 and MB3302) and in part for Millstone Nuclear Power Station, Unit 3, on August 27, 2001 (TAC NO. MB 1785).

The Index and the associated Bases for these Technical Specifications will be modified as a result of the proposed changes.

In addition, pursuant to 10 CFR 50.12, North Atlantic requests an exemption from the specific requirements of 10 CFR 50.60(a) and 10 CFR 50 Appendix G, based on American Society of Mechanical Engineers Code Case N-641, to support the revised reactor vessel analyses.

Enclosure 1, LAR Section I.B contains the justification for the exemption request. A similar exemption was approved for Arkansas Nuclear One, Unit 2, on April 2, 2002 (TAC NO.

MB3301).

U. S. Nuclear Regulatory Commission NYN-02093/Page 2 The Station Operation Review Committee and the Nuclear Safety Audit Review Committee have reviewed LAR 02-04.

As discussed in Enclosure 1, LAR Section IV, the proposed change does not involve a significant hazard consideration pursuant to 10 CFR 50.92. A copy of this letter and the enclosed LAR has been forwarded to the New Hampshire State Liaison Officer pursuant to 10 CFR 50.91(b). North Atlantic requests NRC Staff review of LAR 02-04, and issuance of a license amendment by September 30, 2003 (see Enclosure 1,Section V enclosed).

North Atlantic has determined that LAR 02-04 meets the criterion of 10 CFR 51.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement (see ,Section VI enclosed).

Should you have any questions regarding this letter, please contact Mr. James M. Peschel, Manager - Regulatory Programs, at (603) 773-7194.

Very truly yours, NORTH ATLANTIC ENERGY SERVICE CORP.

Dirý,ecr - Engineering cc: H. J. Miller, NRC Region I Administrator R. D. Starkey, NRC Project Manager, Project Directorate 1-2 G. T. Dentel, NRC Senior Resident Inspector Mr. Donald Bliss, Acting Director New Hampshire Office of Emergency Management State Office Park South 107 Pleasant Street Concord, NH 03301

ENCLOSURE 1 TO NYN-02093 North Atlantic SEABROOK STATION UNIT 1 Facility Operating License NPF-86 Docket No. 50-443 License Amendment Request 02-04, "Revision To Technical Specifications A~slsociated With Pressure/Temperature Curves and ]Low Temperature Overpressure Protection Limits" This License Amendment Request is submitted by North Atlantic Energy Service Corporation pursuant to 10CFR50.90. The following information is enclosed in support of this License Amendment Request:

0 Section I - Introduction and Safety Assessment for Proposed Changes 0 Section II - Markup of Proposed Changes 0 Section III - Retype of Proposed Changes 0 Section IV - Determination of Significant Hazards for Proposed Changes 0 Section V - Proposed Schedule for License Amendment Issuance And Effectiveness 0 Section VI - Environmental Impact Assessment I, Joe M. Vargas, Director - Engineering of North Atlantic Energy Service Corporation hereby affirm that the information and statements contained within this License Amendment Request are based on facts and circumstances which are true and accurate to the best of my knowledge and beliefff, ...

J9 M. Vargas Sworn and Subscribed /1irector - Engineering before me this 111h day of October, 2002 1~~~~~,

V% '.6? )JMe6A&

N%-ar Public

SECTION I INTRODUCTION AND SAFETY ASSESSMENT FOR PROPOSED CHANGES

,Section I I. INTRODUCTION AND SAFETY ASSESSMENT OF PROPOSED CHANGES A. Introduction and Description of Change Reviewer's Note Throughout this amendment request, three terms are used which are defined as follows to ease the reviewer's interpretation of the material presented.

LTOP - Low Temperature Overpressure Protection - This term is used when referring to the cold overpressure requirements of ASME Section XI Appendix G and 10CFR50 Appendix G.

COPPS - Cold Overpressure Protection System - This is used when referring to the cold overpressure setpoints in Seabrook Station Unit 1 Technical Specification 3.4.9.3.

The title of the Technical Specification is "Overpressure Protection Systems."

COMS - Cold Overpressure Mitigating System - This is used when referring to the cold overpressure study performed by Framatome, which establishes the setpoints for COPPS.

LTOP is also the system name for the Seabrook system which implements the COPPS setpoints that come from the COMS study.

Use of ASME Code Case N-641 The proposed Technical Specification changes to modify the pressure-temperature (P/T) limits and overpressure protection system setpoints rely in part on the use of the American Society of Mechanical Engineers (ASME) Code Case N-641. The revised P/T limits, as specified in ASME Code Case N-641, use a higher allowable stress intensity factor, Kic instead of KIR, which results in higher allowable pressures. KIR is a reference stress intensity factor, based on the lower bound values of Kic and KtA. P/T curves and overpressure protection system setpoints based on the Kic curve will enhance overall plant safety by opening the P/T operating window, with the greatest safety benefit in the region of low temperature operations. In addition, enhanced safety during critical plant operational periods, heatup and cooldown evolutions is expected.

The primary safety benefits in opening the low temperature operating window are a reduction in the challenges to pressurizer power-operated relief valves (PORVs), and additional margin to maintain reactor coolant pump (RCP) net positive suction head (NPSH) requirements. In addition, the pressure undershoot due to the relief capacity of one PORV and the time delay for the valve to close after opening for pressure relief due to a Cold Over Pressure Protection System (COPPS) event can result in damage to the RCP seals due to inadequate seal differential pressure. Damage to the RCP seals can require an unplanned shutdown to replace the seals. By raising the COPPS setpoints at low reactor coolant system (RCS) temperatures, the likelihood of challenging the pressurizer PORVs will be reduced, and operation at higher pressures to provide additional margin for RCP seal protection will be allowed.

The P/T limits determined using ASME Code Case N-641 are less restrictive than the requirements of 10 CFR 50 Appendix G, Section IV.A.2.b, which requires the use of methods Page 1 of 10

,Section I equivalent to those provided by Appendix G to ASME Section XI. Since ASME Section XI Code Case N-641 was employed in the development of the reactor vessel beltline P/T limits, an exemption to 10 CFR 50.60(a), based on ASME Code Case N-641, is required to support the proposed Technical Specification changes (Section HI).

Pressure/Temperature Curve Revision License Amendment Request (LAR) 02-04 proposes changes to the Seabrook Station Technical Specifications 3.4.9.1, "Reactor Coolant System - Pressure/Temperature Limits" and 3.4.9.3, "Reactor Coolant System - Overpressure Protection Systems". Specifically, the proposed changes will replace Technical Specification Figure 3.4-2, "Reactor Coolant System Heatup Limitations", Figure 3.4-3, "Reactor Coolant System Cooldown Limitations" and Figure 3.4-4, "RCS Cold Overpressure Protection Setpoints". The proposed change of the reactor vessel P/T limit curves is required because the existing curves referenced above are valid through the attainment of 11.1 Effective Full Power Years (EFPY). Based on current projections, 11.1 EFPY will be achieved early in the next operating cycle, Cycle No. 10, at the end of the fourth quarter of 2003. The revised reactor vessel P/T limit curves will remain valid for 20 EFPY as demonstrated in the analysis documented in Reference (1) and included as Enclosure 2. The associated Bases will be modified as necessary as result of the proposed changes. In addition to the proposed Technical Specification, North Atlantic requests an exemption to 10 CFR50.60(b),

based on ASME Code Case N-641, Reference (2), to support the development of the revised reactor vessel P/T limit curves.

Overpressure Protection System Revision License Amendment Request (LAR) 02-04 proposes a change to Seabrook Station Technical Specification 3.4.9.3 "Reactor Coolant System - Overpressure Protection Systems".

Specifically, the proposed change replaces Technical Specification Figure 3.4-4, "RCS Cold Overpressure Protection Setpoints" and revises the COPPS arming temperature. The maximum PORV COPPS setpoints specified in Figure 3.4-4 and COPPS arming temperature are revised to reflect the higher allowable low temperature overpressure protection (LTOP) pressure limit afforded by the use of ASME Code Case N-641. The revised pressurizer PORV COPPS setpoints have been derived for operation of Seabrook Station's reactor vessel to a cumulative exposure of 20 EFPY, consistent with the proposed revisions to the P/T limits in Figures 3.4-2 and 3.4-3. Credit is taken for the fact that experience shows that LTOP events are most likely to occur at isothermal conditions. The LTOP allowable pressure limit for the COPPS is therefore taken to be the steady-state (isothermal) P/T pressure limit curve. The allowable pressure is limited below 150TF, to comply with the closure head/vessel flange region limitation on system pressure imposed by 10 CFR Part 50, Appendix G. The revised COPPS setpoints provide overpressure protection for the Seabrook reactor vessel and closure head/flange region in accordance with ASME Code Case N-641 and 10 CFR Part 50 Appendix G.

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,Section I B. Evaluation of Proposed Changes Use of ASME Code Case N-641 The following information provides the basis for the exemption request to 10CFR50.60 for use of American Society of Mechanical Engineers (ASME)Section XI Code Case N-641, "Alternate Pressure/Temperature Relationship and Low Temperature Overpressure Protection System Requirements,Section XI, Division I", in lieu of methods specified in 10CFR50, Appendix G.

The requested exemption will allow use of ASME Code Case N-641 to (a) determine stress intensity factors for postulated circumferential defects in circumferential welds, and for postulated axial defects in plates, forgings and axial welds, and (b) use the KIc fracture toughness curve shown on ASME XI, Appendix A, Figure A-2200-1, in lieu of the KLA fracture toughness curve of ASME XI, Appendix G, Figure G-2210-1, as the lower bound fracture toughness.

10CFR50.12 states that the Commission may grant an exemption from requirements contained in 10CFR50 provided that:

1. The requested exemption is authorized by law:

No law exists which precludes the activities covered by this exemption request.

10CFR50.60(b) allows the use of alternatives to 10CFR50, Appendices G and H when an exemption is granted by the Commission under 10CFR50.12.

2. The requested exemption does not present an undue risk to the public health and safety:

10CFR50, Appendix G, requires, in part, that Article G-2120 of ASME Section XI, Appendix G, be used to determine the maximum postulated defects in reactor pressure vessels when determining pressure/temperature (P/T) limits for the vessel. These limits are determined for normal operation and pressure test conditions.

Article G-2120 specifies, in part, that the postulated defect be in the surface of the vessel material and normal to the direction of maximum stress. ASME Section XI, Appendix G, also provides a methodology to determine the stress intensity factors for a maximum postulated defect normal to the maximum stress. The purpose of this article is, in part, to prevent non-ductile fractures by providing procedures to identify the most limiting postulated fractures to be considered in the development of P/T limits.

Due to progress made in non-destructive examination (NDE) techniques over the last thirty years, it is unlikely that undetected defects will be present in the beltline region of reactor vessels. It is further unlikely to have axial cracks originating from a circumferential weld perpendicular to the weld seam orientation in reactor vessels. Both experience and engineering studies indicate that the primary degradation mechanism affecting the beltline region of the reactor vessel is neutron embrittlement. No other service reduced degradation mechanism exists at a pressurized water reactor to cause a prior existing defect located in the beltline region of the reactor vessel to grow while in Page 3 of 10

,Section I service. Based on these considerations, and the fact that the P/T limit for reactor operation is the limiting pressure for any of the materials in the vessel, it is not necessary to include additional conservatism in the assumed flaw orientation for circumferential welds. ASME Section XI, Code Case N-641, and a previousSection XI, Appendix Code change correct this inconsistency in assumed flaw orientation for circumferential welds in vessels when calculating operating P/T limits.

Code Case N-641 provides benefits in terms of calculating P/T limits by revising the Section XI, Appendix G reference flaw orientation for circumferential welds in reactor vessels. The reference flaw is a postulated flaw that accounts for the possibility of an existing defect that may have gone undetected during the fabrication process. When considering a reference flaw with respect to a weld, the reference flaw would represent any prior existing defect that may have been introduced during fabrication. Thus, the intended application of a reference flaw is to account for prior existing defects that could physically exist within the geometry of the weldment. The currently endorsed ASME Section XI, Appendix G approach mandates consideration of an axial reference flaw in circumferential welds for purposes of calculating P/T limits. Postulating the Appendix G reference flaw in a circumferential weld is physically unrealistic and overly conservative, because the length of the flaw is 1.5 times the vessel thickness, which is much longer than the width of the reactor vessel girth weld. The possibility that an axial flaw may extend from a circumferential weld into a plate/forging or axial weld is already adequately covered by the requirement that axial defects be postulated in plates/forging and axial welds.

ASME Code Case N-641 reflects fabrication and NDE experience by allowing consideration of maximum postulated defects oriented circumferentially within the welds.

Code Case N-641 also provides appropriate procedures to determine limiting circumferential weld defects and associated stress intensity factors for use in developing P/T limits per ASME Section XI, Appendix G procedures. The procedures allowed by Code Case N-641 are conservative and provide a margin of safety in the development of P/T operating and pressure test limits that will prevent nonductile fractures.

The revised P/T limits and overpressure protection system limits being proposed for Seabrook Station have been developed using the KIc fracture toughness curve shown on ASME Section XI, Appendix A, Figure A-2200-1, in lieu of the Kup fracture toughness curve of ASME XI, Appendix G, Figure G-2210-1, as the lower bound for fracture toughness. Use of the Kic curve in determining the lower bound fracture toughness in the development of P/T operating limits curve is more technically correct than the KIA curve.

The KIc curve models the slow heatup and cooldown process of a reactor vessel.

Use of this approach is justified by the initial conservatism of the KIA curve when the curve was codified in 1974. This initial conservatism was necessary due to limited knowledge of reactor pressure vessel materials over time and usage. Since 1974, additional knowledge has been gained about the effect of usage on reactor pressure vessel materials. The additional knowledge demonstrates the lower bound on fracture toughness provided by the KI curve provides a margin of safety that is adequate to protect the public health and safety from potential reactor pressure vessel failure.

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,Section I The COMS analyses for Seabrook Station were also performed using the method provided in Code Case N-641. Use of Code Case N-641 methodology in the determination of the LTOP conditions is more technically correct than the generic value included in earlier versions of ASME Section XI and eliminates inconsistencies in the margin of safety between reactor vessels of various geometries. Code Case N-641 provides bounding reactor vessel low temperature integrity protection during LTOP design basis transients. The LTOP lift setpoint utilizes 100% of the pressure determined to satisfy Appendix G, Paragraph G-2215 of ASME Section XI, Division 1, as a design limit. The approach is justified by consideration of the overpressurization design basis events and the resulting margin to reactor vessel failure.

P/T limit curves based on Code Case N-641 will enhance overall plant safety by opening the pressure/temperature operating window with the greatest safety benefit in the region of low temperature operations. The primary safety benefit in opening the low temperature operating window is a reduction in the challenges to LTOP valves. The proposed P/T limits include restrictions on allowable operating conditions and equipment operability requirements to ensure that operating conditions are consistent with the assumptions of the accident analysis. Specifically, the reactor coolant system (RCS) pressure and temperature must be maintained within the heatup and cooldown rate dependent pressure/temperature limits that are specified in TS 3.4.9. Therefore, this exemption does not present an undue risk to the public health and safety.

3. The requested exemption will not endanger the common defense and security:

The common defense and security are not endangered by this exemption request.

4. Special circumstances are present which necessitate the request for an exemption to the regulations of 10CFR50.60:

Pursuant to 10CFR50.12(a)(2), the NRC will consider granting an exemption to the regulations if special circumstances are present. This exemption meets the special circumstances of Paragraphs:

10CFR50.12(a)(2)(ii) demonstrates that the underlying purpose of the regulation will continue to be achieved; (a)(2)(iii) would result in undue hardship or other cost that are significant if the regulation is enforced and; (a)(2)(v) will provide only temporary relief from the applicable regulation and the licensee has made good faith efforts to comply with the regulations.

10CFR50.12(a)(2)(ii), Underlying Purpose of the Regulation Will Continue to be Achieved:

The underlying purpose of 10CFR50, Appendix G and ASME Section XI, Appendix G, is to satisfy the requirement that the reactor coolant pressure boundary be operated in a manner having sufficient margin to ensure that, when stressed, the vessel boundary behaves in a non-brittle manner and the probability of a rapidly propagating failure is minimized. Accordingly, that the P/T operating and test curves provide adequate margin Page 5 of 10

,Section I in consideration of the uncertainties in determining the effects of irradiation on material properties.

Application of Code Case N-641 to determine P/T operating and hydrostatic test limit curves per ASME Section XI, Appendix G, provides appropriate procedures to determine limiting maximum postulated defects and considering those defects in the P/T limits.

This application of the code case maintains the margin of safety originally considered for reactor pressure vessel materials. Therefore, use of Code Case N-641, as described above, satisfies the underlying purpose of the ASME Code and the NRC regulations to ensure an acceptable level of safety.

10CFR50.12(a)(2)(iii), Result In Undue Hardship or Other Cost:

The P/T operating window is defined by the P/T operating and test curves developed in accordance with the ASME Section XI, Appendix G procedure. Continued operation with these more restrictive P/T curves without the relief provided by ASME Code Case N-641 would unnecessarily restrict the pressure/temperature and LTOP operating window for Seabrook Station. Use of Case N-641 will minimize the potential for reactor coolant pump (RCP) impeller cavitation wear while operating in the LTOP region and reduce the potential for inadvertent actuation of the LTOP relief valves. Use of ASME Code Case N-641 in the development of the proposed P/T curves and Overpressure Protection System setpoint and enable temperature alleviates any unwarranted burden.

Implementation of the proposed P/T curves and LTOP parameters as allowed by ASME Code Case N-641 does not reduce the margin of safety originally considered by either the NRC or ASME.

Compliance with the specified requirements of 10CFR50.60 would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

ASME Code Case N-641 allows:

" Postulation of a circumferential defect in circumferential welds is appropriate in lieu of requiring the defect to be oriented across the weld from one plate or forging to the adjoining plate or forging. This circumstance was not considered at the time ASME Section XI, Appendix G was developed and imposes restrictions on P/T operating limits beyond those originally contemplated.

" A reduction in the fracture toughness lower bound is appropriate in lieu of the ASME Section XI, Appendix G, in the determination of reactor coolant pressure/temperature limits. This proposed alternative is acceptable because the Code Case maintains the relative margin of safety commensurate with that which existed at the time ASME Section XI, Appendix G was approved in 1974. Therefore, application of Code Case N-641 for Seabrook Station will ensure an acceptable margin of safety. The approach is justified by consideration of the overpressurization design basis events and the resulting margin to reactor vessel failure.

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,Section I 10CFR50.12(a)(2)(v), Licensee Has Made Good Faith Efforts to Comply with the Regulations.

The exemption provides only temporary relief from the applicable regulation and North Atlantic has made a good faith effort to comply with the regulation. We request that the exemption be granted until such time that the NRC generically approves ASME Code Case N-641 for use by the nuclear industry.

Restrictions on allowable operating conditions and equipment operability requirements have been established to ensure that operating conditions are consistent with the assumptions of the accident analysis. Specifically, RCS pressure and temperature must be maintained within the heatup and cooldown rate dependent pressure/temperature limits that are specified in the proposed amendment. Therefore, this exemption does not present an undue risk to the public health and safety.

