ML20247F955

Safety Evaluation Supporting Amends 135 & 127 to Licenses DPR-42 & DPR-60,respectively

ML20247F955
Person / Time
Site:	Prairie Island
Issue date:	05/04/1998
From:	NRC (Affiliation Not Assigned)
To:
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ML20247F921	List: ML20247F916 ML20247F955
References
NUDOCS 9805200100
Download: ML20247F955 (5)

Text

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4 UNITED STATES f

,j NUCLEAR REGULATORY COMMISSION g

'f WASHINGTON, D.c. 30666 4 001 e,,,,*

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 135 TO FACILITY OPERATING LICENSE NO. DPR-42 AND AMENDMENT NO. 127 TO FACILITY OPERATION LICENSE NO. DPR-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306

1.0 INTRODUCTION

By application dated March 6,1998, as supplemented March 30,31, and April 13,1998, the Northern States Power Company (NSP or the licensee) requested amendments to the Technical Specifications (TS) appended to Facility Operating License No. DPR-42 for the Prairie Island Nuclear Generating Plant, Unit 1, and Facility Operating License No. DPR-60 for the Prairie Island Nuclear Generating Plant, Unit 2. The proposed amendments would update the TS heatup and cooldown rate curves and extend their reactor vessel fluence limit from the current 20 effective full power years (EFPYs) to a new value of 35 EFPYs, incorporate into TS the use of a Pressure and Temperature L!mits Report (PTLR), and change the power-operated l

relief valves (PORVs) temperature requirement for operability.

j NSP supplemented the March 6,1998, submittal by letters dated March 30, 31, and April 13, 1998. The supplemental submittals provided additional clarifying information within the scope of the original FederalRegister notice and did not change the staff's initial proposed no significant hazards consideration determination.

2.0 BACKGROUND

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

l 10 CFR 50.36 identifies four criteria to be used in determining whether a particular matter is required to be included in the TS, as follows: (1) installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant 9805200100 980504 PDR ADOCK 05000282 P

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pressure boundary; (2) a process variable, design feature, or operating restriction that is an

' initial condition of a design-basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; (3) a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design-basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; (4) a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. As a result, existing TS requirements that fall within or satisfy any of these criteria must be retained in the TS, while those TS requirements that do not fall within or satisfy these criteria may be relocated to other licensee-controlled documents.

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3.0 EVALUATION All components of the reactor coolant system (RCS) are designed to withstand the effects of cyclic loads resulting from system pressure and temperature changes. These loads are introduced by heatup and cooldown operations, power transients, and reactor trips. In

. accordance with Appendix G to 10 CFR Part 50, TS limit the pressure and temperature changes during RCS heatup and cooldown within the design assumptions and the stress limits for cyclic operation. These limits are defined by the pressure-temperature (P/T) limit curves for heatup and cooldown. Each curve defines an acceptable region for normal operation.- The curves are used for operational guidance during heatup and cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The licensee submitted its revised vessel material surveillance reports in a letter dated March 6, 1998. In these reports the licensee documents its evaluation of recently removed vessel I

material capsule specimens from Prairie Island Units 1 and 2 reactor vessels. These specimens were pulled from both Units 1 and 2 during recent outages and analyzed to determine how much radioactive fluence they have received over the Units 1 and 2 reactor core lives thus far. The licensee then performed a metallurgical analysis which provided a best estimate of the degree of radiation induced embrittlement that will occur in the Prairie Island Units 1 and 2 reactor vessels through 35 EFPYs. The licensee's best estimate results indicated that the Prairie Island vessels will have sufficient protection against the occurrence of pressurized thermal shock (PTS), and will meet the fracture toughness requirements specifed in 10 CFR Part 50 Appendix G. The new heatup and cooldown curves which dictate operationallimits were developed based on the results of these analyses. The staff finds it acceptable for the licensee to extend the period of applicability of the heatup and cooldown curves from 20 to 35 EFPYs, based on the staff's review of the licensee's analyses.

The licensee used the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) Case N-514, and requested an exemption from 10 CFR 50.60, i

" Acceptance criteria for fracture prevention measures for light water nuclear power reactors for normal operation " The licensee demonstrated that the exemption was warranted under 10 CFR 50.12(a)(2)(ii). The exemption to permit use of Code Case N-514 was granted on

April 30,1998, for Prairie Island Units 1 and 2.

The temperature dependency for PORV operability has been changed from 310 'F to 350 *F.

