ML20247F916

Amends 135 & 127 to Licenses DPR-42 & DPR-60,respectively, Updating TS Heatup & Cooldown Rate Curves & Extending Reactor Vessel Fluence Limit from Current 20 Effective Full Power Yrs to New Value of 35 EFPY

ML20247F916
Person / Time
Site:	Prairie Island
Issue date:	05/04/1998
From:	Wetzel B NRC (Affiliation Not Assigned)
To:
Shared Package
ML20247F921	List: ML20247F916 ML20247F955
References
NUDOCS 9805200089
Download: ML20247F916 (21)

Text

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i UNITED STATES g

j NUCLEAR REGULATORY COMMISSION o

2 WASHINGTCN, D.C. 20sS64001 NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.135 License No. DPR-42 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Northern States Power Company (the licensee) dated March 6,1998, as supplemented March 30, 31, and April 13, 1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfMHf.

j i

- 2.

Accordingly, the license is amended to approve the relocation of certain Technical l

Specification requirements to licensee-controlled documents, as described in the j

licensee's application dated March 6,1998, as supplemented March 30,31, and j

April 13,1998, and evaluated in the staffs safety evaluatir:. dated May 4,1998.

This license is also hereby amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR 42 is hereby amended to read as follows:

9905200089 990504 PDR ADOCK 05000292 P

PM

l Technical Specifications The Technical Specifications contained in Appendix A, as revised through l

Amendment No. 135

, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

- This license amendment is effective as of the date of issuance, with full implementation -

within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION h.

Beth A. Wetzel, Senior P oject Manager Project Directorate ill-1 Division of Reachr Projects -lil/IV Office of Nuclers teactor Regulation

Attachment:

Changes to the Technical Specifications j

Date of issuance: hy 4,1998

)

1 i

ATTACHMENT TO LICENSE AMENDMENT NO.135 FACILITY OPERATING LICENSE NO. DPR-42 l

l DOCKET NO. 50-282 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

REMOVE INSERT TS-xiii TS-xiii TS.1-4 TS.1-4 TS.3.1-2 TS.3.1-2 TS.3.1-4 TS.3.1-4 TS.3.1-5 TS.3.1-5 TS.3.1-6 TS.3.1-6 Figure TS.3.1-1 Figure TS.3.1-2 TS.3.3-1 TS.3.3-1 TS.3.3-3 TS.3.3-3 Table TS.4.1-ic (p. 4 of 4)

Table TS.4.1-1c (p. 4 of 4) f TS.6.7-4 TS.6.7-4 j

B.3.1-3 B.3.1-3 B.3.1-5 B.3.1-5 i

B.3.1-6 B.3.1-6 B.3.3-2 B.3.3-2 B.3.3-2a

_A

TS-xiii l

APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES l

TS FIGURE TITLE l

2.1-1 Reactor Core Safety Limits 3.1-3 DOSE EQUIVALENT I-131~ Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I-131 3.8-1

. Spent Fuel Pool Unrestricted Region Burnup and Decay Time Requirements - 0FA Fuel 3.8-2 Spent Fuel Pool Unrestricted Region Burnup and Decay Time Requirements - STD Fuel 3.10-1 Required Shutdown Margin Vs Reactor Boron Concentration 4.4-1 Shield Building Design In-Laakage Rate 5.6-1 Spent Fuel Fool Burned / Fresh Checkerboard Cell Layout 5.6-2 Spent Fuel Pool Checkerboard Interface Requirements 5.6-3 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - OFA Fuel, No GAD i

j 5.6 4 Spent Fuel Pool Checkerboard Region Burnup and Decay Time i

Requirements - STD Fuel. No GAD 5.6 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - 0FA Fuel, 4 CAD 5.6-6 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel, 4 CAD 5.6-7 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - 0FA Fuel, 8 GAD 5.6-8 Spent Fuel Pool Checkerboard Region Burnup and Decay Time ka uirements - STD Fuel, 8 GAD 5.6-9 Spect, Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - 0FA Fuel, 12 GAD 5.6-10 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel, 12 CAD 5.6-11 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - 0FA Fuel. 16 or More GAD 5.6-12 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements.- STD Fuel, 16 or More GAD B.2.1-1 Origin of Safety Limit curves at 2235 psig with delta-T Trips and Locus of Reactor Conditions at which SG Safety Valves Open l

f.'

