ML20083B625

Application for Amends to Licenses DPR-42 & DPR-60.Amends Would Revise Plant MSSV Lift Setting Tolerances

ML20083B625
Person / Time
Site:	Prairie Island
Issue date:	05/04/1995
From:	Wadley M NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20083B627	List: ML20083B625 ML20083B635
References
NUDOCS 9505120140
Download: ML20083B625 (19)

Text

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-[ Northem States Power Crmpany Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. East Welch, Minnesota 55089 May 4, 1995 10 CFR Part 50 Section 50.90 U S Nuclear Regulatory Commission Attn; Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PIANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 License Amendment Request Dated May 4, 1995 Pressurizer Safety Valves and Main Steam Safety Valves Lift Setting Tolerance Chance and Safety Limit Curve Chances Attached is a request for change to the Technical Specifications, Appendix A of the Operating Licenses, for the Prairie Island Nuclear Generating Plant.

This request is submitted in accordance with the provisions of 10 CFR Part 50, Section 50.90. The requested amendment will revise the Prairie Island pressurizer and main steam safety valve lift setting tolerances from i 1% to 3%, revise the Safety Limit curves and revise Technical Specification Section 2 including to conform to Standard Technical Specifications.

Exhibit A contains a description of the proposed changes, the reasons for requesting the changes, and the supporting safety evaluation and significant hazards determination. Exhibit B contains current Prairie Island Technical Specification pages marked up to show the proposed changes. Exhibit C contains the revised Technical Specification pages.

In the 1994 Unit I outage, two Prairie Island main steam safety valve lift settings were determined to be out of the 1% setpoint tolerance allowed by the Prairie Island Technical Specifications. This was reported to the Commission in Licensee Event Report (LER) 1-94-04. This license amendment request fulfills the commitment made in LER l-94-04 to propose a change to Technical Specification 3.4.A.l.a to allow as-found main steam safety valve measurement tolerance of 3% and makes the same provision for the pressurizer safety valves. The proposed Technical Specification changes require that the pressurizer and main steam safety valve setpoints be restored within 11% of 3 their nominal setpoints following testing.

9505120140 950504 PDR ADOCK 05000282 f(j\I P PDR V i

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USNRC May 4, 1995 l Page 2 i

This license amendment request also proposes changes to the Safety Limit curves which incorporate the revised safety valve lift setting tolerance and account for changes in Prairie Island's fuel design and removal of thimble i plugs from the core. With these latter changes, the Safety Limit curves more '

accurately and conservatively present the plant safety limits. Northern States Power Company is also using this opportunity to propose other changes to  !

Technical Specifications Section 2 to correct typographical errors and conform portions to the Standard Technical Specifications.

If you have any quentions related to this License Amendment Request please contact myself or Dale Vincent at 612-388-6758 X4107.

M. D. Wadley i Plant Manager Prairie Island Nuclear Generating Plant 1

c: Regional Administrator-III, NRC -

NRR Project Manager, NRC Senior Resident Inspector, NRC i State of Minnesota Attn: Kris Sanda ,

J E Silberg Attachments:

Affidavit Exhibit A - Description of Proposed Changes the Reasons for Requesting the Changes, and the supporting Safety Evaluation and Significant Hazards Determination Exhibit B - Technical Specification Marked Up Pages Exhibit C - Revised Technical Specification Pages i

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a UNITED STATES NUCLEAR REGUIATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PIANT DOCKET NO. 50-282 50-306 REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 & DPR-60 LICENSE AMENDMENT REQUEST DATED May 4,1995 Northern States, Power Company, a Minnesota corporation, requests authorization for changes to the Prairie Island Operating License, Appendix A as shown on the attachments labeled Exhibits A, B, and C. Exhibit A describes the proposed changes, reasons for the changes, and the supporting safety evaluation and significant hazards determination. Exhibit B contains current Prairie Island Technical Specification pages marked up to show the proposed changes. Exhibit C contains the revised Technical Specification pages.

This letter contains no restricted or other defense information.

NORTHET.N STATES POWER COMPANY By M M. D. Wadley /['

Plant Manager U Prairie Island Nuclear Generating Plant On this County, personally appe E ay of Ad h before me a notary public in and for said NuclearGeneratingPlant,pM.D.Wadley,PlantManager,PrairieIsland and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the statements made in it are true and that it is not interposed for delay.