. The historical value of 310 'F was used in place of the lower calculated value of 243 *F, which results from the summation of the analytical limit of 225 'F and the indicating instrument i

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_ _ _ _ - _ _ _ _ _ _ _ _ _ _ - r'. nnel uncertainty of 18 'F. Analysis has shown that an overpressure protection system (OPPS) enable temperature of 310 'F with the constant value OPPS pressure relief setpoint of 500 psig will protect the 10 CFR Part 50 Appendix G tirittle fracture P/T limits as modified by the ASME Code Case N-514 from the minimum unvented RCS temperature through rated operating temperature. Changing the temperature from 310 'F to 350 'F will be consistent with the temperature in NUREG-1431, " Standard Technical Specifications, Westinghouse Plants."

This provides for a 40 *F temperature band to permit lining up the PORVs from operation with the OPPS enabled (RCS<310 'F) to operation in the normal pressure relief mode (RCS>350 'F).

The licensee-proposed changes to the TS are in accordance with the guidance in Generic Letter (GL) 96-03 " Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," as follows:

1.

TS 1.0 The definition contained in GL 96-03 for the term Pressure and Temperature Limits Report (PTLR) is added with two exceptions. The PTLR is not unit specific for Prairie Island, and the Prairie Island specific terminology, Over Pressure Protection System (OPPS) is used in place of the generic letter terminology, low temperature overpressure l

protection system (LTOP).

2.

TS 3.1. A.1.c(4). TS 3.1.A.2.c(2) TS 3.3.A.3. and Table TS.4.1-1c The temperature limit, "310*F", has been replaced with the wording, "the Over Pressure Protection System Enable Temperature specified in the PTLR" and the footnote " Valid until 20 EFPY" has been deleted. The OPPS enable temperature value and the associated reactor fluence limit have been relocated to the PTLR.

3.

TS 31.A.2.c(2). TS 3.1.A.2.c(3). and TS 3 3.A.4 The temperature limit, *200*F", has been replaced with the wording, "the temperature specified in the PTLR for disabling both safety injection pumps." The mass addition transient calculation in OPPS Setpoint Analysis used 200 'F for an analytical limit and assumed that only three charging pumps and no SI (safety injection) pumps would be i

available to inject into the RCS when below 200 'F. This analytical limit has been relocated to the PTLR.

4.

TS 3.1.A.2.c i

The temperature dependency for PORV limiting conditions for operation has been changed from "310 'F" to "350 'F". The reference to the reactor fluence limit associated with the OPPS Enable Temperature has been removed. The 350 'F temperature value is not related to the brittle fracture limitations on the reactor vessel material.

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IS 3.1.B 1.a Changed wording to require that the RCS temperature and pressure limits and heatup and cooldown rates shall be maintained within the limits specified in the PTLR. Deleted the specific wording identifying maximum heatup and cooldown rates.

t 6.

Fiaure TS.3.1 1 and Fiaure TS.3.1-2 These figures have been deleted. Revised figures are provided in the PTLR.

7.

TS 3.3.A.1.c Reference to the OPPS enable temperature of "310 *F" has been removed. Reference to TS 3.1.A.d(2) and the footnote have been removed.

8.

TS 3.3.A J Added the requirements that the RCS accumulators are isolated from the RCS whenever RCS temperature is less than the OPPS enable temperature. This specification will not apply whenever the accumulator is depressurized or the reactor vessel head is removed.

g.

TS 6.7.A.7 Added the requirements for the PTLR and reporting in accordance with the guidance of GL 96-03.

10.

Basis 3.1.A. Basis 3.1.B and Basis 3.3 The bases for Specifications 3.1.A,3.1.B and 3.3 are revised in accordance with the changes made in the specifications stated above.

Relocation of the P/T curves and OPPS setpoints does not eliminate the requirement to operate in accordance with the limits specified in Appendix G to 10 CFR Part 50. The requirement to l

operate within the limits in the PTLR is specified in, and controlled by, the TS, Only the figures, values, and parameters associated with the P/T limits and OPPS setpoints are to be relocated to the PTLR. In order for the curves to be relocated to a PTLR, a methodology for their development must be reviewed and approved in advance by the NRC. The methodology to be approved by the NRC is to be developed in accordance with GL 96-03. This generic letter provides guidance regarding referencing the methodology and development of the PTLR including, but not limited to, the requirements of Appendix G to 10 CFR Part 50. Since the methodology is referenced in the TS, changes to the methodology must be approved by the o

NRC. Further, when changes are made to the figures, values, and parameters contained in the l

PTLR, the PTLR is to be updated and submitted to the NRC upon issuance.

On this basis the NRC staff concludes that the licensee provided an acceptable means of establishing and maintaining the detailed values of the P/T limit curves and OPPS system limits. Further, because plant operation continues to be limited in accordance with the requirements of 10 CFR Part 50 Appendix G and the P/T and OPPS limits in the TS will be

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~ established using's methodology approved by the NRC, these changes will not impact plant safety.'