Prairie Island Unit 1 Amendment No. H3,129,135 Prairie Island Unit 2 116, 121, 127

TS.1-4 I

OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s).

Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

When a system. subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided:

(1) its corresponding normal or emorgency power source is OPERABLE; and (2) all of its redundant system (s),

subsystem (s), train (s), component (s) and device (s) are OPERABLE, or likewise i

satisfy the requirements of this paragraph.

The OPERABILITY of a system or component shall be considered to be estab-lished when:

(1) it satisfies the Limiting Conditions for Operation in Specification 3.0, (2) it has been tested periodically in accordance with Specification 4.0 and has met its performance requirements, and (3) its condition is consistent with the two paragraphs above.

OPERATIONAL MODE - MODE An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table TS.1.1.

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental characteristics of the core and related instrumentation.

PHYSICS TESTS are conducted such that the core power is sufficiently reduced to allow for the perturbation due to the test and therefore avoid exceeding power distribution limits in Specification 3.10.B.

Low power PHYSICS TESTS are run at reactor powers less than 24 of rated power.

PRESSURE AND TEMPERATURE LIMITS REPORT (PTIJO The PTLR is the document thct provides reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period.

These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 6.7.A.7.

Plant operation within these operating limits is addressed in the individual specifications.

I Prairie Island Unit 1 Amendment No.

111, 12f, 135 Prairie Island Unit 2 194*, LF5,127

a I

(

TS.3.1-2 3.1.A.l.c.

Reactor Coolant System Average Temperature Below 350*F (and Reactor Coolant Level Above the Reactor Vessel Flante)

(1) Whenever the reactor coolant system average temperature is below 350*F, except during REFUELING, at least two methods for removing decay heat shall be OPERABLE with one in operation * (except as specified in 3.1.A.l.c.(2) below).

Acceptable methods for removing decay heat are at least one reactor coolant pump and its associated steam generator; or a residual heat removal loop including a pump and its associated heat exchanger.

(2) With only one OPERABLE method of removing decay heat, initiate prompt action to restore two OPERABLE methods of removing decay heat.

If the remaining operable method is an RHR loop, be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(3) With no OPERABLE methods of removing decay heat, suspend all operations involving a reduction in boron concentration of the reactor coolant system and initiate prompt action to restore one OPERABLE method of removing decay heat.

(4) A reactor coolant pump may be started at RCS temperature less than the Over Pressure Protection System Enable Temperature specified in the PTLR, only if either of the following conditions is met:

There is a steam or gas bubble in the pressurizer, or The (steam generator minus RCS) temperature difference for the steam generator in that loop is less than 50*F.

d.

Reactor Coolant Level Below or at the Reactor Versel Flance (1) Both residual heat removal loops, each consisting of a pump and its associated hect exchanger, shall be OPERABLE with one in operation * (except as specified in 3.1.A.1.d.(2) below).

(2) With one or both residual heat removal loop (s) inoperable, prompt action shall be taken to restore the inoperable

~

residual heat removal loop (s) to an OPERABLE status.

During reduced inventory conditions, a safety injection pump may be run as required to maintain adequate core cooling and RCS inventory in the event of a loss of Residual Heat Removal System cooling.

• All pumps may be shutdown for up to one hour provided the reactor is suberitical, no operations are permitted that would cause dilution of the reactor coolant boron concentration and core outlet temperature is main-tained at least 10*F below saturation temperature.