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. I LICENSE AMENDMENT REQUEST DATED May 4, 1995  ;

Pressurizer Safety Valves and Main Steam Safety Valves Lift '

Settina Tolerance Chanas and Safety Limit Curve' Chances l EXHIBIT A Description of the Proposed Changes, The Reasons for Requesting the Changes, and the Supporting Safety Evaluation /Significant Hazards Determination Pursuant to 10 CFR Part 50, Sections 50.59 and 50.90, the holders of Operating Licenses DPR-42 and DPR-60 hereby propose the ,

following changes to the Facility Operating Licenses'and Appendix A, Technical Specifications:

BACKGROUND This license amendment request proposes to change the lift setting tolerance of the Pressurizar Safety Valves and Main Steam Safety Valves from the current Technical Specification tolerance of 1% to 3%. Following testing, the lift settings shall be I left within 1%. Also, the SAFETY LIMIT curves have been revised i to accommodate this and other changes. Since changes are required in Chapter 2, Northern States Power Company also proposes changes -

to conform portions of Technical Specification Section 2 to the Standard Technical Specifications. 1 1

Over-pressure protection for the Reactor Coolant System and Main Steam System is provided in part by the Pressurizer Safety Valves ,

and the Main Steam Safety valves located on the pressurizar and i the two main steam lines respectively. For the Reactor Coolant I I

system there are two Pressurizer Safety Valves set at 2485 psig.

The two main steam lines have a total of 10 Main Steam Safety Valves, five on each line with settings of 1077, 1093, 1110, 1120 and 1131 psig.

The Pressurizer Safety Valves and Main Steam Safety Valves were designed to meet the requirements of USAS B31.1.0 - 1967. The original design established the setpoint tolerance of i 1%. This l also agrees with ASME Section III for Class 1 valves which requires the setpoint tolerance to be within i 1% when operating pressure is greater than 1000 psi.

In 1994, two Prairie Island Unit I Main Steam Safety Valves were outside the i 1% tolerance which was reported to the Nuclear Regulatory Commission in Licensee Event Report 1-94-04. In this Licensee Event Report, Northern States Power Company committed to prepare a license amendment request which would propose e change to Technical Specification 3.4.A.1.a to allow as-found valve measurement tolerances of i 3%.

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As-found lift setting variation from nominal setpoints of Pressurizer Safety Valves and Main Stehn Safety Valves is an l industry-wide occurrence not unique to Prairie Island. The industry, through the technical consensus of ASME, has addressed the issue in ASME OM-1987, Part 1 which allows the test lift setting to vary by i 3% from the nominal setpoint for safety valves operating above 1000 psig. Also the Nuclear Regulatory ,

Commission has addressed the issue of Pressurizer Safety Valva l and Main Steam Safety Valve setpoint variability as detailed in i Information Notice IN 86-56, " Reliability of Main Steam Safety l Valves," dated July 10,1986. i These requested amendments will revise the setpoint tolerances for both the Pressurizer Safety Valves and Main Steam Safety Valves to assure they remain within Technical Specification tolerance and reduce the potential for filing subsequent Licensee ,

Event Reports. ,

i Pressurizer Safety Valve and Main Steam Safety Valve testing will [

continue to be performed in accordance with ASME Section XI. As i required by ASME Section XI, Prairie Island Nuclear Generating Plant performs inservice testing on these safety valves in  !

compliance with ASME OM-1987, Part 1 which allows the test lift tolerance to vary by 3%. The i 3% tolerance proposed by this license amendment request will be used for the OPERABILITY  !

determination and test acceptance criteria in lieu of the current  ;

i 1% setpoint tolerance. The lift setting of the Pressurizer Safety Valves and Main Steam Safety Valves will be restored to within 1% whenever a test determines the lift settings are outside of the nominal i 1%. i Northern States Power Company notes that similar amendment 'p requests on relaxing Pressurizer Safety Valve and Main Steam Safety Valve setpoint tolerances have been approved by the -

Nuclear Regulatory Commission for the Seabrook Station, Unit 1, the Virgil C. Summer Nuclear Station, the Donald C. Cook Nuclear .

Plant, Unit Nos. 1 and 2 and more recently for Vogtle Electric '

Generating Plant Units 1 and 2 and Joseph M. Farley Nuclear Plant Units 1 and 2. The portions of this proposed amendment related to changing the Pressurizer Safety Valve and Main Steam Safety Valve setpoint tolerances are similar to these previously approved by the Nuclear Regulatory Commission.  :

Additionally, Northern States Power Company is seeking approval  !

for changing the Prairie Island SAFETY LIMITS curves in Figure  :

TS.2.1-1 which will make them more conservative, that is, more restrictive. This request proposes to revise the SAFETY LIMIT l curves to adjust for the impact of relaxing the lift setting tolerances for the Pressurizer Safety Valves and Main Steam ,

Safety valves and other changes. The most significant change is  ;

in the DNB limit part of the SAFETY LIMIT curves. The DNB limit J curves were significantly lowered (less DNB margin) because the  !

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. 1 bypass flow fraction was increased when the fuel thimble plugs were removed; the Fu limit was increased with the LOCA analysis performed for PI2-16; and rod bow penalty was added to the DNB calculation. Other changes to Figure TS.2.1-1, are changes in  !

format which will be discussed in more detail below.