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~ The staff also concludes that the above-relocated requirements relating to the P/T limits and OPPS limits are not required to be in the TS under 10 CFR 50.36, and are not required to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety, Accordingly, the staff concludes that the proposed changes are acceptable and that these requirements may be relocated from the TS to the PTLR.

A detailed discussion of the staffs basis for acceptance of the licensee's proposed i

methodology is provided in the attached letter from C. A. Carpenter, NRC, to R. Anderson, NSP, " Prairie Island Nuclear Generating Plant,~ Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report," dated April 29,1998.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Minnesota State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change i

surveillance requirements. The amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration and there has been no public comment on such finding (63 FR 14972). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental

- impact statement or environmental assessment need be prepared in connection with the

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issuance of the amendments.

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6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

1 PrincipalContributor: B.Wetzel l

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Date:'

May 4, 1998

Attachment:

As stated

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UNITED STATE 8 g

NUCLEAR RE2ULATORY COMMISSION U

WAsMINGTON, o.C. anseedoot April 29, 1998 l

Mr. Roger O. Anderson, Director Nuclear Energy Engineering Northem States Power Company 414 Nicollet Mall Minneapolis, Minnesota 55401 l

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2, ACCEPTANCE FOR REFERENCING OF PRESSURE TEMPERATURE LIMITS REPORT (TAC NOS. MA1121 AND MA1122)

REFERENCES:

1. WCAP-14040-NP-A, Revision 2 Westinghouse Electric Corporation,

" Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January 15,1996.

2. NRC Generic Letter 96-03," Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits,"

January 31,1996.

3. Letter from Joel P. Sorensen, Northem States Power Company, to NRC Document Control Desk, " Amendment of Technical Specifications to Update the Heatup and Cooldown Rate Curves, incorporate the Use of a Pressure Temperature Limits Report, and Change the Pressurizer Power Operated Relief Valves Operability Temperature," March 6,1998. (WCAP-14780 and l

WCAP-14637 are Attached).

4. Letter from Joel P. Sorensen, Northem States Power Company, to NRC Document Control Desk, " Revised Prairie Island Units 1 and 2 Reactor Vessel Material Surveillance Repor1s," March 6,1998. (WCAP-14779, i

Revision 2, WCAP-14781, Revision 3, WCAP-14613, Revision 2, and l

WCAP-14638, Revision 2 are Attached).

5. Letter from Joel P. Sorensen, Northem States Power Company, to NRC Document Control Desk, " Response to March 16 and 19,1998 Requests for Additional Information for License Amendment Request dated March 6, 1998," March 30,1998.

6. Letter from Joel P. Sorensen, Northem States Power Company, to NRC Document Control Desk, " Response to March 13,1998 Request for Additional Information for License Amendment Request dated March 6, 1998," March 31,1998.

7. Letter from Joel P. Sorensen, Northem States Power Company, to NRC Document Control Desk, " Supplement to the License Amendment Request dated March 6,1998," April 13,1998.

l The NRC staff has completed its review of the pressure temperature (P/T) limit curves and low temperature overpressure protection (LTOP) system limits methodology and the pressure I

temperature limits report (PTLR) submitted by the Northem States Power Company (NSP), We I

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Contact:

Maggalean W. Weston, (301) 415-3151 i

Attachment to safety evaluation dated May 4, 1998 Os AM J O VJ uJ& a q %

ff s

R. O. Anderson 2

April 29, 1998 find the methodology to be acceptable for referencing in the administrative controls section of the Prairie Island Nuclear Generating Plant, Units 1 and 2. Technical Specifications to the extent specified and under the limitations delineated in your submittals and the associated NRC Safety,_

Evaluation, which is enclosed. The Safety Evaluation defines the basis for acceptance of the submittals. Our acceptance applies only to the matters described in the submittals.

The NRC notes that NSP should address the method for assessing the credibility of the reactor pressure vessel surveillance capsule data consistent with the criteria provided in the revised rule (10 CFR 50.61) or in Regulatory Guide 1.99, Revision 2, in future revisions to the PTLR.

The methodology for review relating to the P/T limit curves and the LTOP system limits was provided in the references listed above. WCAP-14040 NP-A provided, in part, the methodology used for determining the acceptance of the Prairie island methodology.

The methodology in WCAP-14040 NP A, along with supplements provided by NSP, will be used to calculate future changes to the P/T limit curves. NSP may generate new P/T limit curves in accordance with this methodology without prior approval of the staff. However, changes to the methodology must first be reviewed and approved Dy the staff. System limits may be subject to audit by the staff through inspections as necessary.

We do not intend to repeat our review of the matters described in the submittals if the submittals appear as references in other license applications relating to your plants, except to ensure that the material in the submittals is still applicable to your plants as indicated in the conclusion section of the Safety Evaluation.