Prairie Island Unit 1 Amendment No. )9", 9(, 135 Prairie Island Unit 2

$8', $4T 127

TS.3.1 4 3.1.A.2.c Pressurizer Power Ooerated Relief Valves (1) Reactor Coolant System average temperature greater than or equal to 350*F l

(a)

Reactor coolant system average temperature shall not exceed 350*F, unless two power operated relief valves (PORVs) and l

their associated block valves are OPERABLE (except as specified in 3.1.A.2.c(1)(b) below).

(b)

During STARTUP OPERATION or POWER OPERATION, any one of the following conditions of inoperability may exist for each unit.

If OP N ILITY is not restored within the time specified or the required action cannot be completed, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce reactor coolant system average temperature below 350*F, within the following 6 l

hours.

1. With one or both PORVs inoperable because of excessive seat leakage, within one hour either restore the PORV(s) to OPFMLE status or close the associated block valve (s) with power maintained to the block valve (s).

t

2. With one PORV inoperable due to causes other than excessive seat leakage, within one hour either restore the PORV to OPERABLE status or close and remove power from the associated block valve. Restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

3. With both PORVs inoperable due to causes other than excessive seat leakage, within one hour either restore at least one PORV to OPERABLE status or close and remove power from the associated block valves and be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce reactor coolant system average temperature below 350*F, within the following l

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

4. With one block valve inoperable, within one hour either restore the block valve to OPERABLE status or place its associated PORV in manual control. Restore the block valve to OPERABLE status within the.following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

5. With both block valves inoperable, within one hour either restore the block valves to OPERABLE status or place the P0kVs in manual control. Restore at least one block valve to OPERABLE status within the next hour.

1 (2)

Reactor Coolant System average temperature greater than or equal to the temperature specified in the PTLR for disabling both safety injection pumps and below the Over Pressure Protection System Knable Temperature soecified in the PTLR i

With Reactor Coolant System temperature greater than or equal to the temperature specified in the PTLR for disabling both safety injection pumps and less than the Over Pressure Protection System Enable Temperature specified in the PTLR; both pressurizer power operated relief valves (PORVs) shall be OPERABM (except as specified in 3.1.A.2.c.(2).(a) and 3.1.A.2.c.(2).(b) below) with the Over Pressure Protection System enabled, the associated block valve open, and the associated backup air supply charged.

Prairie Island Unit 1 Amendment No. ff, 196, 135 Prairie Island Unit 2 S4', 99", 127 L

TS.3.1-5 3.1.A.2.c.(2).(c) Ona PORV may bn in:p3reble for 7 days.

If thsca ccnditiens cannot be met, depressurize and vent the reactor coolant system 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 />.

(b) With both PORVs inoperable, complete depressurization and venting of the RCS 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 />.

.(3) Reactor Coolant System average temperature below the temperature snecified in the PTLR for disabline both safety iniection numns With Reactor Coolant System te g erature less than the temperature specified in the PTLR for disabling both safety injection pumps, when the head is on the reactor vessel and the reactor coolant system is not vented through a 3 square inch or larger vent; both Pressurizer power operated relief valves (PORVs) shall be OPERABLE (except as specified in 3.1.A.2.c.(3).(a) and 3.1.A.2.c.(3).(b) below) with the Over Pressure Protection System enabled, the associated block valve open, and the associated backup air supply charged.

(a) 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,.depressurize and vent the reactor coolant system 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 />.

(b) With both PORVs inoperable, complete depressurization and venting of the RCS 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 />.

3.1.A.3 Reactor Coolant Vent System A reactor shall not be made or maintained critical nor shall a.

reactor coolant system average temperature exceed 200*F unless Reactor Coolant Vent System paths from both the reactor vessel head and pressurizer steam space are OPERABLE and closed (except as specified in 3.1.A.3.b and 3.1.A.3.c below).

b.

During STARTUP OPERATION and POWER OPERATION, any one of the following conditions of inoperability may exist for each unit provided STARTUP OPERATION is discontinued until OPERABILITY is restored.