To summarize, these proposed alternatives are acceptable because the Code Case maintains the relative margin of safety commensurate with that which existed at the time ASME Section XI, Appendix G, was approved in 1974. Therefore, application of Code Case N-641 Seabrook Station will ensure an acceptable margin of safety.

Pressure/Temperature Curve Revision The recalculated reactor vessel P/T limit curves for normal operation are valid through 20 EFPY.

The technical justification and methodologies employed in their development are documented in WCAP-15745, Reference (1) and included in Enclosure 2. The P/T limit curves were generated using the most recent reactor vessel surveillance capsule data which is documented in Duke Engineering and Services Report DES-NFQA-98-01, Reference (3). North Atlantic has previously submitted DES-NFQA-98-01 via North Atlantic Letter NYN98078, Reference (4).

Capsule Y, the second Seabrook Station Unit 1 Surveillance Capsule, was removed from the reactor vessel after completion of Operating Cycle No. 5 in May 1997. At that point in time, the reactor vessel had accrued 5.572 EFPY. Analyses of the neutron dosimetry determined that the capsule had received an average neutron fast fluence (E>I Mev) of 1.15 X 1019 n/cm 2 . At that rate, this is equivalent to the fluence that will be received at the reactor vessel inner diameter after approximately 15 EFPY of operation. The second reactor vessel material capsule specimens were destructively tested and evaluated using the methodologies prescribed in Regulatory Guide 1.99, Revision 2, Reference (5), for predicting beltline material radiation embrittlement. It was concluded that the plate and weld material upper shelf energies will be maintained above 50 ft-lb throughout reactor vessel life as required by 10 CFR50, Appendix G.

Also, based on the surveillance capsule data, the adjusted RTNDT values for the plate and weld material were within the two standard deviations of Regulatory Guide 1.99, Revision 2 predictions. As all the requisite criteria of Regulatory Guide 1.99, Revision 2 were satisfied, Reference (3) concluded that the surveillance data was credible and the beltline material was responding as empirically predicted.

The P/T limit curves derived in WCAP-15745 used the adjusted RTNDT value corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTNDT of the limiting material in the core region of the reactor vessel was determined by using the unirradiated reactor vessel fracture toughness properties, estimating the radiation induced ARTNDT and adding Page 7 of 10

,Section I margin. The unirradiated RTNDT is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35 mil lateral expansion minus 60'F. The P/T limit curves developed in WCAP-15745 used the NRC approved methodology documented in Reference (6) with the exception of the following:

"* The fluence values used were calculated values, not best estimate fluence values.

"* The Kic critical stress intensities were used in lieu of the KtA critical stress intensities. This methodology was taken from approved ASME Code Case N-641, Reference (2).

"* The 1996 Edition of Appendix G to Section XI of ASME Code, Reference (7), was used rather than the 1989 Edition.

Overpressure Protection System Revision Two design basis overpressure events are considered in the COPPS setpoint development: a heatup transient and a mass addition event. The COPPS design basis heatup transient starts from an initial condition where the RCS flow rate is at zero, the steam generators are at a temperature 500F hotter than the rest of the RCS, the pressurizer is water solid (a bubble has not yet been formed) and then a RCP is inadvertently started. The starting of the RCP initiates heat transfer from the secondary to the primary side which causes a pressurization of the primary side.

Simulations of the heatup event were used to determine the peak RCS pressure setpoint overshoot defined as the peak pressure in the vessel adjacent to the beltline minus the nominal PORV setpoint. Calculations were performed over a range of assumed initial water temperatures below the expected COPPS arming temperature.

The second design basis event is a mass addition event. The event considered is mass addition from a single charging pump via both the normal charging flow path, with the charging flow and head control valves fully open, and the charging pump safety injection flow path in parallel, due to an inadvertent opening of one of the two SI flow path block valves. Letdown flow was assumed to be isolated. Current plant Technical Specifications limit the number of operable charging pumps to a single Centrifugal Charging Pump (CCP) in Modes 4 and 5 except for a brief transition period when entering or exiting Mode 3 and/or when swapping pumps.

Therefore the analysis assumed flow from a single CCP. Maximum setpoint overshoot was determined as a function of initial wide-range pressure and PORV setpoint for a cold (100 0 F) reactor coolant condition. Assuming a cold RCS conservatively minimizes the compressibility of the RCS water volume, maximizing the rate of pressure increase and setpoint pressure overshoot.

The pressure overshoot values calculated for the design basis heat and mass addition events were used to determine the maximum allowable PORV COPPS setpoint required to prevent violation of the LTOP allowable pressure limit during the event. The maximum PORV COPPS setpoint is lower than the LTOP allowable pressure limit minus the overshoot predicted for the applicable temperature condition. The analysis included consideration of transient temperature effects on the temperature-dependent COPPS setpoint and vessel beltline water temperature. To assure the COPPS setpoint will not exceed the maximum value specified in Figure 3.4-4 temperature and pressure measurement and instrument uncertainties are applied in determining the nominal PORV COPPS setpoint function hardware settings and COPPS arming temperature. The setpoints of the two PORVs are also staggered to minimize the probability of opening more than Page 8 of 10

Enclosure 1,Section I one valve, since one valve provides adequate relief flow. The recalculated COPPS setpoints are valid through 20 EFPY and provide overpressure protection for the Seabrook reactor vessel and closure head/flange region in accordance with ASME Code Case N-641 and 10 CFR Part 50 Appendix G.

The ability of the Residual Heat Removal (RHR) system suction line safety valves to provide protection of the revised LTOP pressure limit independent of the COPPS as allowed by Technical Specification 3.4.9.3.a was also verified.

Cold Overpressure Miti2ation System (COMS) Arming Temperature The COMS system design has an arming bistable for each Power-Operated Relief Valve (PORV). When wide-range (WR) RCS temperature goes below the selected arming temperature, the bistable arms COMS by causing the associated PORV block valve to open (if it is closed) and enabling the PORV COMS setpoint.

ASME Code Case N-641 indicates the LTOP System Effective Temperature is the "temperature at or above which the safety relief valves provide adequate protection against nonductile failure ". Per the code case, LTOP systems shall be effective below the higher of an inlet coolant temperature of 200'F or a coolant temperature corresponding to a reactor vessel 1/4T metal temperature of RTNDT + 40'F for inside axial surface flaws and RTNDT - 85 0F for inside circumferential surface flaws for all vessel beltline materials. The adjusted 1/4T RTNDT for the limiting beltline material at 20 EFPY is 109'F. Therefore, the code case requires the COMS system to be effective below an inlet coolant temperature of 200'F.

The selected LTOP allowable pressure limit at 2000 F, however, is lower than the safety relief valve setpoint (2560 psig = 2485 psig +/-3%, per TS 3.4.2.1). The pressure limit corresponding to the potentially 50'F lower transient temperature in the reactor vessel (RV) downcomer during a mass addition event is also less than the maximum safety valve setpoint. Therefore, the COMS arming temperature is selected to be greater than or equal to the temperature (273.60 F) at which the allowable COMS relief valve pressure setpoint is > the pressurizer safety relief valve maximum opening setpoint (2560 psig) plus an allowance for the difference between indicated WR pressure and pressure at the safety valve (approximately 67 psi)1 , i.e. > 2627 psig).

This temperature is adjusted to a higher value to include allowances for applicable temperature measurement and arming bistable uncertainties (13.3 0 F). Therefore the revised arming temperature is 273.6 + 13.3 = 286.90 F which is rounded up to 290'F. The nominal PORV COMS setpoints at the arming temperature would be: PCV-456A = 2670.5 psig, and 2610.5 psig for PCV-456B.

1 From the PRESS model input, the elevation difference between the top of the pressurizer to the bottom

3. + 144 in2/ft2 =

of the reactor vessel = 84.55 ft, corresponding to an elevation head = 84.55 ft x 62 lb/ft 36.4 psi. The flow loss with 4 RCPs running is approximately 46 psi at 700 F. Thus the total pressure difference between the bottom of the vessel where WR pressure is tapped and the top of the pressurizer could be as high as 46 + 36.4 = 82.4 psi. The indicated WR pressure is approximately 15.3 psi lower than the pressure at the tap, so the final pressure difference = 82.4 - 15.3 = 67.1 psi.

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,Section I C. Safety Assessment Conclusion of Proposed Changes North Atlantic concludes that based upon the above discussion (Section B, Evaluation of Proposed Changes), as well as the "Determination of Significant Hazards for Proposed Changes," presented in Section IV, that the proposed changes do not adversely affect or endanger the health or safety of the general public or involve a significant safety hazard.

D. References

1. Westinghouse WCAP-15745, Revision 0, "Seabrook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, dated December 2001.

2. ASME Boiler and Pressure Vessel Code Case N-641,Section XI, Division 1, "Alternative Pressure-Temperature Relationship and Overpressure Protection System Requirements", dated January 17, 2000.

3. Duke Engineering and Services Report DES-NFQA-98-01, Revision 0, "Analysis of Seabrook Station Unit 1 Reactor Vessel Surveillance Capsules U and Y", dated May 1998.

4. North Atlantic Letter NYN-98078, T. C. Feigenbaum to U. S. NRC, "Seabrook Station Reactor Vessel Surveillance Capsule Report", dated June 5, 1998

5. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", U. S. Nuclear Regulatory Commission, dated May 1988.

6. Westinghouse WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", dated January 1996.

7. ASME Boiler and Pressure Vessel Code,Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure" dated December 1995, through 1996 Addendum.

8. Framatome Letter NFSB 02-0061, "COMS Setpoints for 20 EFPY," August 30, 2002.

9. "Seabrook Station Cold Overpressure Mitigating System (COMS) Setpoint Development Methodology," August 2002.

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,Section II SECTION II MARKUP OF PROPOSED CHANGES Refer to the attached markup of the proposed changes to the Technical Specifications. The attached markup reflects the currently issued revision of the Technical Specifications listed below. Pending Technical Specifications or Technical Specification changes issued subsequent to this submittal are not reflected in the enclosed markup.

The following Technical Specification changes are included in the attached markup:

Technical Specification Title Page Index Limiting Conditions for Operation vi&x&xi and Surveillance Requirements 3.4.9.1 Reactor Coolant System 3/4 4-31 & 32 Pressure/Temperature Limits Reactor Coolant System 3.4.9.3 3/4 4-34 & 36 Overpressure Protection Systems 3/4.4.9 Pressure/Temperature Limits B 3/4 4-7 through 9 B 3/4 4-11 through 15 Page 1 of 29

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE FIGURE 3.4-1 DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY> IJLCi / gram DOSE EQUIVALENT 1-131 ............................................................... 3/4 4-28 TABLE 4.4-3 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR O G RA M ..................................................................................... 3/4 4-29 3/4.4.9 PRESSURE/TEMPERATURE LIMITS G eneral ........................................................................................ 3/4 4-30 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 1 -4M Y ...................................................... 3/44-31 I

FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 44-4 EFPY ...................................................... 3/44-32 Pressurizer ........................................................................... 3/4 4-33 Overpressure Protection Systems .............................................. 3/4 4-34 3/4 4-36 FIGURE 3.4-4 RCS COLD OVERPRESSURE PROTECTION SETPOINTS ...........

3/4 4-37 3/4.4.10 STRUCTURAL INTEGRITY ..............................................................

3/4.4.11 REACTOR COOLANT SYSTEM VENTS ............................................... 3/4 4-38 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS 3/4 5-1 Hot Standby, Startup, and Power Operatibn ...........................................

3/4 5-3 Shutdow n ..........................................................................................

314 5-4 3/4.5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350"F ........

3/4 5-8 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 350OF ......................................

3145-10 ECCS SUBSYSTEMS - Tavg Equal To or Less Than 200°F ................

3/4 5-11 3/4.5.4 REFUELING WATER STORAGE TANK .................................................

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity .................................................................... 3/4 6-1 Containment Leakage ................................................................... 3/4 6-2 SEABROOK - UNIT 1 vi Amendment No. -?&

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.12.2 (THIS SPECIFICATION NUMBER IS NOT USED) ........................................ 3/4 12-3 3/4.12.3 (THIS SPECIFICATION NUMBER IS NOT USED) ........................................ 3/4 12-5 3.0/4.0 BASES 3/4.0 APPLICABILITY ...................................................................................................... B 3/4 0-1 314.1 REACTIVITY CONTROL SYSTEMS 314.1.1 BORATION CONTROL ..................... .................. B 3/4 1-1 3/4.1.2 BORATION SYSTEMS .................................................................................... B 3/4 1-2%

3/4.1.3 MOVABLE CONTROL ASSEMBLIES ............................................................. B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS ............................................................................ B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE ............................................................................ B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR ................................................. B 3/4 2-2 3/4.2.4 QUADRANT POWER TILT RATIO ................................................................. B 3/4 2-3 3/4.2.5 DNB PARAMETERS ....................................................................................... B 3/4 2-4 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION ............................ B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION ............................................................. B 3/4 3-3 3/4.3.4 (THIS SPECIFICATION NUMBER IS NOT USED) ........................................ B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION .................. B 3/4 4-1 3/4.4.2 SAFETY VALVES ............................................................................................ B 3/4 4-1 3/4.4.3 PRESSURIZER ............................................................................................... B 3/4 4-2 3/4.4.4 RELIEF VALVES ............................................................................................. B 3/4 4-2 3/4.4.5 STEAM GENERATORS .................................................................................. B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE ................................................... B 3/4 4-3 3/4.4.7 CHEM ISTRY .................................................................................................... B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY ....................................................................................... B 3/4 4-5 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ............................................................ B 3/4 4-7 THIS RIG NS OT USD FIGURE B 3/4.4-1 I ,., N - I ,v, A rUf,;T-,e;q eF 3/4.2 B

""FGR (This number not.gur.e used)................................ B 3/4 4-9 FIGURE B 3/4.4-2 (This figure number not used) ............................................................ B 314 4-10 SEABROOK - UNIT I X Amendment No. 5,66, 74-

INDEX BASES SECTION I PAGE B

B ~

3/4.4-1 ~ RF=AFeR, TO 3/4.4-1,,

TABLE 7ABLE~ ',-,*_--,,,,,,,-,

VEJGEI ,#1 ll NE # ........... ,....,......e...=....-...o..-.....................

B34 B3441

-1 3/4.4.10 STRUCTURAL INTEGRITY ................................................................................. B 3/4 4-16 3/4.4.11 REACTOR COOLANT SYSTEM VENTS ............................................................ B 3/4 4-16 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS ............................................................................................... B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS .......................................................................... B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK ............................................................. B 314 5-2 3/4.6 CONTAINMENT SYSTEMS 314.6.1 PRIMARY CONTAINMENT .................................................................................. B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS ............................................ B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES ................................................................ B 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTROL ........................................................................ B 3/4 6-3 3/4.6.5 CONTAINMENT ENCLOSURE BUILDING ......................................................... B 3/4 6-4 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ................................................................................................. B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION .................. B 3/4 7-3 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM ..................................... B 3/4 7-3 3/4.7.4 SERVICE WATER SYSTEM ................................................................................ B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK ........................................................................................ B 3/4 7-3 3/4.7.6 CONTROL ROOM SUBSYSTEMS ...................................................................... B 3/4 7-4 3/4.7.7 SNUBBERS .......................................................................................................... B 3/4 7-4 3/4.7.8 SEALED SOURCE CONTAMINATION ............................................................... B 3/4 7-5 3/4.7.9 (THIS SPECIFICATION NUMBER IS NOT USED) ............................................. B 3/4 7-5 3/4.7.10 (THIS SPECIFICATION NUMBER IS NOT USED) ............................................. B 3/4 7-5 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION ....................................................................... B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES ....................................... B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION ................................................................................ B 3/4 9-1 3/4.9.2 INSTRUMENTATION ........................................................................................... B 3/4 9-1 3/4.9.3 DECAY TIME ........................................................................................................ B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS .................................................... B 3/4 9-1 3/4.9.5 COMMUNICATIONS ............................................................................................ B 3/4 9-1 3/4.9.6 REFUELING MACHINE ....................................................................................... B 3/4 9-1 314.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING ................................... B 3/4 9-2 SEABROOK - UNIT 1 xi Amendment No. 56,-6 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE/TEMPERATURE LIMITS SFor Information Only GENERAL LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup of 100*F in any 1-hour period,

b. A maximum cooldown of 100°F in any 1-hour period, and

c. A maximum temperature change of less than or equal to 10'F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS Tavg and pressure to less than 200OF and 500 psig, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.9.1 The Reactor Coolant System temperature and pressure'shall be deter mined to be within the limits at least once per 30.minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

SEABROOK - UNIT 1 3/4 4-30

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,Section II Insert A, Page 3/4 4-31 MATERIAL PROPERTY BASIS Umiting material: LOWER SHELL PLATE R-1808-1 Limiting ART values at 20 EFPY.I/4T, 109'F 314T, 88°F Curves applicable for the first 20 EFPY and contain margins of 20°F and 100 psig for possible instrument errors 2800 ja _1a _ La - a - -.. L - JIa 1 a I.- a I - JI -

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"- 1600 a a 80Deg. F/Hr , 80 Deg F/Hr , a a a a I a aI a a a I

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Li i base on a

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0 0 100 200 300 400 500 RCS TEMPERATURE (Deg. F, 20 Deg. F PER DIVISION)

FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 20 EFPY Page 7 of 29

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0 100 200 300 400 500 RCS TEMPERATURE (Deg. F, 20 Deg. F PER DIVISION)

FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS - APPLICABLE UP TO 20 EFPY Page 9 of 29

REACTOR COOLANT SYSTEM PRESSURE/TEMPERATURE LIMITS OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 The following Overpressure Protection Systems shall be OPERABLE:

a. In MODE 4 when the temperature of any RCS cold leg is less than or equal to-9*9OF; and in MODE 5 and MODE 6 with all Safety Injection pumps at least one of the following groups of two overpressure protection devices shall be OPERABLE when the RCS is not depressurized with an RCS vent area of greater than or equal to 1.58 square inches:

1) Two residual heat removal (RHR) suction relief valves each with a setpoint of 450 psig +0, -3 %; or

2) Two power-operated relief valves (PORVs) with lift setpoints that vary with RCS temperature which do not exceed the limit established in Figure 3.4-4, or

3) One RHR suction relief valve and one PORV with setpoints as required above.

b. In MODE 5 and MODE 6 with all Safety Injection pumps except one inoperable:

1) The Reactor Coolant System (RCS) depressurized with an RCS vent area equal to or greater than 18 square inches, or

2) The RCS in a reduced inventory condition*.

APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less thian or equal to N F; MODE 5 and MODE 6 with the reactor vessel head on and the vessel head closur bolts not fully detensioned.

ACTION:

a) In MODE 4 with all Safety Injection pumps inoperable and with one-of the tW.0 required overpressure protection devices inoperable, either restore two overpressure protection devices to OPERABLE status within 7 days or Within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (a) depressurize the RCS and (b) vent the RCS through at least a 1.58-square-inch vent.

A reduced inventory condition exists whenever reactor vessel (RV) water level is lower than 36 inches below the RV flange.