Should our criteria or regulations change so that our conclusions as to the acceptability of the methodology is invalidated, licensees referencing these documents will be expected to revise and resubmit their respective documentation, or submit justification for the continued effective applicability of the documents without revision of their respective documentation.

Sincerely, gnthia A. Carpenter, Director Project Directorate ill-1 Division of Reactor Projects - Ill/IV l

Office of Nuclear Reactor Regulation Docket Nos. 50282 and 50-306

Enclosure:

Safety Evaluation ocw/ encl: See next page I

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_, _ - - - - - - - -, - - - - - - - - - - - - - - ' - - - ~ ~ - - -

Mr. Roger O. Anderson, Director Prairie Island Nuclear Generating Northem States Power Company Plant l

cc:

l J. E. Silberg, Esquire Sito Uconsing Shaw, Pittman, Potts and Trowbridge Prairie Island Nuclear Generating

- 2300 N Street, N. W.

Plant i

i Washington DC 20037 Northem States Power Company 1717 Wakonada Drive East Plant Manager Welch, Minnesota 55089 Prairie island Nuclear Generating Plant Tribal Council Northem States Power Company Prairie Island indian Community 1717 Wakonada Drive East ATTN: EnvironmentalDepartment Welch, Minnesota 55089 5636 Sturgeon Lake Road Welch, Minnesota 55089 Adonis A. Nebiett Assistant Attomey General Office of the Attomey General 455 Minnesota Street Suite 900

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St. Paul, Minnesota 55101-2127 I

U.S. Nuclear Regulatory Commission Resident inspector's Office 1719 Wakonade Drive East Welch, Minnesota 55089-9642 Regional Administrator, Region ill U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Mr.Jeff Cole, Auditor / Treasurer Goodhue County Courthouse Box 408 Red Wing, Minnesota 55066-0408 Kris Sanda, Commissioner Department of Public Service 121 Seventh Place East Suite 200 St. Paul, Minnesota 55101-2145 m

en sees i

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p UNITED STATES NUCLEAR REGULATORY COMMISSION WAsHINoToN, o.C. SteeHoot SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

^

l REVIEW OF PRESSURE TEMPERATURE LIMITS REPORT AND METHODOLOGY FOR THE RELOCATION OF THE REACTOR COOLANT SY8 TEM PRESSURE TEMPERATURE LIMIT CURVES AND LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM LIMITS NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING Pl. ANT. UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306 i

1.0 INTRODUCTION

By letter dated March 6,1998 (Reference 3), and supplemented by letters dated March 6,1998 (Reference 4 ), March 30,1998 (Reference 5), March 31,1998 (Reference 6), and April 13,1998 l

(Reference 7), Northem States Power Company (NSP) requested changes to the technical i

specifications (TS) for the Prairie Island Nuclear Generating Plant, Units 1 and 2. The requested changes included (1) revising the reactor coolant system (RCS) pressure temperature (P/T) limit curves and low temperature overpressure protection (LTOP) system limits, (2) relocating the P/T limit curves and LTOP system limits from the TS to a licensee-controlled document identified as a Pressure Temperature Limits Report (PTLR), and (3) changing the affected limiting conditions for operation and bases accordingly. The P/T limit curves and LTOP system setpoints were developed, in part, using the staff approved methodology documented in WCAP-14040-NP-A, l

Revision 2 (Reference 1). These changes are made in accordance with Generic Letter 96-03,

" Relocation of the Pressure Temperature Limit Curves and Low Temperature overpressure Protection System Limits," dated January 31,1996 (Reference 2). Generic Letter 96-03 provides licensees the option to relocate the P/T limit curves and the LTOP system setpoints to a licensee controlled PTLR provided that the limiting curves and setpoints are developed using an NRC approved methodology. The licensee proposes to extend the period of applicability of the P/T limit curves and LTOP system setpoints to 35 effective full power years (EFPYs) of reactor operation.

2.0 BACKGROUND

2.1 Neutron Fluence The fluence evaluation which is the basis for the proposed revised P/T curves is documented in WCAP-14779, Revision 2, and WCAP-14613, Revision 2, for Units 1 and 2, respectively. The evaluation of the pressurized thermal shock is documented in WCAP-14781, Revision 3, and l

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2-WCAP-14638, Revision 2, for Units 1 and 2, respectively. The surveillance capsule reports in WCAP-1477g and WCAP-14613 document the evaluation of capsules S and P for Units 1 and 2, respectively, in addition they document the reevaluation of previously smoved capsules (V,P,R) and (V,T, R) for Units I and 2, respectively. The analyses were performed using the BUGLE g3 cross sections in the DOT computer program. The BUGLE-g3 cross sections are based on the ENDF/B-VI cross section file which is the staff-recommended file. In addition the analyses utilized the Pa and Se approximations for angular elastic scattering and special quadrature, respectively.' These approximations are recommended by the staff.