If any one of these conditions is not restored to an OPERABLE status within 30 days, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />s:

(1) Both of the parallel vent valves in the reactor vessel head vent path inoperable, or (2) Both of the parallel vent valves in the pressurizer vent path inoperable, or (3) The vent valve to the pressurizer relief tank discharge line inoperable, or (4) The vent valve to the containment atmospheric discharge line inoperable, With no Reactor Coolant Vent System path OPERABLE, restore at c.

least one vent path to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Prairie Island Unit 1 Amendment No. %,106,135 Prairie Island Unit 2 K,f9C 127

TS.3.1-6 i

3.1.B.

Pressure /Temocrature Limits l

1.

Reactor Coolant System The Unit 1 and Unit 2 Reactor Coolant Systems (except the a.

pressurizer) temperature, pressure, heatup rates, and cooldown rates si.all be maintained within the limits specified in the Pressure and Temperature Limits Report (PTLR),

b.

If these conditions cannot be satisfied, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the reactor coolant system average temperature and pressure to less than 200*F and 500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

2.

Pressurizer The pressurizer temperature shall be limited to:

a.

1.

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

2.

A maximum cooldown of 200*F in any 1-hour period.

b.

The pressurizer spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320*F.

c.

If these conditions cannot be.catisfied, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the pressurizer than 500 psig within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. pressure to less l

I l

l I

i Prairie Island Unit 1 Amendment No. yd,Pf,135 Prairie Island Unit 2 JMTI)WI,127 j

Ts.3.3 1 3.3 ENGINEERED SAFETY FEATURES Annlicability Applies to the operating status of the engineered safety features.

Obiective To define those limiting conditions that are necessary for operation of engineered safety features:

(1) to remove decay heat from the core in an emergency or normal shutdown situations, and (2) to remove heat from containment in normal operatir.g and emergency situations.

Specifications A.

Safety Iniection and Residual Meat Removal systems 1

A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 200'F unless the following conditions are~ satisfied (except as specified in 3.3.A.2 below):

The refueling water tank contains not less than 200,000 a.

gallons of water with a boron concentration of at least 2500 ppa.

b.

Each reactor coolant system accumulator shall be OPERABLE when reactor coolant system pressure is greater than 1000 psig.

OPERABILITY requires:

(1) The isolation valve is open (2) volume is 1270 20 cubic feet of borated water (3) A minimum boron concentration of 1900 ppm (4) A nitrogen cover pressure of 740 1 30 psig Two safety injection pumps are OPERABLE except as specified in c.

Sections 3.3.A.3 and 3.3.A.4.

d.

Two residual heat removal pumps are OPERABLE.

i Two residual heat exchangers are OPERABLE.

e.

1 Prairie Island Unit 1 Prairie Island Unit 2 Amendment No. 9(, L96, 135 S47 LDf, 127

75.3.3 3 3.3.A.2.g.

The valve position acnitor lights or alarms for motor-operated valves specified in 3.3.A.13 above may be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided the valve position is verified once each shift.

3.

A paximum of one safety injection pump shall be capable of l

injecting into the RCS whenever RCS temperature is less than the Over Pressure Protection System Enable Temperature specified in the PTLR except that both SI pumps may be run for up to one hour while conducting the integrated SI test ** when either of the following conditions is met:

(a)

There it a steam or gas bubble in the pressurizer and an isolation valve between the SI pump and the RCS is shut, or (b)

The reactor vessel head is removed.

4 No safety injection pumps *** shall be capabla of injecting into the RCS whenever RCS temperature is less than the temperature specified in the PILR for disabling both safety injection pumps (except one or both pumps may be run as specified in 3.3.A.3 and 3.1.A.1.d.(2)).

\\

5.

Both reactor coolant system accumulators shall be isolated

• whenever RCS temperature is less than the Over Pressure Protsetion System Enable Temperature specified in the PTLA.

e

• This specification does not apply whenever the reactor coolant system accumulators are depressurized or the reactor vessel head is removed.

• • 0ther 51 system tests and operations may also be conducted undar these conditions.

• • • This specification does not apply whenever the reactor vessel head is removed.