SEABROOK - UNIT 1 3/4 4-34 Amendment No. 3, 5,4 6, REACTOR COOLANT SYSTEM PRESSURE/TEMPERATURE LIMITS OVERPRESSURE PROTECTION SYSTEMS Fonformation Only LIMITING CONDITION FOR OPERATION 3.4.9.3 ACTION: (Continued) b) In MODE 5 and MODE 6 with all Safety Injection pumps inoperable and with one of the two required overpressure protection devices inoperable, restore two overpressure protection devices to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (a) depressurize the RCS and (b) vent the RCS through at least a 1.58-square-inch vent.

c) In MODE 4, MODE 5 and MODE 6 with all Safety Injection pumps inoperable and with both of the two required overpressure protection devices inoperable, within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (a) depressurize the RCS and (b) vent the RCS through at least a 1.58-square-inch vent.

d) In the event the PORVs, or the RHR suction relief valves, or the RCS vent(s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.8.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs, or the RHR suction relief valves, or RCS vent(s) on the transient, and any corrective action necessary to prevent recurrence.

e) in MODE 5 and MODE 6 with all Safety Injection pumps except one inoperable and with the RCS vent area less than 18 square inches or RCS water level not in a reduced inventory condition, immediately restore all Safety Injection pumps to inoperable status.

SEABROOK - UNIT 1 3/4 4-34a Amendment No. ., 4.6, 74 1

REACTOR COOLANT SYSTEM PRESSURE/TEMPERATURE LIMITS SYSTEMS OVERPRESSURE PROTECTION SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE when the PORV(s) are being used for overpressure protection by:

a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, bbt excluding valve operation, at least once per 31 days thereafter when the PORV is required OPERABLE; and

b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and

c. Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valve(s) are being used for overpressure protection as follows:

a. For RHR suction relief valve RC-V89 by verifying at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that RHR suction isolation valves RC-V87 and RC-V88 are open.

b. For RHR suction relief valve RC-V24 by verifying at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that RHR suction isolation valves RC-V22 and RC-V23 are open.

c. Testing pursuant to Specification 4.0.5.

4.4.9.3.3 The RCS vent(s) shall be verified to be open at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s** when the vent(s) is being used for overpressure protection.

4.4.9.3.4 The reactor vessel water level shall be verified to be lower than 36 inches below the reactor vessel flange at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reduced inventory condition is being used for overpressure protection.

"**Exceptwhen the vent pathway is provided with a valve(s) or device(s) that is locked, sealed, or otherwise secured in the open position, then verify this valve(s) or device(s) open at least once per 31 days.

SEABROOK - UNIT I 3/4 4-35 Amendment No. 3, 5,1., 74

I I*

us r L

IL L fn z C

i e4 SEABROOK - UNIT 1 3/4 4-36

,Section II Insert A, Page 3/4 4-36 VALID FOR THE FIRST 20 EFPY, SETPOINT CONTAINS MARGIN OF 50°F FOR TRANSIENT EFFECTS 0

T<s 200.0 F, P = 561.0 PSIG; 200.00 F < T 230.5 0F, P = 12.1*(T-200.0) + 926.0 PSIG; 230.50 F < T < 255.0 0F, P = 23.15*(T-230.5) + 1295.05 PSIG; T > 255.0 0F, P = 34.5*(T-255.0) + 1862.225 PSIG 2500 2250 I F I F F F 2000 F.

F I

I S F I F S I 1750 1500 F

F I j

F II u 1250 F

0

- -- -- -.-..-.--.- F 2 1000 F

1 750

I F F j 500

- I I F

F 250 I I n

5O 100 150 200 250 300 350 RCS TEMPERATURE (DEG. F)

FIGURE 3.4-4 RCS COLD OVERPRESSURE PROTECTION SETPOINTS Page 14 of 29

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS The temperature and pressure changes durin&q heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section +1*, Appendix G:

=Refece Q)

1. The reactor coolant temX ature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon:

a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and

b. Figures 3.4-2 and 3.4-3 define limits to assure prevention ofnon-ductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over c'ertain pressure-temperature ranges.

2. These limit lines shall be calculated periodically using methods provided below, .pag

3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70°F,

4. The pressurizer heatup and cooldown rates shall not exceed 100"F/h and 200°F/h, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320F, and

5. System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of t ASME Boiler and Pressure Vessel Code,Section XI.

-- rt-c ar --

tha at n coraccwthAp-~ C-d-th 1972 t *,1 calculettia 1 metheds deser-ibed 4n W9AP 7924 A, "Basiz for Heatup and Geeldew.,

Hillit eCurvc:," April 197 Is.t]

Heatup and cooldown limit curves are calculated using the 9.6t-l4! i4#4÷"

value ef the nil ductility refcrcPncc temperature, RT amh n f 11.1 effee tive lull power ,ye lar (E,..1 ) ef se, Me l,.,*w.

ife. 6 ". . '3. r,..

-- F

sery.i eew.,i l , per

.if SEABROOK - UNIT 1 B 3/4 4-7 Amendment No. +-l-

,Section II Insert A, Page B3/4 4-7 Operation within the limits of the appropriate heatup and cooldown curves assures the integrity of the reactor vessel's ferritic material against fracture induced by combined thermal and pressure stresses. As the reactor vessel is subjected to increasing fluence, the toughness of the limiting beltline region material continues to diminish, and consequently, even more restrictive pressure/temperature (P/T) limits must be maintained. Each P/T limit curve defines an acceptable region for normal operation during heatup or cooldown maneuvering as pressure and temperature indications are monitored to ensure that operation is within the allowable region. A heatup or cooldown is defined as a temperature change of greater than or equal to 10F in any one-hour period.

The P/T limits have been established in accordance with the requirements of ASME Boiler and Pressure Vessel Code Section XI, Appendix G, as modified by ASME Code Case N-641, Reference (2), and the additional requirements of I OCFR50 Appendix G, Reference (3). The heatup and cooldown P/T limit curves for normal operation, Figures 3.4-2 and 3.4-3 respectively, are valid for a service period of 20 effective full power years (EFPY). The technical justification and methodologies utilized in their development are documented in WCAP-15745, Reference (4). The P/T curves were generated based on the latest available reactor vessel information and latest calculated fluences.

Insert B, Page B3/4 4-7 the adjusted RTNDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTNDT of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTNDT, and adding a margin.

RTNDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT (IRTNDT). The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, Reference (5). Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + ARTNDT + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region.

Page 16 of 29

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS (Continued) schos~4than schthet sgat theoflimitin the 1 i.it-ngTND at th 1/'1/aion in The unirradiatedx-aterial.

co e regionof\

th 'lection h suca limitinri assures hat a components *n\the Reactor .6o nt System jwie opera d cons vat ely in ac rdance w* h ap\icable C e re irements.I The reactor vessel materials have been tested to determine their initial RTNDT, thc r.... f th...tt.... 2..... .o. in T-a.b. B ;j4. . Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an incr. Be in the RTRDT. Therefore, an adjusted reference et Based upon the fluence, copper eontent, and nickel content of the materia .

can be predicted using 14RT. and the value of compu edbvby Regulatory Guide f., Revision .. 1 mbrittlcmr.tt 2,P"Idiatiar Rcictor aessel Materials.* The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predi ed adjustments for this shift in RTWDT at the end of . EFPY as well as adjustm ts for possible errors in the pressure and tem rature sensin instruments Yt ... -,_- werer~u nner may:"

ma.4* use:d ukti4 he resul ts t from the material srvei lance p ogram evaluated according to AST1E185Ja.e aail4ab*. Capsules j re ved inaccordance with the requirements'of ASTM E185-73 and 10CFR5O, Appendix H The lead factor represents the relationship between the fast neutron flux ensity at the location of the capsule and the inner wall of the reactor v eel. Therefore, the results obtained from the surveillance specimens used to predict future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule.ý EvalidetioT ef survcillanee capsule data wiill be eacndueted in Al owable sure-te erature r ionship for iou heatup an cooldown a are calculate usi et ds deriveW rom Appen ix G ie- tion II of eASME Boi er and Pr ure Vess Cod s equired by p ix G 010 art and the e m ods re discus in deta in WCAP 24-A.

The ge er meth d for cal ing he tup and co own limit curves i bas dupon princip s of e inear, elas ic fr ur mechanics LEFMtwi tech gy. I the cal ion pr cedures, a iellipt cal surface fect wih p h of one- ar otewl hike ,,and a le th of I~nsertýB B 3/4 4-8 Amendment No. --4 SEABROOK - UNIT I

,Section II Insert A, Page B3/4 4-8 Surveillance capsule data, documented in Reference (6), is available for two capsules (Capsules U and Y) having already been removed from the reactor vessel. This surveillance capsule data was used to calculate chemistry factor (CF) values per Position 2.1 of Regulatory Guide 1.99, Revision 2. It also noted that Reference (6) concluded that all the surveillance data was credible as the beltline material was behaving as empirically predicted.

Insert B, Page B3/4 4-8 The fluence values used to determine the CFs are the calculated fluence values at the surveillance capsule locations. The calculated fluence values were used for all cases. All calculated fluence values (capsule and projections) are documented in Reference (6). These fluences were calculated using the ENDF/B-VI scattering cross-section data set. The measured ARTNDT values for the weld data were adjusted for chemistry using the ratio procedure given in Position 2.1 of Regulatory Guide 1.99, Revision 2.

Page 18 of 29

a I

I (THIS FIGURE NUMBER IS NOT USED) Amendment No.

S......FIGURE B 3/4.4-1 FAST NEUTRO1N fLuENCE (>-,Ve,) AS A FUNCTION 6F FUEL P,.3. E SERVICE bIFE SEABROOK - UNIT 1 B 3/4 4-9

TABLE B 3/4.4-1 (THIS1 TABLE NUMBER IS NOT USEDý)

- _,,TfI t iV_, i. . "Y I 1L"*

ca 0

0 "C

--I CA) 0.-J

PrA£TnP rfI ANT qYqTFM BASES (

3/4.4.9 PRESSURE/TEMPERATURE LIMITS(Cniud I ssumed to exis t the inside oNhe vessel walloowe as at th o ide of the ves e wall. The d en- ons of thi post*ated crac efe ed to in A endi G of ASM Sectio III as e refer ce fl amply exceed he cur nt capab itie of inservi in ection tech *qu Therefore t reactor ope ti n limit curve eveloped for th reference crack are nservative and vide sufficig fety margins or rotection against n d tile failur T assure tha the d'ation e rittle ent Affects re acc nted fo in the alcula on of t limit rves, th mos imiti g value o the 1-ductilit re rence temper u RTNOT, is u an his includes t radiation-indu shift, ART rresponding te e of the pe od which h tup a cooldow curves re gene ated.

heatup and ldown rates sp ifies that the al stress inte si factor, K1 , for th c bined therma an pressure s esse at any ti dur heatup or cool n cann be gr ter than he re rence st s i ensity fact refe f t metal tempe t re at that tim KKIR is obtaine from the refer fr re toughness r , defined in pp dix G to t AS Code. Th KIR c ye given by he eq tion:

KIR = 6. + 1.223 exp 0 14 5 (T-RTNDT + 0)(

Where: K is e referen stre intensi facto as a fu ction o he me pera re T and e me 1 nil-duct it reference te e ture RTNOTs the verning equatia for the heatup- Idown analysi defined in A en *x G of e SME Code as ol s:

Where: KIM*- stesit i fcor.ca d b ebrane rressu strs t= the ses'tniyf or uedb hehmlgain K'R rel ati'y tohe TDoth tril C2 for insetvi h tat a lea tes oper ins.

At an time du

g t eatup or oodntase sdetermn the t temperature he tip of the stulated flaw, appropriate SEABROOK - UNIT 1 B 3/4-12 Amendment No.7Z T

,Section II Insert A, Page B314 4-12 The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KI., for the metal temperature at that time. K1 t is obtained from the reference fracture toughness curve, defined in Code Case N-641, Reference (2). The Kic curve is given by the following equation:

K., = 3 3 . 2 + 2 0. 7 3 4*e°[O2(T-RTNDT)] (1)

where, Kic = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This Kic curve is based on the lower bound of static critical KI values measured as a function of temperature on specimens of SA-533 Grade B Classl, SA-508-1, SA-508-2, and SA-508-3 steel.

The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C* Kim + Kit < Kic (2)

where, Kim = stress intensity factor caused by membrane (pressure) stress Kit= stress intensity factor caused by the thermal gradients Kl,= function of temperature relative to the RTNDT of the material C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical At any time during the heatup or cooldown transient, K1 c is determined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Kit, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

Page 22 of 29

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS (Continued) reswltir.g frern temperature gradeiint: through the Yessel w&41 are eeakulated and then the orr...onding thermal -stress intensity f..tr, I for the

-referenec flaw: 4s computcd. From Eguaticn (2) the pressure stress 4ntensity are

.bta.ne faeters and, fr... these, the all. abl: pressues are

,aleti'ated COO LDOWN .

COODON I1of Ap endix Gto the ASME Code For the calculation of the allowmabipressure versus coolant temperature during cooldown, the %oe reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the "1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. Thi.s condition, of course, is not true for the steady-state situa tion. It follows that at any given reactor coolant temperature, the AT devel oped during cooldown results in a higher value of Ki at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermo~re, if condi tions exist such that the increase in Ki exceeds Kit, the calculated allowabl pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because ther6 is no direct control on temperature at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.

HEATUP surface Three separate calculations are requ'red to determine the limit curves

.* for finite heatup rates. As is done in e cooldown analysis, allowable

{ pressure-temperature relationships are d veloped for steady-state conditions as well as finite heatup rate conditions ssuming the presence of a 1/4T defect at the inside of the vessel wall. he thermal gradi:nts drnn heatup

.predda compressive stresses at the insideof the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the KiRm for the 1/4T crack duringheatup is lower than the Ki* foýrthe 1/4T crack during steady-state C ~ ~ - Amedtnt No.

SEABROOK - UNIT 1 B 3/4 4-13

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS (Continued)

HEATUP (Continued) . v conditions at the same coolant t mpe ature. During heatup, especially at the end of the transient, conditions;ma* exist such that the effects of compressive thermal stresses and diffeenr KI.R for steady-state and finite heatup rates do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. flaw located at the %Tlocation from the outside The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep awts4de surface f-Iaw-is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stressesT of esps~e7 are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermei-erince the thermal stresses at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.

Rather, each heatup rate of interest must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressuri is taken to be the lesser of the three values taken from the curves under consideration. . wherein, CThe use of the composite curve is necessary to set conservative heatup limitations becaus it if'~ssible for conditions to exist s.eý 4h over the course of the heatup r ý e controlling condition switches from the inside to the outside and the p) ssure limit must at all times be based on analysis of the most critical criterion.

Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.

~Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

SEABROOK - UNIT 1 B 3/4 4-14

, Section HI Insert A, Page B3/4 4-14 10 CFR Part 50, Appendix G, Reference (3), addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNDT by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3106 psi),

which in this case is 621 psig. The limiting unirradiated RTNDT of 30TF occurs in the vessel flange of the reactor vessel, consequently the minimum allowable temperature of this region is 150'F at pressures greater than 621 psig. This limit is shown as the horizontal lines in Figures 3.4-2 and 3.4-3. (Note: Figures 3.4-2 and 3.4-3 include a compensation of 20°F and 100 psig for possible instrument errors.)

Insert B, Page B3/4 4-14 References

1. ASME Boiler and Pressure Vessel Code,Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure", dated December 1995, through 1996 Addendum.

2. ASME Boiler and Pressure Vessel Code Case N-641,Section XI, Division 1, "Alternative Pressure-Temperature Relationship and Overpressure Protection System Requirements", dated January 17, 2000.

3. 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements", U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, dated December 19, 1995.

4. Westinghouse WCAP-15745, Revision 0, "Seabrook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation", dated December 2001.

5. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", U. S. Nuclear Regulatory Commission, dated May 1988.

6. Duke Engineering and Services Report DES-NFQA-98-01, Revision 0, "Analysis of Seabrook Station Unit I Reactor Vessel Surveillance Capsules U and Y", dated May 1998.

Page 25 of 29

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS (Continued)

COLD OVERPRESSURE PROTECTION (Continued)

The OPERABILITY of two PORVs, or two RHR suction relief valves, or a combination of a PORV and RHR suction relief valve, or an RCS vent opening of at least 1.58 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the, 290)

RCS cold legs are less than or equal to 2-F. Either PORV or either RHR suction relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 501F above the RCS cold leg temperatures, or (2) the start of a centrifugal charging pump and its injection into a water-solid RCS.

The Maximum Allowed PORV Setpoint for the Cold Overpressure Mitigation System (COMS) is derived by analysis which models the performance of the COMS assuming various mass input and heat input transients. Operation with a PORV Setpoint less than or equal to the maximum Setpoint ensures that Appendix G criteria will not be violated with consideration for: (1) a maximum pressure overshoot beyond the PORV Setpoint which can occur as a result of time delays in signal processing and valve opening; (2) a 50°F heat transport effect made possible by the geometrical relationship of the RHR suction line and the RCS wide range temperature indicator used for COMS; (3) instrument uncertainties; and (4) single failure. To ensure mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require both Safety Injection pumps and all but one centrifugal charging pump to be made inoperable while in MODES 4, 5, and 6 with the reactor vessel head installed and not fully detensioned, and disallow start of an RCP if secondary coolant temperature is more than 50°F above reactor coolant temperature. Exceptions to these requirements are acceptable as described below.

Operation above 350°F but less than 375°F with only one centrifugal charging pump OPERABLE and no Safety Injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . As shown by analysis, LOCAs occurring at low temperature, low pressure conditions can be successfully mitigated by the operation of a single centrifugal charging pump and a single RHR pump with no credit for accumulator injection. Given the short time duration and the condition of having only one centrifugal charging pump OPERABLE and the probability of a LOCA occurring during this time, the failure of the single centrifugal charging pump is not assumed.

Operation below 350°F but greater than 325 0 F with all centrifugal charging and Safety Injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . During low pressure, low temperature operation all automatic Safety Injection actuation signals except Containment Pressure - High are blocked. In normal conditions, a single failure of the SEABROOK - UNIT 1 B 3/4 4-15 Amendment No. 4 74

REACTOR COOLANT SYSTEM , O-I BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS (Continued)

COLD OVERPRESSURE PROTECTION (Continued)

ESF actuation circuitry will result in the starting of at most one train of Safety Injection (one centrifugal charging pump, and one Safety Injection pump). For temperatures above 3250F, an overpressure event occurring as a result of starting two pumps can be successfully mitigated by operation of both PORVs without exceeding Appendix G limit.

A single failure of a PORV is not assumed due to the short duration that this condition is allowed and the low probability of an event occurring during this interval in conjunction with the failure of a PORV to open. Initiation of both trains of Safety Injection during this 4-hour time frame due to operator error or a single failure occurring during testing of a redundant channel are not considered to be credible accidents.

Operation with all centrifugal charging pumps and both Safety Injection pumps OPERABLE is acceptable when RCS temperature is greater than 350 0 F, a single PORV has sufficient capacity to relieve the combined flow rate of all pumps. Above 350°F two RCPs and all pressure safety valves are required to be OPERABLE. Operation of an RCP eliminates the possibility of a 50°F difference existing between indicated and actual RCS temperature as a result of heat transport effects. Considering instrument uncertainties only, an indicated RCS temperature of 350°F is sufficiently high to allow full RCS pressurization in accordance with Appendix G limitations. Should an overpressure event occur in these conditions, the pressurizer safety valves provide acceptable and redundant overpressure protection.