2.2 Pressure Temperature Limits The methodologies for assessing P/T limits and reactor pressure vessel (RPV) surveillance programs are discussed, in part, in the following documents: (1) 10 CFR Part 50, " Appendix G -

Fracture Toughness Requirements"; (2) 10 CFR Part 50, " Appendix H - Reactor Vessel Material Surveillance Program Requirements"; (3) 10 CFR 50.60 " Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation"; (4) 10 CFR l

l 50.61 - Tracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events"; and (5) Regulatory Guide (RG) 1.gg, Revision 2

• Radiation Embrittlement of Reactor Vessel Materials."

NSP has applied the methodology of WCAP-14040-NP-A, Revision 2, as the general methodology for generating the P/T limit curves for heatup, cooldown, and hydrostatic testing conditions of the Pl-1 and Pl-2 [ Prairie Island] reactor coolant pressure boundaries (RCPBs).

Westinghouse Electric Corporation (WEC) submitted this methodology to the staff as its basis for developing the cold overpressure mitigating system setpoints and RCPB heatup and cooldown limit curves for WEC-designed nuclear reactors.

I Pursuant to 10 CFR Part 50, Appendix G, the P/T limits and minimum temperatures established l

for RPVs must meet the requirements for these parameters as set forth in Table 1 of the rule. In addition to the minimum requirements, the P/T limit curves are required to be at least as I

conservative as those that would be obtained by following the methods of analysis and the safety margins found in the ig8g Edition of the American Society of Mechanical Engineers (ASME)

Code,Section XI, Appendix G.

t 2.3 Low Temperature Overornamure Protadan Svatem The licensee designated the LTOP system as the over Pressure Protection System (OPPS).

The OPPS mitigates overpressure transients at low temperatures so that the integrity of the RCPB !s not compromised by violating the 10 CFR Part 50, Appendix G P/T limits under steady-state operating conditions. Prairie Island Units 1 and 2 OPPS uses the pressurizer power-operated relief valves (PORVs) or an RCS vent with the reactor depressurized to accomplish this function. The system is manually enabled by operators and uses a single setpoint as the lift pressure for the PORVs. The design basis of Prairie Island Units 1 and 2 OPPS considers both mass-addition and heat addition transients. The mass-addition analyses in the supporting PTLR account for the injection from up to three charging pumps to the RCS in the full range of P/T conditions starting from 68 T. For an RCS temperature greater than or equal to 200 T an inadvertent injection from one safety injection pump and a maximum of three charging pumps are assumed. The heat-addition analyses account for heat input from the secondary side of the steam generators into the RCS upon starting a single reactor coolant pump l

(RCP) when the RCS temperature is as much as 50 Y lower than the steam generator l

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• secondary side temperature. The proposed TS provided restriction in plant operation within the configuration assumed in the analysis for OPPS design.

The Prairie laland Units 1 and 2 proposed design of OPPS including the determination of its i

enable temperature and the PORV actuation setpoint was established using the sta#-approved' methodology documented in WCAP-14040-NP-A. Also, the licensee has applied for an exemption from certain requirements of 10 CFR Part 50, Appendix G, and adopted a provision in ASME Code Case N 514 that permits a 10% relaxation of the P/T limits in its design of OPPS.

3.0 EVALUATIONS 3.1 Neutron Fluence i

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The surveillance capsule reports document the measured and calculated values of four capsules for each plant. The measured to calculated (M/C) ratios of the pressure vessel fluence values i

are consistent, and their deviation from unity is reasonable. Likewise, the different dosimeter response M/C ratios are consistent and close to unity. Overall, the measured values are slightly higher than the corresponding calculated values. The licensee adopted the calculated values (for each unit) for the estimation of the P/T curves. This is conservative and, tnerefore, it is acceptable.

The Unit 1 estimated pressure vessel inside diameter fluence value for 35 EFPYs is 3.95x10

l n/cm'. The corresponding value for Unit 2 is 4.18x10 n/cm'. These values were taken into account in the P/T curves. These values were derived using staff-recommended cross sections and approximations; therefore, we find them acceptable.

3.2 Pressure Temperature (P/T) Limits 3.2.1 Revised P/T Lirnit Heatun and CMdown Curves for the PI-1 and PI-2 RPVs The staff performed an independent analysis using the methods described in 10 CFR Part 50, Appendix G and in Standard Review Plan (SRP) 5.3.2,

• Pressure Temperature Limits," in order to determine whether NSP's methods for determining the minimum allowable RCS pressures and temperatures during heatup, cooldown, and hydrostatic testing conditions were conservative relative to the stafs analysis. For the stafs evaluations of the beltline materials, the staff applied the methodology found in 10 CFR Part 50, Appendix G. The stafs analysis methods i

were consistent with those applied by NSP with the following exceptions:

The stars pressure stress equation was based on a simple hoop stress equation.