Prairie Island Unit 1 Amendment No.,9f, 42T, 135 Prairie Island Unit 2

$47 1,k97127 I

e

TS.6.7-4 WCAP-10054-P-A, " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code", August, 1985 WCAP-10924-P-A, "Westin Methodology", December,ghouse Large-Break LOCA Best-Estimate 1988 WCAP-10924-P-A, Volume 1, Addendum 4, " Westinghouse Large Break LOCA Best Estimate Methodology", August, 1990 XN NF-77-57 (A), XN-NF-77-57, Supplement 1 (A), " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase II", May, 1981 WCAP-13677, "10 CFR 50.46 Evaluation Model Report: E-COBRA / TRAC 2-Loop Upper Plenum Injection Model Update to Support ZIRLOm Cladding Options", April 1993 (approved by NRC SE dated November 26, 1993).

NSPNAD-93003-A, " Transient Power Distribution Methodology",(latest approved version)

The core operating (e.mits shall be determined such that all li c.

applicable limits g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

d.

The CORE OPERATINC LIMITS REPORT, including any mid-cycle revisions l

or supplements thereto, shall be supplied upon issuance, for each reload cycle, to-the 1RC Document Control Desk with copies to the i

Regional Administrator and Resident Inspector.

6.7.A.7 Pressure and Temocrature Limits Report (PTLR)

RCS pressure and temperature limits for heatup, cooldown, low a.

temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following Technical Specification sections; 3.1.A.1.c(4), 3.1.A.2.c(2), 3.1.A.2.c(3), 3.1.B.1.a.

3.3.A.3, 3.3.A 4, 3.3.A.5, and Table 4.1-1C.

b.

The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigation System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

WCAP-14040-NP-A, Revision 2. " Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (Including any exemption granted by NRC to ASME Code Case N-514)

The PTLR shall be provided to the NRC upon issuance for each reactor c.

vessel fluence period and for any revision or supplement thereto.

j Changes to the curves, setpoints, or parameters in the PTLR resulting from new or additional analysis of beltline material properties will be submitted to the NRC prior to issuance of an updated PTLR.

B.

REPORTABLE EVENTS The following actions shall be taken for REPORTABLE EVENTS:

The Commission shall be notified by a report submitted pursuant to a.

the requirements of Section 50.73 to 10 CFR Part 50, and b.

Each REPORTABLE EVENT shall be reviewed by the Operations Committee and the results of this review shall be submitted to the Safety Audit Committee and the Vice President Nuclear Generation.

Prairie Island Unit 1 Amendment No. 001, L2f', 135 Prairie Island Unit 2

%.kF5, 127

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B.3.1-3 1

3.1 REACTOR C001 ANT SX$ Igg Bases continued A.

Operational Components (continued) b.

Maintaining the integrity of the reactor coolant pressure boundary.

This is a function that is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage.

I Manual control of the block valve to:

c.

(1) unblock an isolated PORV to allow it to be used for manual control of reactor coolant system pressure (Item a above), and (2) isolate a PORV with excessive seat leakage (Item

b. above).

d.

Manual control of a block valve to isolate a stuck-open PORV.

The OPERABILITY of two PORVs or an RCS vent opening of at least 3 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when the RCS temperature is less than the Over Pressure Protection System Enable Temperature specified in the PTLR.

The PORV control switches are three position switches, Open-Auto-Close.

A PORV is placed in manual control by placing its control switch in the Closed position.

a The RCS safety valves and normal setpoints on the pressurizer.PORV's do not provide overpressure protection for certain low temperature operational transients.

Inadvertent pressurization of the RCS at temperatures below the Over Pressure Protection System Enable Temperature specified in the PTLR could result in the ASME Appendix G brittle fracture pressure / temperature limits specified in the PTLR being exceeded.

The setpoint for the low temperature overpressure protection system is derived by analysis which models the performance of the low temperature overpressure protection system assuming various mass input and heat input transients.

The low temperature overpressure protection system setpoint is updated whenever the RCS heatup and cooldown curves specified in the PTLR are revised.