When operating below 200OF in MODE 5 or MODE 6 with the reactor vessel head on and the vessel head closure bolts not fully detensioned, Technical Specification 3.5.3.2 allows one Safety Injection pump to be made OPERABLE whenever the RCS has a vent area equal to or greater than 18 square inches or whenever the RCS is in a reduced inventory condition, i.e., whenever reactor vessel water level is lower than 36 inches below the reactor vessel flange. Cold overpressure protection provided by the venting method utilizes an 18 square inch or greater mechanical opening in the RCS pressure boundary. This mechanical opening is larger in size than the 1.58 square inch opening required for normal overpressure protection and is of sufficient size to ensure that the Appendix G limits are not exceeded when an SI pump is operating in MODE 5 or MODE 6 with the reactor vessel head on and the vessel head closure bolts not fully detensioned. When the reactor has been shut down for at least 7 days, the larger vent area also enhances the ability to provide a gravity feed to the RCS from the Refueling Water Storage Tank in the unlikely event that the CCP and SI pumps were unavailable after a loss of RHR. Additionally, when steam generator nozzle dams are installed for maintenance purposes and the reactor vessel water level is not in a reduced inventory condition, the larger vent area limits RCS pressure during overpressure transients to reduce the possibility of adversely affecting steam generator nozzle dams.

SEABROOK - UNIT 1 B 3/4 4-16 Amendment No. 3, 74

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS (Continued)

COLD OVERPRESSURE PROTECTION (Continued)

When the reactor vessel head is on and the vessel head closure bolts are fully detensioned, i.e., when the closure nuts have been removed from the studs, a substantial vent area exists by the gap underneath the reactor vessel head, created by the internal spring forces. A measured gap of greater than or equal to 0.03 inches is of sufficient size to provide for cold overpressure protection, for gravity feed from the RWST, and ensuring nozzle dam integrity. Verification of sufficient gap will be performed prior to crediting the gap as a means for cold overpressure protection.

Cold overpressure protection can also be provided when operating at a reduced inventory condition, i.e., whenever reactor vessel water level is lower than 36 inches below the reactor vessel flange. With RCS water level lower than 36 inches below the RV flange in Mode 5 or Mode 6 with the RV head on and the closure bolts not fully detensioned, a mass addition transient involving simultaneous operation of a CCP and a SI pump without letdown will not result in a cold overpressurization condition because of the relatively large void volume in the RCS. This void volume consists of the upper plenum of the reactor vessel and the RV head, the pressurizer and steam generator tubes, as a minimum. The relatively large void volume affords ample time for operator action, (e.g., diagnose the water level increase on main control board instrumentation and stopping the pumps) to mitigate the transient. A minimum time of 50 minutes has been determined based on one charging pump operating at 120 gpm without letdown and a Safety Injection pump injecting into the RCS.

The charging pumps and Safety Injection pumps are rendered incapable of injecting into the RCS through removing the power from th6 pumps by racking the motor circuit breakers out under administrative control. An alternate method of preventing cold overpressurization may be employed. The alternate method uses at least two independent means to prevent cold overpressurization such that a single action will not result in an inadvertent injection into the RCS. This may be accomplished through the pump control switch being placed in Pull-to-Lock position and at least one valve in the discharge flow path closed. The alternate method provides the ability to respond to abnormal situations, expeditiously, from the main control room.

During charging pump swap operation two charging pumps may be made capable of injecting into the RCS for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This provision prevents securing charging for the purpose of not having more than the allowable pumps operable in order to limit thermal fatigue cycles on piping and impact seal injection to the Reactor Coolant Pumps (RCP) which has seal degradation potential. Given the short time duration of the evolution and the evolution controlled under administrative controls, e.g., prohibiting pump swap operation during RCS water-solid conditions, a cold overpressurization condition occurring as a result of an uncontrolled mass addition transient is unlikely.

SEABROOK - UNIT 1 B 3/4 4-16a Amendment No. 74

REACTOR COOLANT SYSTEM For Information Only BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS (Continued)

COLD OVERPRESSURE PROTECTION (Continued)

Charging and/or Safety Injection pumps, normally rendered inoperable for cold overpressure protection may be operated as required under administrative controls during abnormal situations involving a loss of decay heat removal capability or an unexpected reduction in RCS inventory. Maintaining adequate core cooling and RCS inventory during these abnormal situations is essential for public health and safety.

Administrative controls ensure that a cold overpressurization condition will not occur as a result of an uncontrolled mass addition transient.

The Maximum Allowed PORV Setpoint for the Cold Overpressure Mitigation System will be revised on the basis of the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H.

SEABROOK - UNIT 1 B 3/4 4-16b Amendment No. 74 1

,Section III SECTION III RETYPE OF PROPOSED CHANGES Refer to the attached retype of the proposed changes to the Technical Specifications. The attached retype reflects the currently issued version of the Technical Specifications. Pending Technical Specification changes or Technical Specification changes issued subsequent to this submittal are not reflected in the enclosed retype. The enclosed retype should be checked for continuity with Technical Specifications prior to issuance.

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE FIGURE 3.4-1 DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > liCi / gram DOSE EQUIVALENT 1-131 .............................................................................. 3/4 4-28 TABLE 4.4-3 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PRO G RA M....................................................................................................... 3/4 4-29 3/4.4.9 PRESSURE/TEMPERATURE LIMITS G eneral ............................................................................................................. 3/4 4-30 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 20 EFPY ......................................................................... 3/4 4-31 FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 20 EFPY ......................................................................... 3/4 4-321 Pressurizer ....................................................................................................... 3/4 4-33 Overpressure Protection Systems ................................................................... 3/4 4-34 FIGURE 3.4-4 RCS COLD OVERPRESSURE PROTECTION SETPOINTS .................... 3/4 4-36 3/4.4.10 STRUCTURAL INTEGRITY ............................................................................... .3/4 4-37 3/4.4.11 REACTOR COOLANT SYSTEM VENTS ........................................................... 3/4 4-38 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS Hot Standby, Startup, and Power Operation ........................................................ 3/4 5-1 S hutdow n ............................................................................................................. 3/4 5-3 3/4.5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350°F ............ 3/45-4 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 350°F ............................................. 3/4 5-8 ECCS SUBSYSTEMS - Tavg Equal To or Less Than 200°F ............................ 3/4 5-10 3/4.5.4 REFUELING WATER STORAGE TANK .......................................................... 3/45-11 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containm ent Integrity ........................................................................................ 3/4 6-1 Containm ent Leakage ....................................................................................... 3/4 6-2 SEABROOK - UNIT 1 vi Amendment No. 7-0

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.12.2 (THIS SPECIFICATION NUMBER IS NOT USED) ........................................ 3/4 12-3 3/4.12.3 (THIS SPECIFICATION NUMBER IS NOT USED) ........................................ 3/4 12-5 3.0/4.0 BASES 3/4.0 APPLICABILITY ...................................................................................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL ................................................................................... B 3/4 1-1 3/4.1.2 BORATION SYSTEMS ................................................................................. B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES ............................................................. B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS ............................................................................ B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE ............................................................................ B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR ......................................... B 3/4 2-2 3/4.2.4 QUADRANT POWER TILT RATIO ............................................................... B 3/4 2-3 3/4.2.5 DNB PARAMETERS .................................................................................... B 3/4 2-4 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION ............ B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION ........................................................... B 3/4 3-3 3/4.3.4 (THIS SPECIFICATION NUMBER IS NOT USED) ...................................... B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION .................. B 3/4 4-1 3/4.4.2 SAFETY VALVES ............................................................................................ B 3/4 4-1 3/4.4.3 PRESSURIZER ............................................................................................... B 3/4 4-2 3/4.4.4 RELIEF VALVES ............................................................................................. B 3/4 4-2 3/4.4.5 STEAM GENERATORS .................................................................................. B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE ................................................... B 3/4 4-3 3/4.4.7 CHEM ISTRY .................................................................................................... B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY ....................................................................................... B 3/4 4-5 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ............................................................ B 3/4 4-7 FIGURE B 3/4.4-1 (THIS FIGURE NUMBER IS NOT USED) ....................................... B 3/4 4-9 FIGURE B 3/4.4-2 (This figure number not used) ............................................................ B 3/4 4-10 74 SEABROOK - UNIT I X Amendment No. 50

, 66 ,

INDEX BASES SECTION PAGE TABLE B 3/4.4-1 (THIS TABLE NUMBER IS NOT USED) ............................................ B 3/4 4-11 3/4.4.10 STRUCTURAL INTEGRITY ........................................................................... B 3/4 4-16 3/4.4.11 REACTOR COOLANT SYSTEM VENTS ...................................................... B 3/4 4-16 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS ......................................................................................... B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS .................................................................... B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK ....................................................... B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT ........................................................................... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS ...................................... B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES ......................................................... B 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTROL .................................................................. B 3/4 6-3 3/4.6.5 CONTAINMENT ENCLOSURE BUILDING ................................................... B 3/4 6-4 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ........................................................................................... B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION ............ B 3/4 7-3 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM .............................. B 3/4 7-3 3/4.7.4 SERVICE WATER SYSTEM ......................................................................... B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK ................................................................................. B 3/4 7-3 3/4.7.6 CONTROL ROOM SUBSYSTEMS ............................................................... B 3/4 7-4 3/4.7.7 SNUBBERS .................................................................................................... B 3/4 7-4 3/4.7.8 SEALED SOURCE CONTAMINATION ......................................................... B 3/4 7-5 3/4.7.9 (THIS SPECIFICATION NUMBER IS NOT USED) ....................................... B 3/4 7-5 3/4.7.10 (THIS SPECIFICATION NUMBER IS NOT USED) ....................................... B 3/4 7-5 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION ................................................................. B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES ................................. B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION ......................................................................... B 3/4 9-1 3/4.9.2 INSTRUMENTATION ..................................................................................... B 3/4 9-1 3/4.9.3 DECAY TIM E ................................................................................................. B 3/4 9-1 3/4.9.4 CONTINMENT BUILDING PENETRATIONS ................................................ B 3/4 9-1 3/4.9.5 COMMUNICATIONS ...................................................................................... B 3/4 9-1 3/4.9.6 REFUELING MACHINE ................................................................................. B 3/4 9-1 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING ............................. B 3/4 9-2 SEABROOK - UNIT 1 xi Amendment No. 56,63

MATERIAL PROPERTY BASIS Umiting material: LOWER SHELL PLATE R-1808-1 Limiting ART values at 20 EFPY.1/4T, 1097F 314T, 88"F Curves applicable for the first 20 EFPY and contain margins of 20'F and 100 psig for possible Instrument errors 2800

- I- T a I I - -I - T a a a-1 r a a a - - I.a a a a I a a a a a a I a a a a a a a a la a a a I I I I I I I I I I I a I a a i a a I a I I I I 2600 S a i a a a a a a i

"---"-------------T--r----------------------

a 1 a T a a .- a a 2400 a LeaklTestLmlrnt- a ,ir rn ie aIaa a a a a a aI I a a 2200 pt I '...L------ -U -L4....a=......------... ...

z 5 2000 SI a A.41-_°I--4IOperation Operation nacaa a balae I .. -I a i - AI-,a a a a a a ap4 ea a al a a a n eT T io n -t -, l - -lI -- - T----, n ert -I T n ,r'n'-I-,

Uo 080 -- . - -I- - -------------- ------ --------------

ar3 ir1800" 1600"

.. ... e.. -41 - .- . .. e -1Iar- 4- - 1- . L4-. -1 -- - -

a a H atp ate I_ - - - -

a a a Critical Limita L J _LI I a a J_

a a L J a a 1400" -- a--- - - - - - -F- -r--l a a Hea'upRate --- -

a a a a a CriticalLimit

"= - r -

a da a a a a a

.--- 1 -.. .....-i Ia-4--..aL - .I..- .a.-. - a- a ----.4 .--.----. a--I- - 4. J--a..

aI ..a - a..4..a1aI ..aI.. a -.. 4-a. -. a 1200 U) 0 - - -- - - - -- -

n 1000 ... a+/-.L..J........L.J...LJI..a...L...a... L...4...L.

-I -1 -' - 4I -

aL ._.

4 w --. I- -a - -.-- a-a. --- a- a-.,. .a- -a &---a.-,-a- ,--a--a---,-,, - ,.-,.-I-,-..,--a-a U)800" a. a .. a aI a a a a

-r

-a--------------- Crtcit Limit-,-,-

bae -r on, F",---

6000 -.--.J _a.--.a. J. - ..L- -I. .L, J.. I-- I.- InservJ- Icehy / rsai .I- test. r.-J -L- J . .t.L

- - ----- - --- -- - a-,-a-I- ---- -- - - -- I--a--a-- - --- a---- -a--a--

a a I BotuIa a II I I I Ia a a 400 "-i----a Temp--------------r-1-a Cia a.. ml basedaon a a a a aI a a a a a a a ai a a a a ansaaeaarvaa a, ,,a. a a. . ae a _a . a . .

a a a a a a a a a aI 200 a a-- a.. a. a a .aa a1 a0"a a'-, 20 E aPa aa-a .aa.

a ' aU seva peid 0 I I I I 0 100 200 300 400 500 RCS TEMPERATURE (Deg. F, 20 Deg. F PER DIVISION)

FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -APPLICABLE UP TO 20 EFPY SEABROOK - UNIT 1 3/4 4-31 Amendment No. 4-9

MATERIAL PROPERTY BASIS Limiting material: LOWER SHELL PLATE R-1808-1 Limiting ART values at 20 EFPY.I/4T, 1097 3/4T, 88"F Curves applicable for the first 20 EFPY and contain margins of 20°F and 100 psig for possible Instrument errors 280O S- I a I I a -J a a i L a

. I- T I - - I - T -I - - , -I - - - - I-- , - 1 T - - - 1 T I- - r 2600 I a ,a a a i aa

..- . . L_ . . ,L . -J_,_,, _a, . . ..,L 4 a

. J_- . -.. .I-_ . ... , . . ..-1L . . ... I. , _ _-_

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a a a I I

a Ia a a a a a I

.. aL , _ L_,- j.J_-, - ... .L .. -L...4 - ,- -- .,- - -, - ,-a 2400 a a - 4Ia - i - aI I a a a a a a a a a 2200 I tI I +. - I :- -I ai- aTa - I a ia a I -.--I Ia a L a I a a a - I I I a a I a

, OperatIon T - -i - - I - - TI ,- - I-I -,-- I ----- r Oeration T -r o 2000

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n- r . .-- -i r - -t- -rL - J-1 - -r.- -

I- rJ

- -J -I m 1800 ---- "--'-------- oldown Rate ---- *- -

a. -, 10 De-a , u1to100De F/Hr* a a a a a j 1600 -F I T Ia a a aI -a I e.

T 1400 -- 4.. . J --- --. & -. ,-,-,----.--.,--.,.--. .,----,. ---. ..

a.5 I 6 I D eg /H 4- -4I - t- I I I 4 - -I 4 I I- - 4 - + I4 4 - I

-- 1200

- ""1----r-----"-"1---- * --- - --- ---.----- ,-------,----,-------,-------,--.--------,-------,---

cc -Iolo n aes a a a a a a a a a a aI I6 Deg F/Hr II U, 1000 - -- - - - - - I1 I I {_.1

I LU cc a ~ltu " ' "T "1. . . a a a- a*

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a. . aI a - a a a.. a -a Ia I

I , I l 8O1 I OO, giFlhl I I I J I I I I l cl 600 -,- ,a--,a-,a ,

I

.a-,a.

a a-a a

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a a a a a a aI a_. a,. ,a 0

cC a a I B lu F l a a a a a a a a a a a a aI 600 a a I e p a a a a a a a a a a a 1a, 400 200 0 I I "II 0 100 200 300 400 500 RCS TEMPERATURE (Deg. F, 20 Deg. F PER DIVISION)

FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS - APPLICABLE UP TO 20 EFPY SEABROOK - UNIT 1 3/4 4-32 Amendment No. 419

REACTOR COOLANT SYSTEM PRESSURE/TEMPERATURE LIMITS OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 The following Overpressure Protection Systems shall be OPERABLE:

a. In MODE 4 when the temperature of any RCS cold leg is less than or equal to 290 0F; and in MODE 5 and MODE 6 with all Safety Injection pumps inoperable at least one of the following groups of two overpressure protection devices shall be OPERABLE when the RCS is not depressurized with an RCS vent area of greater than or equal to 1.58 square inches:

1) Two residual heat removal (RHR) suction relief valves each with a setpoint of 450 psig +0, -3 %; or

2) Two power-operated relief valves (PORVs) with lift setpoints that vary with RCS temperature which do not exceed the limit established in Figure 3.4-4, or

3) One RHR suction relief valve and one PORV with setpoints as required above.

b. In MODE 5 and MODE 6 with all Safety Injection pumps except one inoperable:

1) The Reactor Coolant System (RCS) depressurized with an RCS vent area equal to or greater than 18 square inches, or

2) The RCS in a reduced inventory condition*.

APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 290°F; MODE 5 and MODE 6 with the reactor vessel head on and the vessel head closure bolts not fully detensioned.

ACTION:

a) In MODE 4 with all Safety Injection pumps inoperable and with one of the two required overpressure protection devices inoperable, either restore two overpressure protection devices to OPERABLE status within 7 days or within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (a) depressurize the RCS and (b) vent the RCS through at least a 1.58-square-inch vent.

A reduced inventory condition exists whenever reactor vessel (RV) water level is lower than 36 inches below the RV flange.

SEABROOK - UNIT 1 3/4 4-34 Amendment No. , 5, 6, -74

VALID FOR THE FIRST 20 EFPY, SETPOINT CONTAINS MARGIN OF 50°F FOR TRANSIENT EFFECTS 0

T

200.0 F, P = 561.0 PSIG; 200.0°F < T 5 230.5 0F, P = 12.1*(T-200.0) + 926.0 PSIG; 230.50F < T

255.00F, P = 23.15*(T-230.5) + 1295.05 PSIG; T> 255.0°F, P = 34.5*(T-255.0) + 1862.225 PSIG 2500 2250 I

2000 1750

I II

?/ S 1500 0

'n 1250 11000 750 500 I I I*

S..

. ... . . .. * . . . .L1... . . . . . . . . . . . .. .i . . . ... .. . . ... . . .

250 0

510 100 150 200 250 300 350 RCS TEMPERATURE (DEG. F)

FIGURE 3.4-4 RCS COLD OVERPRESSURE PROTECTION SETPOINTS SEABROOK - UNIT 1 3/4 4-36 Ame ndment No.

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section XI, Appendix G, Reference (1):

1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon:

a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and

b. Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2. These limit lines shall be calculated periodically using methods provided below,

3. The secondary side of the steam generator must not be pressurized above 200 psig ifthe temperature of the steam generator is below 70 0F,

4. The pressurizer heatup and cooldown rates shall not exceed 100°F/h and 200°F/h, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320 0 F, and

5. System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

Operation within the limits of the appropriate heatup and cooldown curves assures the integrity of the reactor vessel's ferritic material against fracture induced by combined thermal and pressure stresses. As the reactor vessel is subjected to increasing fluence, the toughness of the limiting beltline region material continues to diminish, and consequently, even more restrictive pressure/temperature (P/T) limits must be maintained. Each P/T limit curve defines an acceptable region for normal operation during heatup or cooldown maneuvering as pressure and temperature indications are monitored to ensure that operation is within the allowable region. A heatup or cooldown is defined as a temperature change of greater than or equal to 10°F in any one-hour period.