The stafs method for evaluating the thermal gradient across the RPV wall was based on a simple steady-state thermal gradient.

The stats method for determining the stress intensities due to thermal stresses was based on Figure 4-5 of the Welding Research Council (WRC) Bulletin 175.

The staff followed the criteria of the revised rule 10 CFR 50.61 and RG 1.99, Revision 2, as its basis for calculating the end of life (EOL) 1/4t and 3/4t RT, values, and the RTm values for the beltline materials in the Pl-1 and PI-2 RPVs.

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3.2.2 Assessment of the RT.7, Values for the PI-1 and PI-2 Balthne Materials The staff performed an independent assessment of the RTm values for the beltline materials in the Prairie Island reactor vessels. The staff determined that the licensee's calculations of the RTm values were in agreement with those that would be generated if the methods of the reviseo rule 10 CFR 50.61 or RG 1.99, Revision 2, were applied. For purposes of calculating the limiting j

RTm value, the staff verified that the licensee correctly calculated the limiting projected RTm values for the PI-1 and PI-2 vessels to be 162 'F and 143 'F, which are the values calculated for the PI-1 nonle to intermediate shell forging circumferential weld and for the PI-2 upper forging to intermediate forging seam weld W2, respectively. These values are significantly less than the j

screening criterion of 300 'F as stated in the revised rule 10 CFR 50.61, and indicate that the Prairie Island vessels will continue to satisfy the requirements of the rule throughout the projected lives of the plants.

3.2.3 Assessment of the EOL 1/41 and 3/4t RT, Values and the Premand Hamhm Cm!down.

and Hydrostatic Testino Curves for PI-1 and Pl.2 j

The staff also performed an independent assessment of the EOL 1/4t and 3/4t RT values for the beltline materials in the PI-1 and PI-2 reactor vessels, and of the proposed P/T limit curves for the PI-1 and PI 2 reactor vessels during heatup, cooldown, and hydrostatic testing conditions.

The assessment of each unit is discussed below.

.j For the PI-1 reactor vessel, the licensee determined that the most limiting material at the 1/4t and I

3/41 locations is the nonle to intermediate shell circumferential weld. This weld was fabricated using weld wire heat 2269. The licensee calculated an RT value of 154 'F at the 1/4t location and 136 'F at the 3/4t location at 35 EFPYs. The neutron fluence used in the RT, calculation was 1.47 X 10 n/cm' at the 1/4t location and 0.66 X 10 n/cm' at the 3/4t location. The initial RTa value for the limiting weld was 0 'F. The margin term used in the calculation for the limiting weld was 66 'F for both the 1/4t and 3/4t locations. This number is consistent with the number that is generated when a generic mean value is used to establish the unirradiated RT, for a beltline weld.

l The staff performed an independent calculation of the RT, values for the limiting material using the methodology in RG 1.99, Revision 2. Based on these calculations, the staff verified that the l

licensee's limiting matenal for the PI-1 reactor vessel is the nonle to intermediate shell l

circumferential weld that was fabricated using weld wire heat 2269. The staff's calculated RT, value for the limiting material agreed with the licensee's calculated RTa value at 35 EFPYs.

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Substituting the RTa value, for the Pl-1 limiting weld into the equations in SRP 5.3.2, the staff verified that the proposed PR limits satisfy the requirements in paragraph IV.A.2 of Appendix G of 10 CFR Part 50.

For the PI-2 reactor vessel, the licensee determined that the most limiting material at the 1/4t l

and 3/4t locations is the upper to intermediate shell weld seam. This wold was fabricated using i

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weld wire heat 1752. The licensee calculated an RT value of 134 *F at the 1/4t location and 116 'F at the 3/4t location at 35 EFPYs. The neutron fluence used in the RT calculation was 8

1.59 X 10 n/cm at the 1/41 location and 0.71 X 10 n/cm' at the 3/41 location. The initial RT.

value for the limiting weld was -13 'F. The margin term used in the calculation for the limiting weld was 56 'F for both the 1/4t and 3/4t locations. This number is consistent with the number that is generated when a generic mean value is used to establish the unirradiated RT, for a beltline weld.

I

<- 1 The sta# performed an independent calculation of the RT values for the limiting material using the methodology in RG 1.99, Revision 2. Based on these calculations, the sta# verified that the licensee's limiting material for the PI 2 reactor vessel is the upper to intermediate shell weld seem that was fabricated using weld wire heat 1752. The staffs calculated RT value for the limiting material agreed with the licensee's calculated RTa value at 35 EFPYs. Substituting the RT values for the PI-2 limiting weld into the equations in SRP 5.3.2, the sta# verified that the proposed P/T limits satisfy the requirements in paragraph IV.A.2 of Appendix G of 10 CFR Part 50.