Since revision of Technical Specification Section 2 is required by the amended setpoint tolerance change and other proposed changes to the SAFETY LIMIT curves, additional administrative changes are also proposed to bring portions of Section 2 into conformance with the Standard Technical Specifications and correct typographical errors.

PROPOSED CHANGES AND REASONS FOR CHANGES The proposed changes to Prairie Island Operating License Appendix A, Technical Specifications are described below, and the specific wording changes are shown in Exhibits B and C.

1. TABLE OF CONTENTS: Revise to reflect the reformatting of Section 2, deletion of Section 6.4, reformatting of Bases Section 2, and addition of a figure.

Justification: Table of Contents revised to reflect the requested license amendments.

2. Technical Specification Section 2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING: Combine the specifications of Sections 2.1 and 2.2 into one section. On page TS.2.3-2 correct typographical errors the most significant of which is changing "-9%" to "+9%". Make administrative change on page TS.2.3-3 by realigning the margin for Specification 2.3.A.3.

Justification: This change will simplify the requirements of these two sections and bring them into conformance with the guidance of the Standard Technical Specifications.

Also Section 6.4 specifications have been included in this section to further simplify the Technical Specification.

The range of power differences between power in the top and bottom halves of the core is corrected to "+9%" along with other obvious typographical errors in the preceding paragraph. Subsection 2.3.A.3 is currently aligned on page TS.2.3-3 such that it appears to be a third item under the heading of Specification 2.3.A.2.1, " Power range neutron flux rate". In fact, item 3, "Other reactor trips", is a separate subsection under Specification 2.3.A, " Protective instrumentation settings for reactor trip shall be as follows:". Realignment of the margin for this subsection should eliminate the confusion.

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3. Technical Specification LIMITING CONDITION FOR OPERATION 3.4.A.1.a.: Change Main Steam Safety Valve OPERABLE lift setting tolerance from " 1%" to "i 3%".

Justification 1 Northern States Power Company proposes to revise the OPERABILITY and test acceptance lift setting tolerance for the Main Steam Safety Valves to allow for 1 lift pressure variability due to test instrumentation 1 inaccuracies and drift during operation in accordance with current industry guidance.

4. Technical Specification SURVEILLANCE REOUIREMENTS Table TS.4.1-2A: Specify test acceptance criteria as i 3% of the 1 nominal setpoint for both Pressurizer Safety Valves and Main l Steam Safety Valves and reformat the table to read easier and  !

fit on two pages. l Justification: Northern States Power Company proposes to  !

revise the test lift setting acceptance criteria tolerance to conform with guidance provided in Standard Technical Specifications and ASME Code. Reformatting of the table is 1 necessitated by the addition of the proposed changes.

5. Technical Soecification 6.4 SAFETY LIMIT VIOLATION: Delete this Section in its entirety.

Justification: The requirements of this specification have i been relocated to Section 2.2 in conformance with the guidance of Standard Technical Specifications.

6. Bases Section 2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING: Revise the titles for sections 2.1 and 2.2 to match the changes in Technical Specification Section 2, revise the text to support the changes made to the SAFETY LIMIT curves and add Bases conforming to the guidance of the Standard Technical Specifications to support the action statements relocated from Technical Specification 6.4 to Section 2.2.

Justification: These changes support the changes proposed for Technical Specification Section 2. Technical Specification Section 6.4 did not have Bases.

7. Bases Section 3. LIMITING CONDITIONS FOR OPERATION. Subsections 3.1. REACTOR COOLANT SYSTEM. and 3.4. STEAM AND POWER CONVERSION SYSTEMS: Add statement to each subsection that the applicable safety valves are OPERABLE within i 3% of the nominal setpoint and following testing, the valve lift setting will be restored to within 1% of nominal setpoint.

Justification: These changes support the changes proposed for Technical Specification Subsections 3.4 and 4.1.

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SAFETY EVALUATION Setooint Tolerance Chances Northern States Power Company proposes revising the Prairie Island Technical Specification to allow relaxation in the Pressurizer Safety Valve and Main Steam Safety Valve setpoint tolerances to i 3% for operability considerations and ASME Section XI testing acceptance criteria. The proposed Technical Specification changes require that the Pressurizer Safety Valve and Main Steam Safety Valve setpoints be restored to within nominal 1% of their setpoints following testing.

Prairie Island over-pressure protection design for the Reactor Coolant System incorporates two Code safety valves on the system pressurizer. Over-pressure protection for the main steam system is provided by ten Code safety valves, five on each line. The Pressurizer Safety Valves and Main Steam Safety Valves are tested to verify that their lift pressures and seat leakages are acceptable pursuant to the Prairie Island Inservice Testing Program which complies with the ASME Boiler and Pressure Vessel Code,Section XI, 19A9 Edition. However, the testing acceptance criteria has been established at i 1% lift pressure tolerance in conformance with the Prairie Island Technical Specification for these valves.