SEABROOK - UNIT 1 B 3/4 4-7 Amendment No. 4-9

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS (Continued)

The P/T limits have been established in accordance with the requirements of ASME Boiler and Pressure Vessel Code Section XI, Appendix G, as modified by ASME Code Case N-641, Reference (2), and the additional requirements of 10CFR50 Appendix G, Reference (3). The heatup and cooldown P/T limit curves for normal operation, Figures 3.4-2 and 3.4-3 respectively, are valid for a service period of 20 effective full power years (EFPY).

The technical justification and methodologies utilized in their development are documented in WCAP-15745, Reference (4). The P/T curves were generated based on the latest available reactor vessel information and latest calculated fluences.

Heatup and Cooldown limit curves are calculated using the adjusted RTNDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTNDT of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTNDT, and adding a margin.

RTNDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT (IRTNDT). The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, Reference (5). Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + ARTNDT + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region.

The reactor vessel materials have been tested to determine their initial RTNDT.

Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence, best estimate copper and nickel content of the limiting beltline material, can be predicted using surveillance capsule data and the value of ARTNDT computed by Regulatory Guide 1.99, Revision 2. Surveillance capsule data, documented in Reference (6), is available for two capsules (Capsules U and Y) having already been removed from the reactor vessel. This surveillance capsule data was used to calculate chemistry factor (CF) values per Position 2.1 of Regulatory Guide 1.99, Revision 2. It also noted that Reference (6) concluded that all the surveillance data was credible as the beltline material was behaving as empirically predicted. The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RTNDT at the end of 20 EFPY as well as adjustments for possible errors in the pressure and temperature sensing instruments.

SEABROOK - UNIT 1 B 3/4 4-8 Amendment No. 4-9

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS (Continued)

The results from the material surveillance program were evaluated according to ASTM E185. Capsules U and Y were removed in accordance with the requirements of ASTM E185-73 and 10CFR50, Appendix H. The lead factor represents the relationship between the fast neutron flux density at the location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens were used to predict future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. The fluence values used to determine the CFs are the calculated fluence values at the surveillance capsule locations. The calculated fluence values were used for all cases. All calculated fluence values (capsule and projections) are documented in Reference (6). These fluences were calculated using the ENDF/B-VI scattering cross-section data set. The measured ARTNDT values for the weld data were adjusted for chemistry using the ratio procedure given in Position 2.1 of Regulatory Guide 1.99, Revision 2.

FIGURE B 3/4.4-1 (THIS FIGURE NUMBER IS NOT USED)

SEABROOK - UNIT I B 3/4 4-9 Amendment No.

TABLE B 3/4.4-1 (THIS TABLE NUMBER IS NOT USED)

SEABROOK - UNIT 1 B 3/4 4-11 Amendment No.

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS (Continued)

The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K1, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, Kjc, for the metal temperature at that time. Kic is obtained from the reference fracture toughness curve, defined in Code Case N-641, Reference (2).

The K10 curve is given by the following equation:

K,, --3 3 .2 +2 0. 7 3 *e[

4 02r-RTNDT)] ()

where, Kic = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This Kic curve is based on the lower bound of static critical K, values measured as a function of temperature on specimens of SA-533 Grade B Class 1, SA-508-1, SA-508-2, and SA-508-3 steel.

The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C* Kim + Kit < Kic (2)

where, Kim = stress intensity factor caused by membrane (pressure) stress Kt= stress intensity factor caused by the thermal gradients Kjc = function of temperature relative to the RTNDT of the material C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical At any time during the heatup or cooldown transient, Kic is determined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Kit, for the reference flaw are computed.

From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

SEABROOK - UNIT 1 B 3/4 4-12 Amendment No.

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS (Continued)

COOLDOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situation. It follows that at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of Kic at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist such that the increase in Kic exceeds K1t, the calculated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.

HEATUP Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 114T defect at the inside of the vessel wall. The heatup results in compressive stresses at the inside surface of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the Kic for the 114T crack during heatup is lower than the Kic for the 1/4T crack during steady-state conditions at the same coolant temperature.

SEABROOK - UNIT I B 3/4 4-13 Amendment No.

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS (Continued)

HEATUP (Continued)

During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and lower Kic values for steady-state and finite heatup rates do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.

Rather, each heatup rate of interest must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.

SEABROOK - UNIT 1 B 3/4 4-14 Amendment No.

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS (Continued)

HEATUP (Continued) 10 CFR Part 50, Appendix G, Reference (3), addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNDT by at least 120°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3106 psi), which in this case is 621 psig. The limiting unirradiated RTNDT of 30°F occurs in the vessel flange of the reactor vessel, consequently the minimum allowable temperature of this region is 150°F at pressures greater than 621 psig. This limit is shown as the horizontal lines in Figures 3.4-2 and 3.4-3. (NOTE: Figures 3.4-2 and 3.4-3 include a compensation of 20°F and 100 psig for possible instrument errors.)

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

References

1. ASME Boiler and Pressure Vessel Code,Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure", dated December 1995, through 1996 Addendum.

2. ASME Boiler and Pressure Vessel Code Case N-641,Section XI, Division 1, "Alternative Pressure-Temperature Relationship and Overpressure Protection System Requirements", dated January 17, 2000.

3. 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements", U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, dated December 19, 1995.

4. Westinghouse WCAP-15745, Revision 0, "Seabrook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation", dated December 2001.

5. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", U. S. Nuclear Regulatory Commission, dated May 1988.

6. Duke Engineering and Services Report DES-NFQA-98-01, Revision 0, "Analysis of Seabrook Station Unit I Reactor Vessel Surveillance Capsules U and Y", dated May 1998.

SEABROOK - UNIT 1 B 3/4 4-14a Amendment No. I

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS (Continued)

COLD OVERPRESSURE PROTECTION (Continued)

The OPERABILITY of two PORVs, or two RHR suction relief valves, or a combination of a PORV and RHR suction relief valve, or an RCS vent opening of at least 1.58 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 290 0 F. Either PORV or either RHR suction relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50°F above the RCS cold leg temperatures, or (2) the start of a centrifugal charging pump and its injection into a water-solid RCS.

The Maximum Allowed PORV Setpoint for the Cold Overpressure Mitigation System (COMS) is derived by analysis which models the performance of the COMS assuming various mass input and heat input transients. Operation with a PORV Setpoint less than or equal to the maximum Setpoint ensures that Appendix G criteria will not be violated with consideration for: (1) a maximum pressure overshoot beyond the PORV Setpoint which can occur as a result of time delays in signal processing and valve opening; (2) a 50°F heat transport effect made possible by the geometrical relationship of the RHR suction line and the RCS wide range temperature indicator used for COMS; (3) instrument uncertainties; and (4) single failure. To ensure mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require both Safety injection pumps and all but one centrifugal charging pump to be made inoperable while in MODES 4, 5, and 6 with the reactor vessel head installed and not fully detensioned, and disallow start of an RCP if secondary coolant temperature is more than 50°F above reactor coolant temperature.

Exceptions to these requirements are acceptable as described below.

Operation above 350°F but less than 375 0 F with only one centrifugal charging pump OPERABLE and no Safety Injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . As shown by analysis, LOCAs occurring at low temperature, low pressure conditions can be successfully mitigated by the operation of a single centrifugal charging pump and a single RHR pump with no credit for accumulator injection. Given the short time duration and the condition of having only one centrifugal charging pump OPERABLE and the probability of a LOCA occurring during this time, the failure of the single centrifugal charging pump is not assumed.

Operation below 3501F but greater than 325 0 F with all centrifugal charging and Safety Injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . During low pressure, low temperature operation all automatic Safety Injection actuation signals except Containment Pressure-High are blocked. In normal conditions, a single failure of the SEABROOK - UNIT 1 B 3/4 4-15 Amendment No. 3 , 16, 74

SECTION IV DETERMINATION OF SIGNIFICANT HAZARDS FOR PROPOSED CHANGES

,Section IV IV. DETERMINATION OF SIGNIFICANT HAZARDS FOR PROPOSED CHANGES License Amendment Request (LAR) 02-04 proposes changes to the Seabrook Station Technical Specifications (TSs) 3.4.9.1, "Reactor Coolant System - Pressure/Temperature Limits" and 3.4.9.3, "Reactor Coolant System - Overpressure Protection Systems". Specifically, the proposed changes will replace Technical Specification Figure 3.4-2, "Reactor Coolant System Heatup Limitations", Figure 3.4-3, "Reactor Coolant System Cooldown Limitations" and Figure 3.4-4, "RCS Cold Overpressure Protection Setpoints". The proposed change of the reactor vessel pressure-temperature (P/T) limit curves is required because the existing referenced curves are valid through the attainment of 11.1 Effective Full Power Years (EFPY). Based on current projections, 11.1 EFPY will be achieved early in the next operating cycle, Cycle No. 10, at the end of the fourth quarter of 2003. The revised reactor vessel P/T limit curves will remain valid for 20 EFPY as demonstrated in the analysis documented in Reference (1). The P/T limit curves were generated using the most recent reactor vessel surveillance capsule data, which is documented in Reference (3).

An evaluation of the proposed change has been performed in accordance with 10CFR50.91(a)(1) regarding no significant hazards considerations using the criteria in 10CFR50.92(c). A discussion of these criteria as they relate to this amendment request follows:

1. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes to TS 3.4.9.1 and TS 3.4.9.3 do not result in a condition where the design, material, and construction standards that were applicable prior to the proposed changes are altered. The probability of occurrence of an accident previously evaluated for Seabrook Station is not altered by the proposed amendment to the TSs. The accidents remain the same as currently analyzed in the UFSAR as a result of changes to the P/T limits as well as those for Cold Overpressure Mitigation System (COMS). The new P/T limits are based on NRC accepted methodology along with American Society of Mechanical Engineers (ASME) Code alternative methodology. An exemption request to allow use of the alternative ASME methodology is included as part of this LAR. The proposed COMS setpoint limit based on the revised P/T limits satisfies the criteria specified in the alternative ASME methodology and 10 CFR Part 50 Appendix G closure head/vessel flange region pressure limit criteria. The proposed changes do not impact the integrity of the reactor coolant pressure boundary (RCPB) i.e. there is no change to the operating pressure, materials, system loadings, etc., as a result of this change. In addition, there is no increase in the potential for the occurrence of a loss of coolant accident. The probability of any design basis accident is not affected by this change, nor are the consequences of any design basis accident (DBA) affected by this proposed change. The proposed P/T limit curves and the COMS limits are not considered to be an initiator or contributor to any accident currently, evaluated in the Seabrook Station UFSAR. These new limits ensure the long term structural integrity of the RCPB.

Fracture toughness test data are obtained from beltline material specimens contained in Page 1 of 3

,Section IV surveillance capsules that are periodically withdrawn from the reactor vessel. This data allows determination of time conditions under which the vessel can be operated with adequate safety margins against non-ductile fracture throughout its service life. The second Seabrook Station surveillance capsule was removed from the reactor vessel after completion of Operating Cycle No. 5 in May 1997 and was analyzed to predict the fracture toughness requirements using projected neutron fluence calculations. For each analyzed transient and steady state condition, the allowable pressure is determined as a function of reactor coolant temperature considering postulated flaws in the reactor vessel beltline region material. The predicted radiation induced ARTNDT was calculated using the respective reactor vessel beltline materials copper and nickel contents and the neutron fluence predicted for 20 EFPY. The RTNDT and, accordingly, the operating limits for Seabrook Station were adjusted to account for the effects of irradiation on the fracture toughness of the reactor vessel beltline materials. Therefore, new operating limits are established which are represented in the revised operating curves for heatup/cooldown, criticality and inservice hydrostatic testing contained in the technical specifications. The proposed P/T limit curves and COMS setpoint limits are not considered to be an initiator or contributor to any accident currently evaluated in the Seabrook Station UFSAR.

Therefore based on the above discussion, it is concluded that the proposed revisions to TS 3.4.9.1 and TS 3.4.9.3 do not involve a significant increase in the probability or consequences of any accident previously evaluated.

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

The proposed changes to the P/T and COMS limits will not create a new accident scenario. The requirements to have P/T and COMS protection are part of the licensing basis for Seabrook Station. The proposed technical specification amendment reflects the change in reactor vessel material properties as determined by evaluation of the most recently withdrawn surveillance capsule. Based on the surveillance capsule data, the adjusted RTNDT values for the plate and weld material were within the two standard deviations of Regulatory Guide 1.99, Revision 2 predictions. As all the requisite criteria of Regulatory Guide 1.99, Revision 2 was satisfied, it was concluded that the surveillance data was credible and the beltline material was responding as empirically predicted. The new P/T limits are based on NRC accepted methodology along with American Society of Mechanical Engineers (ASME) Code alternative methodology. An exemption request to allow use of the alternative ASME methodology is included as part of this LAR. The proposed COMS setpoint limit based on the revised P/T limits satisfies the criteria specified in the alternative ASME methodology and 10 CFR Part 50 Appendix G closure head/vessel flange region pressure limit criteria. The proposed changes will not alter the way any structure, system or component functions, and will not significantly alter the manner in which the plant is operated. There will be no adverse effect on plant operation or accident mitigation equipment.

Page 2 of 3

,Section IV Since no new failure modes are created by the proposed revisions to TS 3.4.9.1 and TS 3.4.9.3, this change does not create the possibility of a new or different kind of accident from any that was previously evaluated.

3. The proposed changes do not involve a significant reduction in the margin of safety.

The existing P/T and COMS limit curves in the technical specifications are reaching their expiration for the number of years at effective full power operation. The revision of the P/T limits and COMS will ensure that Seabrook Station continues to operate within the operating limits allowed by 10CFR50.60 and the ASME Code. The material properties used in the development of the revised limit curves are based on the evaluation of the most recently withdrawn surveillance capsule. The application of ASME Code Case N-641 presents alternative methods for calculating P/T and COMS temperature and pressure limits in lieu of those established in ASME Section XI, Appendix G-2215. This ASME Code alternative allows analysis features that are less restrictive than those associated with previous methodologies, however these features remain conservative with respect to the requirements delineated ASME Section XI. Therefore it is concluded that the revised P/T and COMS limit curves proposed by this technical specification amendment still provide sufficient margin to preclude non-ductile fracture of the reactor vessel.

Thus, it is concluded that these proposed revisions to TS 3.4.9.1 and TS 3.4.9.3 do not involve a significant reduction in a margin of safety.

Therefore, based upon the evaluation presented above and the previous discussion of the amendment request, North Atlantic concludes that the proposed revisions to TS 3.4.9.1 and TS 3.4.9.3 do not constitute a significant hazard as defined by the criteria in 10CFR50.92(c).

Page 3 of 3

SECTIONS V AND VI PROPOSED SCHEDULE FOR LICENSE AMENDMENT ISSUANCE AND EFFECTIVENESS AND ENVIRONMENTAL IMPACT ASSESSMENT

, Sections V and VI V. PROPOSED SCHEDULE FOR LICENSE AMENDMENT ISSUANCE AND EFFECTIVENESS North Atlantic requests NRC review of License Amendment Request 02-04, and issuance of a license amendment by September 30, 2003, having immediate effectiveness and implementation within 60 days. Issuance of a license amendment by the requested date would afford North Atlantic the flexibility for planning of technical resources in support of Seabrook Station's Tenth Operating Cycle scheduled to commence in the fourth quarter of 2003.

VI. ENVIRONMENTAL IMPACT ASSESSMENT North Atlantic has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. 10CFR51.22(c) provides criteria for and identification of licensing and regulatory actions eligible categorical exclusion from performing 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 hazard consideration, (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite or (3) result in a significant increase in individual or cumulative occupational radiation exposure. North Atlantic has reviewed this proposed license amendment and has determined that it meets the eligibility criteria for categorical exclusion set forth in 10CFR51.22(c)(9). Pursuant to 10CFR51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the proposed license amendment.

The basis for this determination is as follows:

1. The proposed license amendment does not involve a significant hazards consideration described previously in this evaluation (Section IV).

2. As discussed in the significant hazards evaluation, the proposed amendment does not result in a significant change or significant increase in the radiological doses for any Design Basis Accident. The proposed license amendment does not result in a significant change in the types or a significant increase in the amounts of any effluents that may be released off-site.

3. The proposed license amendment does not result in a significant increase to the individual or cumulative occupational radiation exposure because the proposed update of the operating P/T and COMS limits does not impact the exposure of plant personnel.

Based on the preceding discussion, North Atlantic concludes that the proposed changes meet the criterion delineated in 10 CFR 51.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement.

Page 1 of 1

ENCLOSURE 2 TO NYN-02093

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15745 Seabrook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation T. J. Laubham December 2001 Prepared by the Westinghouse Electric Company LLC for the North Atlantic Energy Services Corporation C. H. Boyd, Manager Engineering and Materials Technology Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355

ii PREFACE This report has been technically reviewed and verified by:

J.H. Ledger ________

liii TABLE OF CONTENTS LIST OF TABLES .................................................................................................................................. iv LIST OF FIGURES ................................................................................................................................. v EXECUTIVE

SUMMARY

................................................................................................... vi 1 INTRODUCTION .............................................................................................................. 1 2 FRACTURE TOUGHNESS PROPERTIES ........................................................................... 2 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS .............. 7 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE ..................................... 11 5 HEATUP AND COOLDOWN PRES SURE-TEMPERATURE LIMIT CURVES ................... 19 6 REFERENCES ......................................................................................................................... 28

iv LIST OF TABLES Table 1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RT=r Values for the Seabrook Unit 1 Reactor Vessel Materials ........................................................ 3 Table 2 Calculated Integrated Neutron Exposure of the Surveillance Capsules @ Seabrook Unit I ............................................................................................................................ 4 Table 3 Calculation of Chemistry Factors using Seabrook Unit 1 Surveillance Capsule Data ........ 5 Table 4 Summary of the Seabrook Unit 1 Reactor Vessel Beltline Material Chemistry Factors ...... 6 Table 5 Calculated Neutron Fluence Projections at Key Locations on the Reactor Vessel Clad/Base Metal Interface (1019 n/cra2, E > 1.0 MeV) .............................................. 12 Table 6 Summary of the Vessel Surface, 1/4T and 3/4T Fluence Values used for the Generation of the 16 and 20 EFPY Heatup/Cooldown Curves .................................... 12 Table 7 Summary of the Calculated Fluence Factors used for the Generation of the 16 and 20 EFPY Heatup and Cooldown Curves ................................................................... 13 Table 8 Calculation of the ART Values for the 1/4T Location @ 16 EFPY .............................. 14 Table 9 Calculation of the ART Values for the 3/4T Location @ 16 EFPY ............................ 15 Table 10 Calculation of the ART Values for the 1/4T Location @ 20 EFPY .............................. 16 Table 11 Calculation of the ART Values for the 3/4T Location @ 20 EFPY .............................. 17 Table 12 Summary of the Limiting ART Values Used in the Generation of the Seabrook Unit 1 Heatup/Cooldown Curves ........................................................................................ 18 Table 13 16 EFPYHeatup Curve Data Points Using 1996App. G w/Kic (without Uncertainties for Instrumentation Errors) .......................................................... 22 Table 14 16 EFPY Cooldown Curve Data Points Using 1996 App. G w/Kic (without Uncertainties for Instruminentation Errors) .......................................................... 23 Table 15 20 EFPY Heatup Curve Data Points Using 1996 App. G w/Kic (without Uncertainties for Instrumentation Errors) .......................................................... 26 Table 16 20 EFPY Cooldown Curve Data Points Using 1996 App. G w/Kic (without Uncertainties for Instrumentation Errors) .......................................................... 27

v LIST OF FIGURES Figure 1 Seabrook Unit I Reactor Coolant System Heatup Limitations (Heatup Rates of 80 & 100 0F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors) Using 1996 App. G Methodology w/Kic .................... 20 Figure 2 Seabrook Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100*F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors) Using 1996 App. G Methodology w/Kic ............................... 21 Figure 3 Seabrook Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 80 & I00°F/hr) Applicable for the First 20 EFPY (Without Margins for Instrumentation Errors) Using 1996 App. G Methodology w/Kic ............................... 24 Figure 4 Seabrook Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 20 EFPY (Without Margins for Instrumentation Errors) Using 1996 App. G Methodology w/Kic ............................... 25

vi EXECUTIVE

SUMMARY

This report provides the methodology and results of the generation of heatup and cooldown pressure temperature limit curves for normal operation of the Seabrook Unit I reactor vessel. The PT curves were generated based on the latest available reactor vessel information and latest calculated fluences. The new Seabrook Unit 1heatup and cooldown pressure-temperature limit curves were generated using ASME Code Case N-641t 33 (which allows the use of the K1. methodology) and the axial flaw methodology of the 1995 ASME Code,Section XI through the 1996 Addenda. It should be noted that the Seabrook reactor vessel was limited by the lower shell plate R-1808-1. The pressure-temperature (PT) limit curves and data points are presented in Section 5.