In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes a minimum temperature at the closure head flange based on the reference temperature for the flange material.Section IV.A.2 of Appendix G states that when the pressure exceeds 20% of the preservice system hydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload must exceed the reference temperature of the materialin those regions by at least 120 'F for normal operation and by 90 'F for hydrostatic pressure tests and leak tests. The RT values for the limiting flange materials in the PI-1 and PI-2 reactor vessels are -4 *F and -22 'F, respectively. The staff has determined that the proposed P/T limits satisfy the requirement for the closure flange region during normal operation and hydrostatic pressure tests and leak tests.

3.3 Low Pressure Overpressure Protection System The proposed Limiting Conditions for Operation in TS 3.1.A.2.c require that an OPPS be enabled with two operable PORVs when the RCS temperature is below the OPPS enable temperature.

Also, (1) when the RCS temperature is above the temperature in which the safety injection pumps are not disabled, one PORV may be inoperable for 7 days. If these conditions cannot be met, the RCS must be depressurized and vented through at least a 3-square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. in the case where both PORVs become inoperable, the RCS must be depressurized and vented through at least a 3-square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and (2) when the RCS is below the temperature for disabling both safety injection pumps, one PORV may be inoperable for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If these conditions cannot be met, the RCS must be depressurized and vented through at least a 3-square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. In the case where both PORVs I

become inoperable, the RCS must be depressurized and vented through at least a 3 square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The setpoints related to the design of OPPS and applicable to both Prairie Island Units 1 and 2 are listed in the licensee's PTLR. The sta# evaluation of these setpoints is presented below.

3.3.1 Enable Temperature The OPPS enable temperature is the temperature below which the OPPS is required to be operable. The licensee has established an OPPS enable temperature using the methodology presented in WCAP-14040-NP-A with the provision permitted by ASME Code Case N-514. This Code Case requires the OPPS to be efective at an RCS temperature less than 200 'F or at an RCS temperature corresponding to a reactor vessel metal temperature less than RT

+ 50 'F at the beltline location (1/4t). Therefore, the licensee proposed to calculate the enable temperature as RT, + 50 'F + temperature di#erence between RCS and metal + instrument Uncertainties. Using the above equation with limiting material adjusted reference temperature for Unit 1 as input, the calculated minimum enable temperature applicable for both Prairie Island Units 1 and 2 is 243 'F. The licensee proposed an enable temperature of 310 'F that includes an additional margin of 67 'F.

6-The staff finds that this proposed OPPS enable temperature is conservative with respect to the enable temperature allowed by ASME Code Case N-514 and therefore is acceptable.

3.3.2 Diamhlina Safety Inier*ian Pumo(s) and la*S..c A muk.ic.s in the licensee's analysis for the design of OPPS, it is assumed that when the RCS temperatu is below 200.*F, the OPPS will provide adequate protection for a mass addition from a maximum l

of three charging pumps. When the RCS is between 200 'F and the calculated OPPS enable l

temperature, the OPPS will provide protection for a mass addition from one safety injection pump plus three charging pumps. To support these analysis assumptions, the licensee proposed requirements in TS 3.3.A and the PTLR to disable one safety injection pump when the RCS is above 218 'F and to disable two safety injection pumps when the RCS temperature is below 218 'F. The setpoint of 218 'F includes an instrument uncertainty of 18 'F. Also, TS 3.3.A requires that both accumulators be isolated when the RCS temperature is below the OPPS enable temperature.

The licensee in its letter dated March 31,1998, indicated that the safety injection pumps are rendered incapable of injecting into the RCS by employing at least two independent means to I

prevent a pump start such that a single action will result in an injection into RCS. This may be accomplished through the pump control switch being placed in pullout with a block device Installed over the control switch that would prevent an unplanned start. This method of disabling l

the safety injection pump has been stated in the TS Bases 3.3. We find that ti el licensee-proposed method of disabling the safety injection pump is acceptable.

3.3.3 PORV Actuation Satoolnt OPPS is designed to mitigate overpressure transients at low temperatures to prevent violating 10 CFR Part 50, Appendix G P/T limits. Additionally, since overpressure events most likely occur during isothermal conditions in the RCS, the NRC has accepted the use of the steady-state Appendix G limits for the design of the OPPS. The OPPS actuation setpoint is the pressure at which the PORVs wi!! lift, when the OPPS ts enabled, to limit the peak RCS pressure during a pressurization transient.

Prairie Island Units 1 and 2 use PORVs to provide pressure relief capacity for the OPPS. The methodology used for determining the PORV actuation setpoint is consistent with the methodology presented in WCAP-14040 NP-A.