The 1989 Edition of ASME,Section XI, requires that both the Pressurizer Safety Valves and Main Steam Safety Valves be tested in accordance with ASME/ ANSI OM-1987, Part 1, " Requirements for Inservice Performance Testing of Nuclear Power Plant Pressure Relief Devices." This standard allows the tested lift pressure to exceed the stamped pressure by up to i 3% before declaring a test failure. It also provides a guideline for testing additional valves when a valve exceeds the 1 3% tolerance. Therefore, increasing the Pressurizer Safety Valve and Main Steam Safety Valve setpoint tolerance to 3% for testing acceptance criteria is in compliance with the later ASME Section XI requirements.

It is important to remember that proposed 3% tolerance is the "as-found" acceptance criteria for valve OPERABILITY or testing.

The requested Technical Specification revisions require that the Pressurizer Safety Valve and Main Steam Safety Valve setpoints be restored to within i 1% of their nominal setpoints following testing.

The impact of the relaxed Pressurizer Safety Valve and Main Steam Safety Valve setpoint tolerance on plant transients and accidents has been considered as follows:

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Pressurizer Safety Valve over-Pressure Protection The Pressurizer Safety Valve, in conjunction with the Reactor Protection System, provides over-pressure protection for the Reactor Coolant System. The Pressurizer Safety Valves are designed to prevent the system pressure from exceeding the system SAFETY LIMIT of 2735 psig which is 110% of the Reactor Coolant System design pressure. The nominal setting of the Pressurizer Safety Valve is 2485 psig with a tolerance of i 1%.

Under the provisions of this license amendment request, the Pressurizer Safety Valve would continue to be OPERABLE and would meet test acceptance criteria if the setpoint is to i 3% of nominal. By the nature of their design, safety valves may require an additional increase in pressure from the initial lift pressure to fully open the valves. ASME Code allows up to 3% additional pressure which is referred to as a 3% accumulation. Accounting for 3% upward setpoint variance from nominal and 3% accumulation the Pressurizer Safety Valve will limit the Reactor Coolant System pressure to 2637 psig which is well below the 2735 psig SAFETY LIMIT.

The relaxed tolerance on Pressurizer Safety Valve lift settings could impact the stress on the relief header and associated piping. These additional stresses would result from the increased jetting forces of steam and water released through the valve at higher pressures and velocities. Each manifold comprises one safety valve and relief header which has been analyzed to assure the stresses are within code allowables. In support of this license amendment request, the pressurizer relief header has been reanalyzed for stresses resulting from a setpoint tolerance of 5% of the nominal setpoint with an additional 3% accumulation.

This analysis concluded the stresses in the relief header and associated piping comply with the original design code ANSI B31.1.0 - 1967 including consideration of two phase flow in accordance with more recent guidance. Therefore, the higher tolerance on the Pressurizer Safety Valves will not result in stresses in excess of design code allowables.

Motor operated valves in the Reactor Coolant System, which are part of the Motor Operated Valve Program as required by Nuclear Regulatory Commission Generic Letter 89-10 are analyzed for a maximum differential pressure across the valve. This differential corresponds to the system pressure the valve experiences in accordance with the requirements of the Program. The applicable valves that are dependent on the Pressurizer Safety Valves for differential pressure analysis have been analyzed assuming the Reactor Coolant System maximum pressure is at the Pressurizer Safety Valve lift setpoint plus a setpoint tolerance of + 3%.

This analysis demonstrated the valves will perform as required with the Pressurizer Safety Valves controlling pressure at this higher level.

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- - -- - - . . -- . . . . . . . - ~- - . . - - . _

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'I' Pressurizar Safety Valve' Transient and Accident Analyses  ;

Transients and accidents that could potentially result in over-  !

pressurization include: l

a. Uncontrolled rod withdrawal from full power  ;

.b. Loss of reactor coolant flow i

c. Loss of external electrical load i

d. Loss of normal feedwater i

e. Loss of all AC power to station auxiliaries

f. Locked RCP rotor Northern States Power Company's Nuclear Analysis and Design group i has performed the reload safety evaluations for the current fuel '

cycles, PIl-17 and PI2-16, and subsequent Prairie Island fuel cycles using a i 3% tolerance on the initial lift pressure and a 3% accumulation for the Main Steam and Pressurizer Safety Valves. l The reload safety evaluations considered all of the transients and accidents listed above and determined that each event -

satisfies the acceptance criteria with the Pressurizer Safety i Valves initial lift pressure +3% above its' nominal setpoint.