1 INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RT r (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RThiyr of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced AkT-r, and adding a margin. The unirradiated RTmT is designated as the higher of either the drop weight nil ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb 0

of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60 F.

RTNDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTr at any time period in the reactor's life, ART=T due to the radiation exposure associated with that time period must be added to the unirradiated RTmw (IRT~T). The extent of the shift in RTNT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, 'Radiation Embrittlement of Reactor Vessel Materials.'['] Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTN*D + ART~w + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at tle beItline region measured from the clad/base metal interface.

The heatup and cooldown curves documented in this report were generated using the most limiting ART values and the NRC approved methodology documented in WCAP-14040-NP-A, Revision 2 1'3, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" with exception of the following: 1) The fluence values used in this report are calculated fluence values, not the best estimate fluence values. 2) The K10 critical stress intensities are used in place of the K1, critical stress intensities. This methodology is taken from approved ASME Code CaseN-641E'l. 3) The 1996 Version of Appendix G to Section XId4l will be used rather than the 1989 version.

2 2 FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic materials in the reactor coolant pressure boundary are determined in accordance with the NRC Standard Review Plan!S]. The beltilne material properties of the Seabrook Unit 1 reactor vessel is presented in Table 1.

Best estimate copper (Cu) and nickel (Ni) weight percent values used to calculate chemistry factors (CF) in accordance with Regulatory Guide 1.99, Revision 2, are provided in Table 1. Additionally, surveillance capsule data is available for two capsules (Capsules U and Y) already removed from the Seabrook Unit 1 reactor vessel. This surveillance capsule data was also used to calculate CF values per Position 2.1 of Regulatory Guide 1.99, Revision 2 in Table 2. These CF values are summarized in Table 3. It should be noted that all the surveillance data has been determined to be credible.

The Regulatory Guide 1.99, Revision 2 methodology used to develop the heatup and cooldown curves documented in this report is the same as that documented in WCAP-14040, Revision 2.

3 TABLE 1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RTMDT Values for the Seabrook Unit 1 Reactor Vessel Materials Material Description Cu (%) Ni(%) Initial RTXrrD )

Closure Head Flange -- - 0.76 10°F Vessel Flange -- - 0.73 30cF Intermediate Shell Plate R-1806-1(d) 0.045 0.61 40°F Intermediate Shell Plate R-I806-2(d) 0.06 0.64 0°F Intermediate Shell Plate R-1806-3("' 0.075 0.63 10°F Lower Shell Plate R-1808-1(d) 0.06 0.58 40°F Lower Shell Plate R-1808-2(d) 0.06 0.58 10F Lower Shell Plate R-1808-3(d) 0.07 0.59 40OF Beltline Weld Seams 0.047 0.049 .60o*1c)

(Heat # 4P6052)()

Seabrook Unit 1 Surveillance Weld 0.02 0.075 --

(Heat # 4P6052)(c)

Notes (a) The Beltline Weld Seams Consist of the Intermediate Shell Longitudinal Welds (101-124AB,C), the Lower Shell Longitudinal Welds (101-142A,B,C) and the Intermediate to Lower Shell Girth Weld (101 171). These welds were fabricated with Wire Heat No. 4P6052, Flux Type 0091, Flux Lot No. 0145. The copper and Nickel weight percents were taken from CE Reports NPSD-1039, Rev. 21'] & NPSD-1 119, Rev. 1M.

It should be noted that these Cu & Ni values do not Match RVID2 however, they would produce a more conservative Table Chemistry factor versus those in RVID2. This fact is negligible since the welds are not limiting.

(b) The Initial RTNtT values are measured values unless otherwise noted.

(c) Average of the two data points presented in Table A-3 of WCAP-101 10!s].

(d) Average of Lukens Mill Test Report and CE Test (Documented in WCAP-10110).

(e) Measured value documented in WCAP-10110

4 The chemistry factors were calculated using Regulatory Guide 1.99 Revision 2, Positions 1.1 and 2.1.

Position 1.1 uses the Tables from the Reg. Guide along with the best estimate copper and nickel weight percents. Position 2.1 uses the surveillance capsule data from all capsules withdrawn to date. The fluence values used to determine the CFs in Table 3 are the calculated fluence values at the surveillance capsule locations. Hence, the calculated fluence values were used for all cases. All calculated fluence values 91 (capsule and projections) for Seabrook Unit 1 were documented in DES-NFQA-98-011 . These fluences were calculated using the ENDF/B-VI scattering cross-section data set. The measured ARTNDT values for the weld data were adjusted for chemistry using the ratio procedure given in Position 2.1 of Regulatory Guide 1.99, Revision 2.

TABLE 2 Calculated Integrated Neutron Exposure of the Surveillance Capsules @ Seabrook Unit 1 Capsule Fluence Seabrook Unit 1(2)

U 2.55 x 10Ta nblr61,(E > 1.0 MeV)

Y 1.031 x 10'9 n/era2, (E > 1.0 MeV)

NOTES:

(a) Per Table 6-12 of DES-NFQA-98-01t91.

5 TABLE 3 Calculation of Chemistry Factors using Seabrook Unit 1 Surveillance Capsule Data Material Capsule Capsule fPa) FF() ARTDT(c) FF*ART=T FF2 Lower Shell Plate U 0.255 0.629 36.0 22.644 0.396 R-1808-3 (Long.) Y 1.031 1.009 44.0 44.396 1.018 Lower Shell U 0.255 0.629 28.0 17.612 0.396 PlateR-1808-3 Y 1.031 1.009 34.0 34.306 1.018 (Trans.) SUM: 118.958 2.828 CFB.2002.2 = X(FF

RTrT) + Z( FF 2) = (118.958) + (2.828) = 42.1*F Surveillance Weld U 0.255 0.629 11.4 (10.0) 7.171 0.396 Material(d) Y 1.031 1.009 11.4 (10.0) 11.503 1.018 SUM: 18.674 1.414 CF S,.r.Wed = 7(FF

RT-rr) + J( FV2) = (18.6741F) + (1.414) = 13.21F Notes:

(a) f= fluence. See Table 2, (x 10'9 n/cm2, E > 1.0 MeV).

(b) FF = fluence factor= fo28."

9 (c) ARTNT values are the measured 30 ft-lb shift values taken from DES-NFQA-98-011 1.

(cd) The surveillance weld metal ARTr values have been adjusted by a ratio factor 1.14.

The pre-adjusted values are in parenthesis.

6 TABLE 4 Summary of the Seabrook Unit I Reactor Vessel Beltline Material Chemistry Factors Material Reg. Guide 1.99, Rev. 2 Reg. Guide 1.99, Rev. 2 Position 1.1 CF's Position 2.1 CF's Intermediate Shell Plate R-1806-1 28.5 0 F - -

Intermediate Shell Plate R-1806-2 370F --

Intermediate Shell Plate R-1806-3 47.5 0 F --

Lower Shell Plate R-1808-1 370 F --

Lower Shell Plate R-1 808-2 37 0 F --

Lower Shell Plate R-1808-3 440 F 42.1 0F Intermediate & Lower Shell 30.7 0F 13.2 0 F Longitudinal Weld Seams (Heat # 4P6052)

Intermediate to Lower Shell 30.7 0F 13.2 0 F Girth Weld Seam (Heat # 4P6052)

Seabrook Unit 1 Surveillance Weld 270F (Heat # 4P6052)

7 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3.1 Overall Approach The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K1 ,, for the metal temperature at that time. Ki, is obtained from the reference fracture toughness curve, defined in Code Case N-641, "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System RequirementsSection XI, Division 1'43 841 of the ASME Appendix G to Section XI. The KI, curve is given by the following equation:

K10 = 33.2 +20.734* e'(OM(T-RTMT)J (1)

where, K10 = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This K10 curve is based on the lower bound of static critical K, values measured as a function of temperature on specimens of SA-533 Grade B Class 1, SA-508-1, SA-508-2, SA-508-3 steel.

3.2 Methodology for Pressure-Temperature Limit Curve Development The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C*Kim + Kit < Kir (2)

where, K = stress intensity factor caused by membrane (pressure) stress Kt = stress intensity factor caused by the thermal gradients Ki = function of temperature relative to the RTNr of the material C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical

8 For membrane tension, the corresponding KI for the postulated defect is:

Kim = Mmx(pRl/t) (3) where, M. for an inside surface flaw is given by:

M. = 1.85 for -" < 2, Mm = 0.926.H for 2 < 7! <3.464, Mm = 3.21 for J > 3.464 Similarly, Mm for an outside surface flaw is given by:

Mm = 1.77 for-f < 2, Mm = 0.893 f7 for 2*<7

3.464, Mm = 3.09 for ft > 3.464 and p = internal pressure, Ri = vessel inner radius, and t = vessel wall thickness.

For bending stress, the corresponding KI for the postulated defect is:

Krb = Mb

Maximum Stress, where Mb is two-thirds of Mm The maximum KI produced by radial thermal gradient for the postulated inside surface defect of G-2120 is Ki = 0.953xl0" 3 x CR x t25 , where CR is the cooldowv rate in *F/hr., or for a postulated outside surface defect, Ki, = 0.753x10"3 x HU x t2 5,where HU is the heatup rate in OF/hr.

The through-wall temperature difference associated with the maximum thermal K, can be determined from Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from Fig.

G-2214-2 for the maximum thermal K1 .

(a) The maximum thermal KI relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(1) and (2).

(b) Alternatively, the K, for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a 1/4-thickness inside surface defect using the relationship:

Kit = (1.0359C0 + 0.6322C, + 0.4753C2 + 0.3855C3) * (4)

9 or similarly, Krr during heatup for a 1/4-thickness outside surface defect using the relationship:

Krt = (1.043Co + 0.63 0C, + 0.48 1C2 + 0.40 1C3) * \ (5) where the coefficients Co, C1, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

o-(x) = Co+ Ci(x / a)+ C2(x / a)' + C3(x/ a)3 (6) and x is a variable that represents the radial distance from the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth.

Note, that equations 3, 4 and 5 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. Therefore, the P-T curve methodology is unchanged from that described in WCAP-14040, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves"121 Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above.

At any time during the heatup or cooldown transient, K10 is determined by the metal temperature at the tip of a postulated flaw at the 1l4T and 314T location, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIt, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference

flawofAppendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher value of KI, at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in KI, exceeds KIt, the calculated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the l/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

10 Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the KI for the l/4T crack during heatup is lower than the Kic for the l/4T crack during steady state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower KI, values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the l/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

3.3 Closure Head/Vessel Flange Requirements 10 CFR Part 50, Appendix G'10 I addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTmDT by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3106 psi), which is 621 psig for Seabrook Unit 1. The limiting unirradiated RTNDT of 30*F occurs in the vessel flange of the Seabrook Unit 1 reactor vessel, so the minimum allowable temperature of this region is 150°F at pressures greater than 621 psig. This limit is shown in Figures 1 through 4 wherever applicable.

3.4 LTOP System Allowable Pressure Per Code Case N-64113 3,the LTOP system shall limit the maximum pressure in the vessel to 100% of the pressure determined by Equation 2, herein, if KI, is used for determining allowable pressure, which is the case for this report.

11 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

ART = Initial RTNDT + ARTmr + Margin (7)

Initial RT r is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III oftheASME Boiler and Pressure Vessel Code!U]. If measured values of initial RTNT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.

ARTmT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

ARTNT = CF

fols-1 1-1Ogf) (8)

To calculate ARTNr at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.

S= f.*e C0 2 4x) (9) where x inches (vessel beltline thickness is 8.63 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 8 to calculate the ARTMT at the specific depth.

Duke Engineering & Services evaluated the vessel fluence projections in DES-NFQA-98-01191and are also presented in a condensed version in Table 5 of this report. The evaluation used the ENDF/B-VI scattering cross-section data set. This is consistent with methods presented in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves'4t 2 . Table 5 contains the calculated vessel surface fluences values at various azimuthal locations.

Tables 6 and 7 contain the l/4T and 3/4T calculated fluences and fluence factors, per the Regulatory Guide 1.99, Revision 2, used to calculate the ART values for all beltline materials in the Seabrook Unit 1 reactor vessel.

12 TABLE 5 Calculated Neutron Fluence Projectionsý') at Key Locations on the Reactor Vessel Clad/Base Metal Interface (10's n/cm2 , E > 1.0 MeV)

Azimuthal Location EFPY 00 190-21.50 290 31.50 44 0-45 0 5.572(') 0.201 0.355 0.196 0.197 0.369 16(c) 0.577 1.019 0.562 0.565 1.059 20(c) 0.722 1.273 0.702 0.706 1.324 32 1.155 2.037 1.123 1.130 2.119 Notes:

(a) Date of last capsule removal.

(b) Determined by multiplying Best Estimate Fluences by the Calculated-to-Measured Ratio (Ratio = 0.893).

(c) Values have been interpolated between 5.572 EFPY and 32 EFPY.

TABLE 6 Summary of the Vessel Surface, 1/4T and 3/4T Fluence Values used for the Generation of the 16 and 20 EFPY Heatup/Cooldown Curves Material 7 Surface /4 T() 3/4 T(2) 16 EFFY Beltline Materials() 1.059 x lol0 6.31 x 1018 2.24 x 10" 20 EFPY Beltline Materials0() 1.324 x 1019 7.89 x 101" 2.80 x 1018 Note (a) 114T and 3/4T = F(s,.&=) *e('° 24x) , where x is the depth into the vessel wall (i.e. 8.63*0.25 or 0.75)

(b) The beltline materials consist of the intermediate shell plates (R-1806-1,2,3), the lower shell plates (R-1808 1,2,3) intermediate and lower shell longitudinal welds and the intermediate to lower shell girth weld. Since the limiting material is a plate it would be subjected to the peak vessel fluence. Thus, for conservatism the peak vessel fluence was applied to the weld seams that are not at the peak azimuthal fluence location.

13 TABLE 7 Summary of the Calculated Fluence Factors used for the Generation of the 16 and 20 EFPY Heatup and Cooldown Curves Note:

(a) The beltline materials consist of the intermediate shell plates (R-1806-1,2,3), the lower shell plates (R-1808 1,2,3) intermediate and lower shell longitudinal welds and the intermediate to lower shell girth weld. Since the limiting material is a plate it would be subjected to the peak vessel fluence. Thus, for conservatism the peak vessel fluence was applied to the weld seams that are not at the peak azimuthal fluence location.

14 Margi is calculated as, M 2 +A~ . The standard deviation for the initial RTmT margin term, is cr 0°F when the initial RTNDT is a measured value, and 171F when a generic value is available. The standard deviation for the ARTNDT margin term, ca, is 17'F for plates or forgings, and 8.51F for plates or forgings when surveillance data is used. For welds, a, is equal to 28'F when surveillance capsule data is not used, and is 140F (half the value) when credible surveillance capsule data is used. a, need not exceed 0.5 times the mean value of ART=,T .

Contained in Tables 8 through 11 are the calculations of the 16 and 20 EFPYART values used for generation of the heatup and cooldown curves.

TABLE 8 Calculation of the ART Values for the 1/4T Location @ 16 EFFY Material RG 1.99 CF FF IRTDT (a) ARTMTb) Margin(c e) ART(d)

R2 Method (0F) (0F) (01) (CF) (19)

Intermediate Shell Plate Position 1.1 28.5 0.871 40 24.8 24.8 90 R-1806-1 Intermediate Shell Plate Position 1.1 37 0.871 0 32.2 32.2 64 R-1806-2 Intermediate Shell Plate Position 1.1 47.5 0.871 10 41.4 34 85 R-1806-3 Lower Shell Plate R-1808-1 Position 1.1 37 0.871 40 32.2 32.2 104 Lower Shell Plate R-1809-2 Position 1.1 37 0.871 10 32.2 32.2 74 Lower Shell Plate R-1808-3 Position 1.1 44 0.871 40 38.3 34 112 Position 2.1 42.1 0.871 40 36.7 17 94 Intermediate & Lower Shell Position 1.1 30.7 0.871 -60 26.7 26.7 -7 Longitudinal Weld Seams Position 2.1 13.2 0.871 -60 11.5 11.5 -37 (Heat # 4P6052)

Inter. to Lower Shell Girth Position 1.1 30.7 0.871 -60 26.7 26.7 -7 Weld Seam (Heat # Position 2.1 13.2 0.871 -60 11.5 11.5 -37 4P6052)

Notes:

(a) Initial RTNDT values are measured values except for the welds.

(b) ART-T = CF

FF (c) M = 2 *(ac2 + 'a2)"2 (d) ART = Initial RTr + ART,. + Margin (TF)

(e) All surveillance data is credible.