The licensee-proposed PORV actuation setpoint of 500 psig in the PTLR was calculated in accordance with the proposed methodology. The licensee, in its submittal dated March 6,1998, gut.d.d a tabulation listing PORV setpoints, transient pressure overshoot, instrumentation uncertainties, pressure difference between the pressure transmitter and the reactor vessel mid-plane with one or two RCPs in operation and conssponding P/T limits under various temperature conditions below the OPPS enable temperature. The data presented in this tabulation confirms that the proposed PORV setpoint of 500 psig will provide adequate protection to the P/T limits established by 10 CFR Part 50, Appendix G, with the provision of ASME Code Case N-514 under steady-state conditions during a design-basis overpressure transient (mass-addition or heat addition) as described in Section 1.0 of this report. Based on the above discussion, we find the proposed PORV setpoint acceptable.

1

.* 3.3.4 RCS Vent Size The proposed TS 3.1.A.2.c speci6es a vont size of 3 square inches as an altemative to an operable OPPS when the RCS is depressurtzed, The bases for the 3-square inch vent is stated in the TS Bases, page B.3.1.-3. It states that the vent size is based on the 2.956-square inch ^^

cross sectional flow area of a pressurizer PORV. Since the vent size is compatible with the PORV size, which is sufficient to mitigate a design-basis overpressure transient, we find it acceptable.

4.0 CONCLUSION

S Based upon the staf evaluations, as discussed in Section 3.0 above, the NRC staW concludes that it is acceptable for NSP to relocate the P/T limit curves and LTOP system limits from the

- Prairie Island Units 1 and 2 TS to a licensee-controlled PTLR. The proposed heatup, cooldown, and hydrostatic testing curves for PI-1 and PI-2 will expire at 35 EFPYs.

The staff has reviewed the proposed fluence values for Prairie island Units 1 and 2 for the revision of the P/T curves and finds that the proposed values are conservative and, therefore, acceptable. In addition, the staff has determined that the proposed P/T limit curves are acceptable for use and are consistent with the requirements of Appendix G to 10 CFR Part 50, and Appendix G to Section XI of the ASME Code.

However, since WCAP-14040-NP-A does not address the credibility of RPV surveillance material, the licensee, in future P/T limit evaluations, should address the credibility of the surveillance material.

The staff also reviewed the licensee's analyses related to the proposed setpoints of the OPPS as

- oiscussed in Section 3.0 above. The licensee has considered instrument uncertainties in its setpoint calculation using instrument Society of America S67.04-1994. The staff finds that the licensee's analyses were performed in a manner consistent with the approved methodology and that the results of the analyses conservatively demonstrated that the P/r limits established by 10 CFR Part 50, Appendix G, with provisions provided by ASME Code Case N-514 will be adequately protected with these setpoints and, therefore, finds NSP's analyses acceptable.

The staff has determined that the proposed PTLR meets the criteria of Generic Letter 96-03, and is acceptable to the staff.

5.0 REFERENCES

1.

WCAP-14040-NP-A, Revision 2, Westinghouse Electric Corporation, " Methodology used to i

Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January 15,1996.

2.

NRC Generic Letter 96-03, " Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," January 31,1996.

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8-3.

Letter from Joel P. Sorensen, Northem States Power Company, to NRC Document Control Desk, "Ame.ndment of Technical Specifications to Update the Hestup and Cooldown Rate Curves, incorporate the Use of a Pressure Temperature Limits Report, and Change the Pressurizer Power Operated Relief Valves Operability Temperature," March 6,1998.

(WCAP-14780 and WCAP-14637 are Attached).

N 4.

Le'. er from Joel P. Sorensen, Northem States Power Company, to NRC Document Control DesP, " Revised Prairie Island Units 1 and 2 Reactor Vessel Material Surveillance Reports,"

March 6,1998. (WCAP-14779, Revision 2, WCAP-14781, Revision 3, WCAP-14613, Revielos 2, and WCAP-14638, Revision 2 are Attached).

5.

Letter frora Joel P. Sorensen, Northem States Power Company, to NRC Document Control Desk, "Respo.we to March 16 and 19,1998 Requests for Additional information for License Amendment Reqast dated March 6,1938,

• March 30,1998.

6.

Letter from Joel P. Sorensen, Northem Sta;05 Power Company, to NRC Document Controi Desk, " Response to March 13,1998 Request for Additional Information for License Amendment Request dated March 6,1998,

• March 31,1998.

7.

Letter from Joel P. Sorensen, Northem States Power Company to NRC Document Control Desk, ' Supplement to the License Amendment Request dated March 6,1998,* April 13, 1998.

Principal Contributors: M. Khanna C.Y. Liang L. Lois M. W. Weston Date: April 29, 1998 l