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In addition to the Pressurizer Safety Valves, the pressurizer :Ls  ;

also equipped with power operated relief valves (PORV) which  !

automatically open at 2335 psig. If the PORV lift setting is 34 l high it would open at 2405 psig which is below the Pressurizar Safety Valve lift setting if it is 3% low at 2410 psig. Therefore the PORV will always open before Pressurizer Safety Valve and will determine the Reactor Coolant System low pressure DNB limiting transients. DNB evaluations conservatively assume PORV operation because lower Reactor Coolant System pressures yield more limiting values of DNB ratios. Consequently, the proposed Pressurizer Safety valve lift setting tolerance of 3% is not a <

consideration in the DNB transients. As another assurance, the reactor will trip at 2385 psig, ensuring that Pressurizer Safety Valve lift settings 3% below nominal do not enter into the  !

consideration for P"E ,sansients. l In LOCA events, the MO+: tor Coolant System is losing inventory, and pressure is eith?c stable or decreasing. Therefore the Pressurizer Safety Viltes are not challenged in LOCA events and the proposed Pressuri pt Safety Valve setpoint tolerance does not affect the LOCA analysos.

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Pressurizer Safety Valve Conclusions The plant response to transient and accident conditions was considered assuming the Pressurizer Safety Valves lift pressure settings range from nominal setpoints by i 3%. It is concluded ,

that the setpoint lift setting range as requested by this amendment does not adversely affect public health and safety. The amendment does, however, bring plant operation and inspection requirements into conformance with industry expectations for performance of these valves and reduces administrative burdens imposed on the Nuclear Regulatory Commission and Northern States Power Company.

Main Steam Safety Valve Over-Pressure Protection l The primary purpose of the Main Steam Safety Valves is to provide over-pressure protection for the secondary system. The Main Steam i Safety Valves also provide protection against over pressurizing the Reactor Coolant System pressure boundary by providing a heat sink for the removal of energy from the Reactor Coolant System if the preferred heat sink, provided by the condenser and i circulating water system, is not available.

Five Main Steam Safety Valves are located on a header on each main steam line, outside of containment, upstream of the main steam isolation valves. The Main Steam Safety Valve capacity  :

criteria is 110% of rated steam flow at 110% of the steam l generator design pressure. The Main Steam Safety Valve design includes staggered setpoints of 1077, 1093, 1110, 1120 and 1131 psig so that only the needed valv?s will actuate. Staggered setpoints reduce the potential for valve chattering that is due to steam pressure insufficient to fully open all valves following a turbine reactor trip.

The Main Steam Safety Valves are required to maintain system pressure within 1195 psig which is 110% of the main steam system design pressure of 1085 psig. The highest Main Steam Safety Valve nominal lift pressure with 3% upward tolerance is 1165 psig. If a 3% linear accumulation is applied to this highest lift setting,  ;

the maximum steam line pressure is 1184 psig with loss of all I heat sink at rated reactor thermal power. Thus the relaxed lift setting tolerance of 3% will enable the Main Steam Safety Valves to maintain Main Steam system pressure below 1195 psig.

Main Steam Safety Valve Transient and Accident Analyses Many Main Steam and Reactor Coolant System events are potentially impacted by the proposed setpoint tolerance relaxation. These events include DNB transients, long term heat removal incidents, loss of external load such as a turbine trip, small and large break LOCAs and steam generator tube rupture.

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Northern States Power Company's Nuclear Analysis and Design group has performed the reload safety evaluations for the current fuel cycles, PIl-17 and PI2-16, and subsequent Prairie Island fuel cycles using a 3% tolerance on the initial lift pressure and a 3% accumulation for the Main Steam and Pressurizer Safety Valves.

The reload safety evaluations considered all of the transients previously evaluated in the USAR Chapter 14, including those events listed above, and determined that each event meets the acceptance criteria with the Main Steam Safety Valves initial lift pressure +3% above its nominal setpoint.

Lift pressure 3% below nominal setpoint was not considered because a main steam system pressure 3% below the lowest Main Steam Safety Valve setpoint is not a concern in the safety analyses and it does not significantly affect the primary system in the safety analyses.

Main Steam Safety Valve Impact on Auxilliary Feedwater(AFW) Flow Changing the Main Steam Safety Valve lift setpoints also will potentially change the post-transient AFW flows to the steam generators due to higher or lower steam system pressures. Higher pressures will always be limiting since this causes reduced AFW flow. In cases where increased AFW flow (lower steam generator pressure) is limiting, the steam generator is considered faulted (depressurizing) and at pressures well below any Main Steam Safety Valve lift setpoints. The limiting transients analyzed were:

a. Loss of main feedwater (associated with loss of offsite power)

b. Rupture of main feedwater pipe

c. Rupture of a main steam pipe inside containment The Loss of Main Feedwater event establishes minimum flows to the intact steam generator. This analysis has been completed assuming an AFW pump minimum flow of 200 gpm. Plant AFW pump discharge pipe sizing design calculations determined that steam generator pressures less than 1180 psig will result in a minimum 200 gpm AFW pump flow. Typically, the steam generator will be at lower pressures relieving on the Main Steam PORVs at 1050 psig or less.