15 TABLE 9 Calculation of the ART Values for the 3/4T Location @ 16 EFPY Material RG 1.99 CF FF IRTNT(a) ARTNDP' Margin*" ARTW R2 Method (°T) (OF) (OF) (OF) (OF)

Intermediate Shell Plate Position 1.1 28.5 0.597 40 17 17 74 R-1806-1 Intermediate Shell Plate Position 1.1 37 0.597 0 22.1 22.1 44 R-1806-2 Intermediate Shell Plate Position 1.1 47.5 0.597 10 28.4 28.4 67 R-1806-3 Lower Shell Plate R-1808-1 Position 1.1 37 0.597 40 22.1 22.1 84 Lower Shell Plate R-1808-2 Position 1.1 37 0.597 10 22.1 22.1 54 Lower Shell Plate R-1808-3 Position 1.1 44 0.597 40 26.3 26.3 93 Position 2.1 42.1 0.597 40 25.1 17 82 Intermediate &Lower Shell Position 1.1 30.7 0.597 -60 18.3 18.3 -23 Longitudinal Weld Seams Position 2.1 13.2 0.597 -60 7.9 7.9 -44 (Heat # 4P6052)

Inter. to Lower Shell Girth Position 1.1 30.7 0.597 -60 18.3 18.3 -23 Weld Seam (Heat # 4P6052) Position 2.1 13.2 0.597 -60 7.9 7.9 -44 Notes:

(a) Initial RTr values are measured values except for the welds.

(b) ART,,= = CF

FF (c) M = 2 *("C + a' 2 )"2 (d) ART = Initial RTNDT + ARTm- + Margin (OF)

(e) All surveillance data is credible.

16 TABLE 10 Calculation of the ART Values for the 114T Location @ 20 EFPY Material RG 1.99 CF FF IRTNT(a) &RTNDTb) Margin(c,) ART(d) 0 R2 Method ( 0F) ( 1F) (OF) (OF) (OF)

Intermediate Shell Plate Position 1.1 28.5 0.934 40 26.8 26.8 94 R-1806-1 Intermediate Shell Plate Position 1.1 37 0.934 0 34.6 34 69 R-1806-2 Intermediate Shell Plate Position 1.1 47.5 0.934 10 44.4 34 88 R-1806-3 Lower Shell Plate R-1808-1 Position 1.1 37 0.934 40 34.6 34 109 Lower Shell Plate R-1808-2 Position 1.1 37 0.934 10 34.6 34 79 Lower Shell Plate R-1808-3 Position 1.1 44 0.934 40 41.1 34 115 Position 2.1 42.1 0.934 40 39.3 17 96 Intermediate & Lower Shell Position 1.1 30.7 0.934 -60 28.7 28.7 -3 Longitudinal Weld Seams Position 2.1 13.2 0.934 -60 12.3 12.3 -35 (Heat # 4P6052)

Inter. to Lower Shell Girth Position 1.1 30.7 0.934 -60 28.7 28.7 -3 Weld Seam (Heat # 4P6052) Position 2.1 13.2 0.934 -60 12.3 12.3 -35 Notes:

(a) Initial RTNm= values are measured values except for the welds.

(b) ARTIm = CF

FF (c) M = 2 *(2,2 + qj2)12 (d) ART = Initial RTN= + ARTNDT + Margin (0F)

(e) All surveillance data is credible.

17 TABLE 11 Calculation of the ART Values for the 3/4T Location @ 20 EFPY Material RG 1.99 CF FF ]IRTNr() ART T*) Margin(e,) ART d)

R2 Method (OF) (OF) (OF) (OF) (OF)

Intermediate Shell Plate Position 1.1 28.5 0.653 40 18.6 18.6 77 R-1806-1 Intermediate Shell Plate Position 1.1 37 0:653 0 24.2 24.2 48 R-1806-2 Intermediate Shell Plate Position 1.1 47.5 0.653 10 31.0 31.0 72 R-1806-3 Lower Shell Plate R-1808-1 Position 1.1 37 0.653 40 24.2 24.2 88 Lower Shell Plate R-1808-2 Position 1.1 37 0.653 10 24.2 24.2 58 Lower Shell Plate R-1808-3 Position 1.1 44 0.653 40 28.7 28.7 97 Position 2.1 42.1 0.653 40 27.5 17 85 Intermediate & Lower Shell Position 1.1 30.7 0.653 -60 20.0 20.0 -20 Longitudinal Weld Seams Position 2.1 13.2 0.653 -60 8.6 8.6 -43 (Heat # 4P6052)

Inter. to Lower Shell Girth Position 1.1 30.7 0.653 -60 20.0 20.0 -20 Weld Seam (Heat # 4P6052) Position 2.1 13.2 0.653 -60 8.6 8.6 -43 Notes:

(a) Initial RTwrT values are measured values except for the welds.

(b) ART*r = CF

FF (c) M =2 *(U,2 + CTA2)V2 (d) ART = Initial RTmT + ART=T + Margin (OF)

(e) All surveillance data is credible.

18 The lower shell plate R-1808-1 is the limiting beltline material for the 1/4T and 3/4T locations at both 16 and 20 EFPY. Contained in Table 12 is a summary of the limiting ARTs to be used in the generation of the Seabrook Unit 1 reactor vessel heatup and cooldown curves.

TABLE 12 Summary of the Limiting ART Values Used in the Generation of the Seabrook Unit 1 Heatup/Cooldown Curves 1/4T Limiting ART I 3/T Limiting ART 16 EFPY 104 184 20 EFPY 109 88

19 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods discussed in Sections 3.0 and 4.0 of this report This approved methodology is also presented in WCAP-14040-NP-A, Revision 2 with exception to those items discussed in Section 1 of this report.

Figures 1 and 3 present the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 80 and 100 0F/hr applicable for the first 16 and 20 EFPY, respectively. These curves were generated using the1996 ASME Code Section XI, Appendix G with the limiting ARTs. Figures 2 and 4 present the limiting cooldown curves without margins for possible instrumentation errors using cooldown rates of 0, 20, 40, 60, 80 and 100*F/hr applicable for 16 and 20 EFPY, respectively. Again, this curve was generated using the1996 ASME Code Section XI, Appendix G with the limiting ARTs. Allowable combination of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 1 through 4. This is in addition to other criteria which must be met before the reactor is made critical, as discussed below in the following paragraphs.

The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 1 and 3. The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in Code Case N-64113 (approved in February 1999) as follows:

1.5 KIm< Kir

where, Kim is the stress intensity factor covered by membrane (pressure) stress, KI- = 33.2 + 20.734 el° 07 Cr'- r)J, T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility temperature.

The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 10. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40*F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 3.0 of this report. For the heatup and cooldown curves without margins for instrumentation errors, the minimum temperature for the in service hydrostatic leak tests for the Seabrook Unit 1 reactor vessel at 16 and 20 EFPY is 164'F and 169°F, respectively. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40'F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.

Figures 1 through 4 define all of the above limits for ensuring prevention of nonductile failure for the Seabrook Unit 1 reactor vessel at 16 and 20 EFPY. The data points used for the heatup and cooldown pressure-temperature limit curves shown in Figures 1 through 4 are presented in Tables 13 through 16.

20 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATE R-1808-1 LIMrTINGART VALUES AT 16 EFPY: 1/4T, 104°F 3/4T, 84-F 2500 2250 2000 1750 Cn

0. 1500 In In 2 1250 CL 0

= 1000 U

C.

750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 1 Seabrook Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 80 &

100*F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology w/Kic

21 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATE R-1808-1 LIMITING ART VALUES AT 16 EFPY: 1/4T, 104-F 3/4T, 84°F 2500 2250 2000 1750 Q5.

1500 Un a

1250 L.)

1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2 Seabrook Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100 0F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology wlKic

22 TABLE 13 16 EFPY Heatup Curve Data Points Using 1996 App. G w/Kic (without Uncertainties for Instnrnentation Errors)

Heatup Curves (Run # 32218) 80 Heatup 80 Critical Limit 100 Heatup 100 Critical LimitI Leak Test Limit T P T P T P I T P T P 60 0 164 0 60 0 164 0 147 2000 22 60 621 164 620 60 621 164 620 164 2485 65 621 164 620 65 621 164 620 70 621 164 620 70 621 164 620 75 621 164 620 75 621 164 620 80 621 164 620 80 621 164 620 85 621 164 620 85 621 164 620 90 621 164 620 90 621 164 620 95 621 164 620 95 621 164 620 100 621 164 620 100 621 164 620 105 621 164 620 105 621 164 620 110 621 164 620 110 621 164 620 115 621 164 620 115 621 164 620 120 621 165 620 120 621 165 620 125 621 170 620 125 621 170 620 130 621 175 620 130 621 175 620 135 621 180 620 135 621 180 620 140 621 185 620 140 621 185 620 145 621 190 620 145 621 190 620 150 621 190 1170 150 621 190 1053 150 621 195 1235 150 621 195 1105 150 1170 200 1307 150 1053 200 1162 155 1235 205 1387 155 1105 205 1227 160 1307 210 1475 160 1162 210 1298 165 1387 215 1574 1A5 1227 215 1378 170 1475 220 1682 170 1298 220 1466 175 1574 225 1803 175 1378 225 1563 180 1682 230 1935 180 1466 230 1671 185 1803 235 2082 185 1563 235 1790 190 1935 240 2244 190 1671 240 1922 195 2082 245 2423 195 1790 245 2068 200 2244 200 1922 250 2228 205 2423 205 2068 255 2406 210 2228 215 2406

23 TABLE 14 16 EFPY Cooldown Curve Data Points Using 1996 App. G w/Kic (without Uncertainties for Instrumentation Errors)

Cooldown Curves (32218)

Cooldown Steady State 20*F/hr. Rate 4 0 'F/hr Rate 60°F/hr. Rate 80°F/hr. Rate 100°F/hr. Rate STP P T P T P T P T P 60 0 60 0 60 0 60 0 60 0 60 0 60 621 60 621 60 621 60 621 60 621 60 588 65 621 65 621 65 621 65 621 65 621 65 614 70 621 70 621 70 621 70 621 70 621 70 621 75 621 75 621 75 621 75 621 75 621 75 621 80 621 80 621 80 621 80 621 80 621 80 621 85 621 85 621 85 621 85 621 85 621 85 621 90 621 90 621 90 621 90 621 90 621 90 621 95 621 95 621 95 621 95 621 95 621 95 621 100 621 100 621 100 621 100 621 100 621 100 621 105 621 105 621 105 621 105 621 105 621 105 621 110 621 110 621 110 621 110 621 110 621 110 621 115 621 115 621 115 621 115 621 115 621 115 621 120 621 120 621 120 621 120 621 120 621 120 621 125 621 125 621 125 621 125 621 125 621 125 621 130 621 130 621 130 621 130 621 130 621 130 621 135 621 135 621 135 621 135 621 135 621 135 621 140 621 140 621 140 621 140 621 140 621 140 621 145 621 145 621 145 621 145 621 145 621 145 621 150 621 150 621 150 621 150 621 150 621 150 621 150 621 150 621 150 621 150 621 150 621 150 621 150 1560 150 1560 150 1560 150 1560 150 1560 150 1560 155 1660 155 1660 155 1660 155 1660 155 1660 155 1660 160 1771 160 1771 160 1771 160 1771 160 1771 160 1771 165 1893 165 1893 165 1893 165 1893 165 1893 165 1893 170 2028 170 2028 170 2028 170 2028 170 2028 170 2028 175 2178 175 2178 175 2178 175 2178 175 2178 175 2178 180 2343 180 2343 180 2343 180 2343 180 2343 180 2343

24 MATERIAL PROPERTY BASIS LIMITING MATERJAL: LOWER SHELL PLATE R-1808-1 LIMITING ART VALUES AT 20 EFPY: 1/4T, 109-F 3/4T, 88°F 2500 Operdim Verslon:5 I Run:17535 2250 2000 Utnaccep'table[ cpalOperation

[ Acceptable iOperation 1750 co 1500

[L Heatup Rate Crtia Limi i.5 1250 100 Deg. FIHrI l/"*Deg._F/Hr 10 I.

0 1000 Ca 750 500 Criticality Limit based on Boltup lInservice hydrostatic test 250 NTemp temperature (169 F) for the service period up to 20 EFPY 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 3 Seabrook Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of 80 &

100F/ihr) Applicable for the First 20 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology w/Kic

25 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATE R-1808-1 LIMTING ART VALUES AT 20 EFPY: 1/4T, 109WF 314T, 88-F 2500 Operlim Version.5 I Run'17535 2250 Unacceptable 2000 Operation

-1750 0~

w 1500 0

IL.

IL 1250 0 1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 4 Seabrook Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100'F/hr) Applicable for the First 20 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology w/Kic

26 TABLE 15 20 EFPY Heatup Curve Data Points Using 1996 App. G w/Kic (without Uncertainties for Instrumentation Errors)

Heatuv Curves (Run # 17535) 80 T HeatupP 80 Critical T Limit P 100 T HeatupP 1100 Critical T Limit P2 Leak T Test Limit p

60 0 169 0 60 0 169 0 152 2000 60 621 169 620 60 621 169 620 169 2485 65 621 169 620 65 621 169 620 70 621 169 620 70 621 169 620 75 621 169 620 75 621 169 620 80 621 169 620 80 621 169 620 85 621 169 620 85 621 169 620 90 621 169 620 90 621 169 620 95 621 169 620 95 621 169 620 100 621 169 620 100 621 169 620 105 621 169 620 105 621 169 620 110 621 169 620 110 621 169 620 115 621 169 620 115 621 169 620 120 621 169 620 120 621 169 620 125 621 170 620 125 621 170 620 130 621 175 620 130 621 175 620 135 621 180 620 135 621 180 620 140 621 185 620 140 621 185 620 145 621 190 620 145 621 190 620 150 621 190 1116 150 621 190 1006 150 621 195 1176 150 621 195 1054 150 1116 200 1242 150 1006 200 1107 155 1176 205 1316 155 1054 205 1166 160 1242 210 1398 160 1107 210 1232 165 1316 215 1488 165 1166 215 1305 170 1398 220 1588 170 1232 220 1386 175 1488 225 1699 175 1305 225 1476 180 1588 230 1822 180 1386 230 1575 185 1699 235 1957 185 1476 235 1685 190 1822 240 2107 190 1575 240 1807 195 1957 245 2272 195 1685 245 1941 200 2107 250 2454 200 1807 250 2089 205 2272 205 1941 255 2253 210 2454 210 2089 260 2433 215 2253 220 2433

Curves 27 TABLE 16 20 EFPY Cooldown Curve Data Points Using 1996 App. G w/Kic (without Uncertainties for Instrumentation Errors)

Cooldown Curves (17535)

Cooldown Steady State 20°F/hr. Rate 40*F/hr Rate 60IF/hr. Rate 80°F/hr. Rate 100°F/hr. Rate T P T P ITITT P P I T I T P 60 0 60 0 60 0 60 0 60 0 60 0 60 621 60 621 60 621 60 621 60 620 60 564 65 621 65 621 65 621 65 621 65 621 65 587 70 621 70 621 70 621 70 621 70 621 70 614 75 621 75 621 75 621 75 621 75 621 75 621 80 621 80 621 80 621 80 621 80 621 80 621 85 621 85 621 85 621 85 621 85 621 85 621 90 621 90 621 90 621 90 621 90 621 90 621 95 621 95 621 95 621 95 621 95 621 95 621 100 621 100 621 100 621 100 621 100 621 100 621 105 621 105 621 105 621 105 621 105 621 105 621 110 621 110 621 110 621 110 621 110 621 110 621 115 621 115 621 115 621 115 621 115 621 115 621 120 621 120 621 120 621 120 621 120 621 120 621 125 621 125 621 125 621 125 621 125 621 125 621 130 621 130 621 130 621 130 621 130 621 130 621 135 621 135 621 135 621 135 621 135 621 135 621 140 621 140 621 140 621 140 621 140 621 140 621 145 621 145 621 145 621 145 621 145 621 145 621 150 621 150 621 150 621 150 621 150 621 150 621 150 621 150 621 150 621 150 621 150 621 150 621 150 1469 150 1469 150 1469 150 1469 150 1469 150 1469 155 1560 155 1560 155 1560 155 1560 155 1560 155 1560 160 1660 160 1660 160 1660 160 1660 160 1660 160 1660 165 1771 165 1771 165 1771 165 1771 165 1771 165 1771 170 1893 170 1893 170 1893 170 1893 170 1893 170 1893 175 2028 175 2028 175 2028 175 2028 175 2028 175 2028 180 2178 180 2178 180 2178 180 2178 180 2178 180 2178 185 2343 185 2343 185 2343 185 2343 185 2343 185 2343

28 6 REFERENCES

1. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlenent of Reactor Vessel Materials," U.S.

Nuclear Regulatory Commission, May 1988.

2. WCAP-14040-NP-A, Revision 2, '"Methodology used to Develop Cold Overpressure Mitigating system Setpoints and RCS Heatup and Cooldown Limit Curves", J.D. Andrachek, et. al., January 1996.

3. ASME Code Case N-641, "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System RequirementsSection XI, Division 1", January 17, 2000.

[Sub

Reference:

ASME Code Case N-640, "AlternativeReference FractureToughnessfor Development of P-TLimit Curvesfor Section A7, Division 1",February26, 1999.]

4.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix (; "Fracture Toughness Criteria for Protection Against Failure." Dated December 1995, through 1996 Addendum.

5. "Fracture Toughness Requirements", Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.

6. CE Report NPSD-1039, Revision 2, "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds", CEOG Task 902, By the CE Owners Group. June 1997.

7. CE Report NPSD-1119, Revision 1, "Updated Analysis for Combustion Engineering Fabricated Reactor Vessel Welds Best Estimate Copper and Nickel Content", CEOG Task 1054, By the CE Owners Group. July 1998.

8. WCAP-101 10, "Public Service Company of New Hampshire Seabrook Station Unit No. 1 Reactor Vessel Radiation Surveillance Preogramn", L.R. Singer, March 1983.

9. DES-NFQA-98-01, "Analysis of Seabrook Station Unit 1 Reactor Vessel Surveillance Capsules U and Y", E.C. Biemiller, et. al., May 1998.

10. Code of Federal Regulations, 10 CFR Part 50, Appendix (; "Fracture Toughness Requirements,"

U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

11. 1989 Section II, Division 1 of the ASME Boiler and Pressure Vessel Code, Paragraph NB-233 1, "Material for Vessels."