In these accident scenarios, the reactor trips early in the event, so the over-pressure protection devices are not challenged to their full capacity but rather would act only to limit system pressure. In this role, not all of the Main Steam Safety Valves would be required to perform their function nor would all of them be required to open fully to the extent of their accumulation to perform their pressure limiting function.

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Thus the highest lift pressure setting of 1165 psig (1131 + 3%)

will maintain the steam generator pressure sufficiently low to allow the minimum required AFW flow.

The Rupture of Main Feedwater Pipe and Rupture of a Main Steam Line Inside Containment events also were analyzed assuming that 200 GPM of AFW flow are required. In all three events the AFW pump will perform as required with the Main Steam Safety Valve setpoint tolerance proposed by this license amendment request.

Main steam safety valve lift settings below their nominal setpoints will not adversely affect AFW pump operation since they are designed to supply 200 gpm with steam pressures of 400 psig which is well below the the lowest allowable safety valve setting in conjunction with 10% blowdown.

Other Main Steam Safety Valve Considerations Motor operated valves in the Main Steam system, which are included in the Prairie Island Motor Operated Valve Program are analyzed in accordance with the Program for a maximum differential pressure corresponding to the maximum system pressure the valves will experience. The system valves included in the Program have been reanalyzed in accordance with the Program using i 3% setpoint tolerance in lieu of i 1%. The new analyses concluded that increasing the Main Steam Safety Valve lift setpoint tolerance to i 3% does not affect the operability of the motor operated valves.

The relaxed tolerance on Main Steam Safety Valve lift settings may impact the stresses on the relief header and associated piping. These additional stresses would result from tha increased jetting forces of steam released through the valves at higher pressures. Each manifold consists of five safety valves which have been analyzed in a limiting sequence of operation to assure the manifold stresses are within code allowables in any sequence of operation. The Main Steam relief header has been l reanalyzed for stresses resulting from a setpoint tolerance of +

5% of the nominal setpoint with an additional 3% accumulation.

This calculation concluded the stresses in the manifold and associated piping comply with the original design code ANSI B31.1.0 - 1967. Therefore, the higher tolerance on the Main Steam Safety Valves will not result in any stresses outside the design code assuming a worst case lifting sequence.

j Relaxing the Main Steam Safety Valve lift setting tolerance I raises the possibility that the safety valve may lift in a sequence different than originally assumed which could cause Page 10

l asymmetric loading stresses on the relief header in excess of those previously evaluated. Fluor Daniel Incorporated performed a calculation assuming the most limiting valve lift sequence and determined that the relief header stresses are within Code allowable.

Offsite Dose Considerations The only event identified with potential offsite dose impacted by the relaxation of safety valve tolerances was the steam generator tube rupture accident. However, the analyses for this accident assume that all releases are made by atmospheric steam dump from the affected steam generator. Since the steam dump opens at 1005 psig the opening time for the steam dump will envelope the opening time for Main Steam Safety Valves at a 3% lower lifting pressure. Thus, the existing analyses in USAR 14.5.4 bound this change in the Main Steam Safety Valve setpoint tolerance.

Main Steam Safety Valve Conclusions The plant response to transient and accident conditions was considered assuming the Main Steam Safety Valves lift pressure settings range from nominal setpoints by i 3%. It is concluded that this lift setting range as requested by this amendment does not adversely affect public health and safety. The amendment will, however, bring plant operation and inspection requirements into conformance with industry expectations for performance of these valves and reduce administrative burdens imposed on the Nuclear Regulatory Commission and Northern States Power Company.

Safety Limit Curves The curves of Technical Specification Figure TS.2.1-1 show the loci of points of reactor core temperature differential (an indication of reactor THERMAL POWER), Reactor Coolant System pressure and average Reactor Coolant System temperature for which the minimum DNBR is no less than 1.17 (WRB-1 correlation for Westinghouse fuel) or 1.3 (W-3 correlation for Exxon fuel), as applicable, or the average enthalpy at the reactor vessel exit is equal to the enthalpy of saturated liquid or 650 F whichever is less. If the RPS trip functions are exceeded these curves are used to determine if a SAFETY LIMIT has been exceeded.

The proposed relaxation of the tolerance on Main Steam safety Valve lift settings changes the locus of points at which the Main Steam Safety Valves open on the SAFETY LIMITS curves in Technical Specification Chapter 2. However, the locus of points at which the Main Steam Safety Valves open is not a plant SAFETY LIMIT and serves no purpose being on Figure TS.2.1-1. Therefore this license amendment request proposes removing the curve representing the locus of points at which the Main Steam Safety Page 11

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1 Valves open from the SAFETY LIMITS figure. This is also consistent with the Standard Technical Specifications SAFETY LIMITS curve.