ENCLOSURE 3 TO NYN-02093 NFSB 02-0061 Seabrook Station Cold Overpressure Mitigating System (COMS)

Setpoint Development Methodology P. J. Guimond August 2002 Prepared by Framatome ANP DE&S for North Atlantic Energy Services Corporation A

FRAMATOME ANP Solomon Pond Park 400 Donald Lynch Boulevard FRAMATOME ANP DE&S Marlborough, Massachusetts 01752

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS Dat z Prepared:

P. I. Guimond, Engineering Consultant Date Nuclear Analysis & Fuel Management Services Reviewed: 4?J421 Cf A. E. Ladieu, Engineering Consultant Date Nuclear Analysis & Fuel Management Services Approved:

P. A. Bergeron, Eanager Date Seabrook Project ii

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS DISCLAIMER OF RESPONSIBILITY Framatome ANP DE&S prepared this document in cooperation with North Atlantic Energy Services Corporation (NAESCO). The use of information contained in this document by anyone other than Framatome ANP DE&S, or NAESCO for whom this document was prepared under contract, is not authorized and, with respect to any unauthorized use, neither Framatome ANP DE&S nor its officers, directors, agents, or employees assume any obligation, responsibility, or liability or make any warranty or representation as to the accuracy or completeness of the material contained in this document.

iii

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS ABSTRACT This report documents the considerations and selection procedures used to determine setpoints for the Cold Overpressure Mitigating System (COMS) at Seabrook Station. COMS limits reactor vessel pressure during cold overpressurization events to values below the low temperature overpressurization (LTOP) limits. The LTOP limits are in part based on the reactor vessel heatup and cooldown pressure/temperature limits specified by 10 CFR 50 Appendix G. The Appendix G limits shift as reactor vessel neutron Irradiation increases over time. Appendix G limits and COMS setpoints are derived to conservatively bound the limits through a specified vessel total exposure. Prior to exceeding the applicable exposure, the Appendix G limits and COMS setpoints must be modified to conservatively bound a higher total exposure. This report describes the application of procedures and associated design-basis event evaluations to support modifications to the COMS setpoints for Seabrook Station.

iv

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS TABLE OF CONTENTS Pagte APPROVALS ................................................................................................................................... ii DISCLAIMER OF RESPONSIBILITY ................................................................................................. iii ABSTRACT ................................................................................................................................... iv TABLE OF CONTENTS ................................................................................................................... v LIST OF TABLES ........................................................................................................................... vi LIST OF FIGURES ......................................................................................................................... vi

1.0 INTRODUCTION

AND

SUMMARY

....................................................................................... 7 1.1 Introduction .................................................................................................... 7 1.2 Summary ................................................................................................................ 7 2.0 SETPOINT SELECTION CONSIDERATIONS ...................................................................... 9 2.1 Pressure Limit Selection ........................................................................................ 9 2.2 Design Basis Event Evaluations ............................................................................. 9 2.3 COMS Setpoint Determination ............................................................................. 13 2.4 Application of Setpoint Uncertainties .................................................................. 17 2.5 COMS Arming Temperature ................................................................................ 18

3.0 REFERENCES

.................................................................................................................... 19 V

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS LIST OF TABLES Table No. Title Paae 2-1 LTOP Pressure Limit for 20 EFPY ....................................................................... 10 2-2 Derivation of the Maximum Allowable PORV COMS Setpoint for 20 EFPY .......... 16 2-3 Comparison of Maximum PORV Setpolnt Equations to Maximum Allowable Setpoint ............................................................................................................... 17 LIST OF FIGURES Figure No. "Title Page 1-1 RCS Cold Overpressure Protection Setpoints ...................................................... 8 2-1 Seabrook PRESS Model Nodalization ................................................................. 12 vi

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS

1.0 INTRODUCTION

AND

SUMMARY

1.1 Introduction One alternative for providing low temperature overpressure protection allowed by Seabrook's Technical Specifications is the pressurizer pressure operated relief valves (PORVs) operating with reduced pressure setpoints. The reduced pressure setpoints are a function of the Indicated reactor coolant system temperature. The PORVs and reduced pressure setpoint functions are referred to as the Cold Overpressure Mitigating System (COMS). This report documents the procedure used to develop the temperature dependent pressure setpoints for the COMS at Seabrook Station. The topics herein addressed are:

"* Pressure limit selection

"* Design basis event evaluations

"* COMS setpoint selection

"* Application of setpoint uncertainties

"* COMS arming temperature 1.2 Summary The reactor vessel heatup and cooldown limitations for Seabrook have been updated to cover an increased total exposure of 20 EFPY [1]. Revised pressurizer PORV COMS setpoints have therefore been derived for operation of Seabrook Station's reactor vessel to a cumulative exposure of 20 EFPY (see Figure 1-1). The considerations Included in the derivation of these setpoints are described in section 2.0 of this report. The revised setpoints provide overpressure protection for the Seabrook reactor vessel and closure head/flange consistent with the requirements of ASME Section XI Division I Appendix G [6], ASME Code Case N-641 [2], and 10 CFR Part 50 Appendix G [3]. Implementation of these revised maximum allowable COMS setpoints requires USNRC approval of an exemption from 10 CFR Part 50 Appendix G LTOP pressure limit requirements for application of ASME Code case N-641.

7

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS Figure 1-1 RCS COLD OVERPRESSURE PROTECTION SETPOINTS Valid for first 20 EFPY, setpoint contains margin of 500 F for transient effects 2500 I a 2250

. . .. ... . .. ...... - .1.7......

2000 L-'1750 I

.. ............. . ..... , . .. . ... o . . . .° .°.. . ,o. .. .° ° ,............ . .. °. * ............ ... . .,.

-1500 U 1250 0

C 1000

. °.. ...... ......... .... ° , . .-,......... . . . . .. . . .........

-.oo ... . .. . ......... . . . . . .°. . .. ° . . . .

S750 500

.. ..... ... ........ ..... *... ... .... . .... ...... ......... . ,°°.,°... ... 6,,

......... ,°. ...... . ... .... . . ..... ... °...... ........... ,

250 0

so 100 150 200 250 300 350 RCS TEMPERATURE (deg. F)

T

200.0, P = 561.0 ; (200.0 <T T; 218.65), P = 18.2"T"- 2221.0 ; T > 218.65, P = 33.0*'T - 5457.0

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS 2.0 SETPOINT SELECTION CONSIDERATIONS 2.1 Pressure Limit Selection New Seabrook Unit 1 heat up and cool down allowable pressure-temperature limit curves for normal operation [1] were derived using the KI, methodology allowed by ASME Code Case N 641[23. Per Code Case N-641, to provide protection against failure during reactor startup and shutdown operation due to overpressure events that have been classified Service Level A or B, low temperature overpessure protection (LTOP) systems shall limit the maximum pressure In the vessel to 100% of the pressure determined to satisfy the equation, 2Kn, +Kut < Krc. Credit is taken for the fact that experience shows that LTOP events are most likely to occur at isothermal conditions. The allowable pressure limit for the COMS is therefore taken to be the steady-state (isothermal) pressure limit curve (see Table 2-1). Below 150'F, to comply with the closure head/vessel flange region limitation on system pressure imposed by 10 CFR Part 50, Appendix G

[3] the allowable pressure Is limited to a maximum of 621 psig.

2.2 Design Basis Event Evaluations Two design basis overpressure events are considered in the COMS setpoint development: a heatup transient and a mass addition event.

The COMS design basis heatup transient starts from an initial condition where the RCS flow rate is at zero, the steam generators are at a temperature 500 F hotter than the rest of the RCS (limited by Technical Specification 3.4.1.4.1), the pressurizer is water solid (a bubble has not yet been formed) and then a RCP is inadvertently started. The starting of the RCP initiates heat transfer from the secondary to the primary side which causes a pressurization of the primary side. The pressure response to the heatup transient is analyzed using a modified version of the PRESS computer code [4). The modifications to the code were made to provide increased flexibility In the modeling of relief valve opening/closing dynamics (delay time, non-linear opening/closing characteristics), control input location (static and dynamic spatial pressure differences), and type of relief valve (PORV vs RHR safety valve). The basic pressure and heat transfer solution techniques were not modified.

9

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS TABLE 2-1 LTOP Pressure Umit for 20 EFPY

~r mrt~~ 1taVSaeesre-.Umlt' RFiafge71Jmit7, T0-Pes06Ul1?

60 0 0 0 60 564' 621 621 65 587' 621 621 70 614' 621 621 75 621 621 621 80 621 621 621 85 621 621 621 90 621 621 621 95 621 621 621 100 621 621 621 105 621 621 621 110 621 621 621 115 621 621 621 120 621 621 621 125 621 621 621 130 621 621 621 135 621 621 621 140 621 621 621 145 621 621 621 150 621 - 621 150 1469 1469 155 1560 1560 160 1660 - 1660 165 1771 - 1771 170 1893 - 1893 175 2028 - 2028 180 2178 - 2178 185 2343 - 2343 190 2508 (extrapolated)' - 2508 195 2673 (extrapolated)' 2673 1Since the slope of limit Increases with temperature, linear extrapolation Is conservative.

10

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS The PRESS computer code models the reactor coolant system, including the pressurizer, the primary and secondary sides of the steam generator tube bundle region, and the steam generator tube walls. The Seabrook PRESS model nodalization is shown in Figure 2-1.

To determine the system pressure transient following the initiation of flow from the reactor coolant pump, PRESS dynamically calculates the primary/secondary heat transfer through the steam generator tube walls, the fluid energy transfer through the primary system via coolant flow, and the discharge rate through the relief valve. For the Seabrook analysis, the loop flows were assumed to Increase linearly from zero to steady-state conditions In 15 seconds.

In addition to simulating the primary/secondary heat transfer, the PRESS model simulates the dynamic opening and closing characteristics of the PORV relief valves. The spatial and dynamic pressure drops affecting the wide range pressure Input to the COMS PORV control system and the PORV inlet pressure are also accounted for In the modified PRESS model.

The PORV relief flow as a function of valve Inlet pressure was obtained from an independent calculation of PORV performance for a range of subcooled and saturated liquid relief conditions using the RETRAN-02 computer code [5]. The RETRAN-02 analysis used the iso-enthalpic expansion (IHE) flow model to conservatively minimize the predicted relief flow.

PRESS simulations of the heatup event were used to determine the peak RCS pressure setpolnt overshoot defined as the peak pressure In the RV downcomer adjacent to the beitline minus the nominal PORV setpoint. Calculations were performed over a range of assumed initial RCS water temperatures below the expected COMS arming temperature.

The second design basis event Is the mass addition event. The event considered is mass addition from all potentially operating charging pumps via both the normal charging flow path, with the charging flow (FCV-121) and head (HCV-182) control valves fully open, and the charging pump safety Injection flow path In parallel, due to an Inadvertent opening of one of the two SI flow path block valves. Letdown flow was assumed to be isolated. Current plant Technical Specifications limit the number of operable charging pumps to a single CCP in Modes 4 and 5 except for a brief transition period when entering or exiting Mode 3 and/or when swapping pumps. Therefore the analysis assumed flow from a single CCP.

11

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS FIGURE 2-1 SEABROOK PRESS MODEL NODALIZATION SrelieF flow SG 2 Tubes So I Tubes SG 2 Secondc-v SO I SeconncrU Loop 2 Cold Leg Loop I Cold Leg Core - 5 nodes Coldaiot Loops - 5 nodes SO - 1 nodes Loop 1 models the single loop where RCP is started, Loop2 models the three loops with Inactive RCPs.

12

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS The mass addition event was analyzed using a RETRAN-02 model of the Seabrook reactor coolant system to determine the pressure response and maximum RCS pressure setpoint overshoot.

Maximum setpoint overshoot was determined as a function of Initial wide-range pressure and PORV setpoint for a cold (100 0 F) reactor coolant condition. Assuming a cold RCS conservatively minimizes the compressibility of the RCS water volume, maximizing the rate of pressure Increase and setpoint pressure overshoot.

2.3 COMS Setpoint Determination The pressure overshoot values calculated for the design basis heat and mass addition events are used to determine the maximum allowable PORV setpoint required to prevent violation of the LTOP allowable pressure limit during the event. The maximum PORV setpoint must be a value lower than the selected LTOP allowable pressure limit minus the overshoot predicted for the applicable temperature condition (determination of the applicable temperature for each event is discussed below).

In the Seabrook COMS design the auctioneered low wide-range hot leg temperature is used as input to the setpoint function generator for one PORV and the auctioneered low wide-range cold leg temperature is used as Input to the setpoint function generator for the second PORV.

During the design basis heatup event, the temperature In the cold legs may Increase by as much as 50 0 F due to heat addition to the loop flow from the hotter steam generator secondary water once the RCP is started. This heated water eventually reaches all four hot legs and thus the setpoint inputs for auctioneered low cold and hot leg temperature are expected to increase during the event. The temperature increase is limited to the temperature of the steam generator secondary water, so the maximum temperature Increase Is limited to 50'F. Since the PORV setpoints are a programmed function of the input temperature, the PORV setpoint may increase during the event due to the Increase in Indicated RCS temperatures. To account for this possibility, the COMS setpoints are chosen such that the maximum PORV setpolnt at a temperature up to 500 F higher than the initial RCS temperature is low enough to preclude violation of the LTOP pressure limit considering the maximum pressure overshoot for the event.

The setpoint at the initial RCS temperature plus 50'F must be less than or equal to the LTOP pressure limit at the initial temperature minus the maximum pressure overshoot.

13

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS During a mass addition event, the injection of cold water downstream of the location in the cold leg where RCS wide-range temperature is measured could result In lower fluid temperatures adjacent to the RV wall in the downcomer region than the temperature Indicated by the cold leg temperature setpoint input. Analysis of the mixed stream fluid temperature entering the reactor vessel showed that the maximum temperature difference is less than 50 0 F for the maximum Injection flow with RHR flow rate at a conservatively low minimum value. Therefore, to prevent pressure from exceeding the LTOP allowable pressure limit applicable to the potentially lower temperature at the vessel beltline, the maximum COMS PORV setpoint at any Indicated wide range temperature must be less than or equal to the LTOP pressure limit corresponding to a temperature 500 F lower, minus the maximum pressure overshoot for a mass addition event at any temperature over the 50'F temperature span. This is the same condition determined to be required for the heat addition event.

The heat and mass addition event setpoint overshoot values are also used to indicate the amount of stagger that should be maintained between the two PORV setpoints. The COMS system consists of two Independently actuated PORVs. In order to minimize the potential undershoot of the setpolnt (which could cause RCS pressure to challenge minimum RCP seal operating pressure limits) It is desirable for only a single valve to be actuated as the result of an overpressure event.

Since a single valve has sufficient relief capacity to limit the RCS pressure to the setpoint plus overshoot, setting one valve to a setpoint less than the setpolnt of the other valve minus the overshoot Ideally eliminates the challenge to the other valve.

A second factor Influencing the selection of the lower PORV setpoint is the effect of pressurizer water temperature on PORV relief volumetric relief rate. The PORV volumetric relief rate was found to decrease as the enthalpy of the pressurizer liquid approached the saturation temperature corresponding to the PORV setpoint. As a result, the pressure overshoot for a mass addition event was found to Increase substantially when the temperature of the fluid In the pressurizer exceeds the value which would result in the PORV volumetric relief capacity at the setpoint pressure dropping below the mass addition rate from the CCP at the corresponding RCS pressure. Therefore, in order to minimize the overshoot, the maximum Initial pressurizer liquid enthalpy/temperature is limited to a value which assures that the fluid relieved at the PORV setpoint Is adequately subcooled, resulting in a relief valve flow at the maximum allowed setpoint pressure greater than or equal to the mass addition rate at that pressure. The mechanism used to implement this limit is to set or assure that the nominal COMS setpoint of the PORV with the lower setpoint limits pressurizer water temperature to the required maximum value. The setpoint 14

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS of the PORV with the minimum COMS setpoint is Indicated In station operating procedures for heatup and cooldown. This assures that the pressurizer water is adequately subcooled when the maximum setpoint pressure is reached, minimizing the pressure overshoot.

Within these constraints, the PORV setpoints are set as high as possible in order to maximize the margin avaliable between the PORV setpoint and the RCP seal minimum operating pressure limit (325 psig). Maximizing this margin minimizes the chance of damaging a seal due to pressure undershoot while the relief valve Is closing. Operation of the RCP at RCS pressures lower than the limit could result In damage to the seal. However, preventing damage to the seal is an economic concern and not a safety limit. Preventing non-ductile failure of the vessel Is the primary safety concern. Therefore, if necessary to prevent violating the LTOP allowable pressure limit, the COMS setpoInt Is set below the value which would challenge the RCP seal limit, since the seal limit is of secondary significance.

Table 2-2 provides an example of the determination of the maximum COMS PORV setpoint for the new LTOP pressure limit for 20 EFPY for Seabrook. The maximum PORV setpolnt, P, specified in TS Figure 3.4-4 as a function of RCS temperature, T, is expressed as a set of linear equations defining a lower bounding fit of the last two columns in Table 2-2:

T 5 200.00 F, P = 561.0 psig; (200.0 0F < T < 218.650 F), P = 18.2"T - 2221.0 psig; T > 218.650F, P = 33.0"T - 5457.0 psig The equations conservatively fit the allowable setpoint limit as shown In Table 2-3.

15

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS TABLE 2-2 16

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS TABLE 2-3 Comparison of Maximum PORV Setpoint Equations to Maximum Allowable Setpoint Ak~ii{k- p-SR.CS -,Te~rnpbigatu-re"F,,radm mA loa]Stdrt sg*Z, 1i!ei~m ~

60-200 561 561 200 1419 1419 205 1510 1510 210 1610 1601 215 1721 1692 220 1848 1803 225 1983 1968 230 2133 2133 235 2298 2298 240 2463 2463 245 2628 2628 2.4 Application of Setpoint Uncertainties The wide range temperature and pressure inputs to the COMS PORV setpoint function generators and the function generator and trip bistable hardware are subject to various uncertainty factors which influence setpoint accuracy, including sensor/instrument accuracy, calibration tolerance, drift, and environmental conditions. The uncertainties on the PORV setpoint are conservatively quantified and statistically combined in the determination of nominal "in-plant" setpoint equations for each PORV. The equations for the lower set PORV are also adjusted downward by the maximum overshoot to provide the desired stagger to minimize the potential for both valves opening in response to an event.

17

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS 2.5 COMS Arming Temperature The COMS system design has an arming bi-stable for each PORV. When wide-range RCS temperature goes below the selected arming temperature, the bi-stable arms COMS by causing the associated PORV block valve to open (if it is closed) and enabling the PORV COMS setpoint.

ASME Code Case N-641 indicates the LTOP System Effective Temperature Is the "temperature at or above which the safety relief valves provide adequate protection against nonductile failure".

Per the code case, LTOP systems shall be effective below the higher of an Inlet coolant temperature of 200OF or a coolant temperature corresponding to a reactor vessel 1/4T metal temperature of RT,1 T-+ 40"F for Inside axial surface flaws and RTADoT - 850 F for inside circumferential surface flaws for all vessel beltline materials. The adjusted 1/4T RTNVDT for the limiting beltline material at 20 EFPY is 109'F. Therefore, the code case requires the COMS system to be effective below an Inlet coolant temperature of 200"F.

Examination of Table 2-2 indicates that the maximum COMS PORV setpoint at 200OF is lower than the safety relief valve setpoint (2485 psig -3%,per TS 3.4.2.1). Therefore, the COMS arming temperature is selected to be greater than or equal to the temperature at which the maximum COMS setpoint pressure limit is ;* safety relief valve maximum opening setpoint plus an allowance for the difference in pressure at the safety valve and the Indicated wide range pressure which actuates the PORV in the COMS mode. This temperature is then adjusted to a higher value to include allowances for applicable COMS setpoint uncertainties.

18

NFSB 02-0061 SEABROOK STATION COLD OVERPRESSURE MITIGATING SYSTEM (COMS) SETPOINTS

3.0 REFERENCES

[1] Laubham, T. J., WCAP-15745, Seabrook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation Westinghouse Non-Proprietary Class 3, December 2001.

[2] ASME Code Case N-641, " Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System RequirementsSection XI, Division 1",

January 17, 2000.

[3] Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," U. S. Nuclear Regulatory Commission, Washington, D. C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

[4] Chapman, J. R., YAEC-1124, PRESS An Analytical Model Used In PWR Overpressurization Analysis. February 1977.

[5] EPRI NP-1850-CCM-A, Computer Code Manual, RETRAN A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, June 1987.

19

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