In the course of reviewing the SAFETY LIMIT curves it was determined that the DNB limit seguent of the SAFETY LIMIT curves should be recalculated to account for increased bypass flow associated with removal of fuel assembly thimble plugs and rod bow penalty. Also the Fu limit was increased with the LOCA analysis performed for the current Unit II reload, PI2-16, and subsequent Prairie Island reload analyses which changes the SAFETY LIMIT curves. The combination of these factors were included in a recalculation of the SAFETY LIMIT curves presented in this license amendment request. These changes all contribute to making the SAFETY LIMIT curves more restrictive.

Although the SAFETY LIMIT curves have been revised, plant operation under the old curves did not constitute a safety hazard because plant operation in the affected region of the curve would not have occurred. Normal operation of the Prairie Island units at power is controlled by plant procedures which specify the normal operating band for both Reactor Coolant System temperature and pressure. The normal operating pressure is 2235 psig and the plant procedures require that the Reactor Coolant System pressure be maintained between 2225 and 2245 psig. The normal Reactor Coolant System average temperature at full power is 560 F i 1.5 F in accordance with the T,, Program.

Furthermore, the plant is prevented from approaching the SAFETY LIMIT curves by the Reactor Protection System reactor trips which are designed to prevent the safety limits from being challenged.

These proposed changes to the SAFETY LIMIT curves have not resulted in any changes to the Reactor Protection System (RPS) reactor trip setpoints. These RPS trips are the Over Temperature AT trip, over Power AT trip and the Power Range High Neutron Flux trip. These RPS trips are developed using the criteria of Technical Specification Section 2.3, Limiting Safety System Settings, Protective Instrumentation and do not rely upon the values of the curve in Figure TS.2.1-1. These RPS trips have not been challenged or exceeded at Prairie Island and therefore the safety limits as provided in Figure TS.2.1-1 has not been challenged.

Also note that the core physical limits themselves remain unchanged, that is, the vessel exit temperature is still restricted to 650*F or T sat, whichever is lower and the minimum DNBR is still 1.17 or 1.3 for the WRB-1 (Westinghouse fuel) and W-3 (Exxon fuel) correlations, respectively.

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l Recalculation of the DNB Limit ,

l Northern States Power Company uses approved methodology described I in the " Prairie Island Nuclear Power Plant Reload Safety l Evaluation Methods for Application to Prairie Island Units",

NSPNAD-8102-A, Rev. 6, to calculate DNBR. The DNB limit portions of the safety limit curves were updated because the current bypass flow, rod bow, and Fu assumptions used in calculating DNB  ;

have reduced the DNB margin since the previous revision.

Bypass flow effectively reduces the amount of coolant flow available for transporting heat out of the fuel. Thus, increasing the bypass flow fraction will increase the average enthalpy rise in the fuel channel if power is held constant. Consequently, the margin to DNB is reduced for an increase in bypass flow.

Currently, Northern States Power company assumes a bypass flow fraction of 6% in its safety calculations which is greater than the bypass flow used in previous calculations of the safety limit curves.

The rod bow penalty has also been applied to the limiting DNBR value in the DNB calculations for this submittal. Northern States Power Company currently uses a rod bow penalty of 2.6%. The previous revision of safety limit curves did not have this penalty applied.

The Fu limit was increased from 1.7 to 1.75 which reflects the new peaking factor limits set by the LOCA analysis performed in support of the startup of the current Unit II reload, PI2-16 and applied to subsequent reload analyses. This increased Fu limit reduces DNB margin by allowing higher pin powers.

All three of these factors contributed to reducing the DNB portion of the safety limit curves.

Format Changes l

Northern States Power Company also seeks approval to change the l format of Figure TS.2.1-1. The most significant change is that  :

the Reactor Coolant System average temperature, T ,,, is plotted l against the Reactor Coolant System hot leg minus cold leg temperature difference, AT, instead of percent power. This l format change is required because the full power AT is different  :

I at different temperatures and pressures because the steam tables are nonlinear. This makes it impractical to accurately plot the curves at each temperature and pressure using the same scale for 1 the percent power axis. l 4

other format changes include removing the locus of points at which the steam generator (Main Steam) safety valves open and replacing the curve at 1685 psig with curves at 1785 and 1885 psig. The locus of points at which the steam generator safety Page 13

valvec open has no purpose being on Figure TS.2.1-1 because it is not e. wafety limit. Like the Over-Temperature AT trip and Over-Powet AT trip curves, this curve represents a limiting condition governed by plant equipment which prevents plant operating conditions from approaching the Safety Limite.

Curves at 1785 and 1885 psig were added uo that the entire ~?nge of pressures allowed by TS.2.3.A.2.c is tounded by the curvut of the new plot. The curve at 1685 psig was removed since ".t ic 1 much lower pressure than would ever be achieved in plan; operation dun to the Technical Specification lirmits of the low pressurizer pressure trip setpoint.

These proposed changes to the safety limit curves de r.ot alter the design, function or operation of the Prairia Island plant. It is concluded that they do not adversely affect public health and safety. They will, however, increase the conservatism of plant evaluations in the unlikely event the plant limiting safety system settings are violated.

DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS The proposed changes to the Operating License have been evaluated to determine whether they constitute a significant hazards consideration as required by 10 CFR Part 50, Section 50.91 using the standards provided in Section 50.92. This analysis is provided below:

1. The proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated The proposed changes lacrease the "as-found" setpoint tolerances for the Pressurizer Safety Valves and Main Steam Safety Valves from 1% to i 3%. The proposed changes do not involve any hardware modifications to plant structures, systems, or components. Analyses have determined that the proposed changes do not significantly affect the structural integrity of either the Reactor Coolant System or the Main Steam system.

The proposed setpoint tolerance of i 3% was included in the assumptions for the performance of the reload safety evaluations for the current fuel cycles, PIl-17 and PI2-16, and subsequent Prairie Island fuel cycle analyses.

These analyses concluded that the minimum acceptable DNBR is maintained, over-pressure protection is maintained, LOCA acceptance criteria are met and offsite dose limits are not exceeded. These changes are consistent with the guidance provi.ded by Section III and XI of the ASME Code and Standard Technical Specifications.

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l The proposed change to Technical Specification Figure l TS.2.1-1 does not affect any existing accident analyses. l This revision ensures that the design bases and safety l limits are accurately and appropriately reflected in the  !

i Technical Specifications and will ensure that plant operations are properly evaluated for DNBR encroachment.

Therefore, the probability or consequences of an accident l previously evaluated are not affected by any of the proposed amendments.

2. The proposed amendment will not create the possibility of a new or different kind of accident from any accident Dreviousiv analyzed The lift setpoint the Pressurizer Safety Valves and Main i Steam Safety Valves will be restored to i 1% following testing, thus the "as-left" setpoint tolerance for the Pressurizer Safety Valves and Main Steam Safety Valves are unchanged. Evaluations of plant normal operation, ,

transient and accident conditions have been performed  !

assuming these safety valve lift settings are i 3% of the nominal setpoint and demonstrated that new or different kinds of accidents are not created by the proposed changes. ]

The proposed changes to Technical Specification Figure TS.2.1-1 do not affect the design, function or operatie of any Prairie Island structures, systems or componen' The curves show the loci of points of reactor core differential temperature (an indication of thermal power), ,

I Reactor Coolant System pressure, and average temperature for which the minimum DNBR is not less than the safety analysis limit, that fuel centerline temperature remains below melting, that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, or that the exit quality is within the limits defined by the applicable DNBR correlation. There are no new failure  ;

modes introduced by the proposed changes to the Figure.

The changes conservatively adjust Figure TS.2.1-1 to current plant conditions and ensure that the design is accurately reflected and that the plant is operated in ace rdance with its design bases.

T* are, the possibility of a new or different kind of accident from any accident previously evaluated would not j be created be these amendments.  ;

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3. The proposed amendment will not involve a significant reduction in the marain of safety The proposed changes to the safety valve lift setting j Page 15 l

tolerances are consistent with the guidance provided by Section III.and XI of the ASME Code and Standard Technical Specifications. Analyses have demonstrated these valves will continue to perform their function of protecting their respective system from over-pressurization under all postulated transients and accidents. The changed setting tolerances do not cause a reduction'in any other safety margin such as DNBR.

1 SAFETY LIMIT curves are provided to define minimum l allowable safety margin for plant steady state operation,  !

normal operational transients and anticipated operational )

occurrences. The SAFETY LIMITS represent a design  ;

requirement for establishment of many of the RPS trip _  ;

setpoints which prevent reactor conditions from i approaching the SAFETY LIMITS. The proposed revision of J the SAFETY LIMIT curves provide the minimum. safety margins with somewhat more conservatism than previously included.  ;

No RPS trips setpoints are changed.  :

Therefore, a significant reduction in the margin of safety  !

would not be involved with these amendments.

Based on the evaluation described above, and pursuant to 10 CFR  !

Part 50, Section 50.91, Northern States Power Company has i determined that operation the Prairie Island Nuclear; Generating r Plant in accordance with the proposed license. amendment request does not involve any significant hazards considerations as  ;

defined by Nuclear Regulatory Commission regulations in 10 CFR Part 50, Section 50.92. ,

ENVIRONMENTAL ASSESSMENT l Northern States Power Company has evaluated the proposed changes and determined that: i

1. The changes do not involve a significant hazards  ;

consideration, or l;

2. The changes do not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or

~i

3. The changes do not involve a significant increase in individual or cumulative occupational radiation exposure.

{

Accordingly, the proposed changes meet the eligibility criterion l for categorical exclusion set forth in 10 CFR Part 51 Section l 51.22 (c) (9) . Therefore, pursuant to 10 CFR Part 51 Section  :

51.22(b), an environmental assessment of the proposed changes is not required.

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