The 3 square inch RCS vent opening is based on the 2.956 square inch cross sectional flow area of a pressurizer PORV. Because the RCS vent opening specification is based on the flow capacity of a PORV, a PORV maintained in the open position may be utilized to meet the RCS vcnt requirements.

Prairie Island Unit 1 Amendment No.

135 Prairie Island Unit 2

, 127 i

l l

B.3.1-5 3.1 REACTOR COOLANT SYSTEM j

Aaaan continued B.

Pressure /Temocrature Limits Appendix G of 10 CFR Part 50, and the ASME Code require that the reactor coolant pressure boundary be desiped with sufficient margin to insure that, when stressed under operating, maintenance, testing, and postulated accident conditions, the boundary behaves in a nonbrittle manner, the probability of rapidly propagating fracture is minimized and the design reflects the uncertainties in determining the effects of irradiation on material properties.

The pressure / temperature limit curves specified in the PTLR are based on the properties of the most limiting material in either unit's reactor vessel (Unit 1 reactor vessel nozzle to intermediate shell forging circumferential weld) and are effective to 35 EFPY. The curves in the PTLR have not been adjusted for pressure and temperature sensing instruments' uncertainties, he curves incorporated into plant operating procedures will incorporate instrument uncertainties.

The curves define a region where brittle fracture will not occur and are determined from the material characteristics, irradiation effects, pressure stresses and stresses due to thermal gradients across the vessel wall.

Heatuo Curves During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. At the inner wall of the vessel, the thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure, which tends to make the coolant temperature limit higher.

However, the coolant temperature is higher than the metal temperature in the heatup condition, which tends to reduce the coolant temperature limit. These two phenomena tend to cancel each other.

Therefore, an inside-radius pressure-temperature curve based on a comparison of the steady state conditions (i.e., no thermal stresses) and the finite heatup rate conditions must be performed.

The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location.

The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present.

The thermal induced stresses at the outer wall of the vessel are dependent on both the rate of heatup and coolant temperature during the,heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined. Therefore, each heatup rate of interest must be analyzed on an individual basis. The heatup limit curve is a composite curve prepared by determining the most conservative case in a point by point comparison, with either the inside steady state curve, the inside finite heatup rate curve, or the outside finite heatup rate curve, for any heatup rate up to 100*F per hour.

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Prairie Island Unit 1 Amendment No. Pf, 1,06,135 i

Prairie Island Unit 2 S(,)MI,127 l

I

5.3.1-6 3.1 REACTOR COOIANT SYSTEM 13333 (continued)

Cooldown curves During cooldown, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from tensile at the inner wall to compressive at the outer wall The thermal induced tensile stresses at the inner wall are additive to the pressure induced tensile stresses which are already present. Therefore, the controlling location is always the inside wall.

The cooldown-limit curves were prepared utilizing the same type of analysis used to calculate the heatup curve except that the controlling location is always the inside wall.

Limit lines for cooldown rates between those presented may be obtained by interpolation.

Criticality Lf=fts Appendix G of 10 CFR Part 50 requires that for a given pressure, the reactor must not be made critical unless the temperature of the reactor vessel is 40'F above the minimum permissible temperature specified on the heatup curve and above the minimum permissible temperature for the inservice hydrostatic For Prairie Island the curves were prepared, requiring that pressure test.

criticalit must occur above the maximum permissible temperature for the inservice ydrostatic pressure test.

ASME Code Section XI Inservice Test Limits The pressure temperature limits for the ASME Code Section XI ' Inservice Test

{

Limits (hydrostatic pressure test) are less restrictive than the heatup and cooldown curves to allow for the periodic inservice hydrostatic test.

These limits are allowed to be less restrictive because the hydrostatic test is based on a 1.5 safety factor versus the 2.0 safety factor built 1nto the heatup and cooldown curves and because the test is run at a constant temperature so the thermal stresses in the vassal are minimal.

Steam Generator Pressure / Temperature Limitations The limitations on steam generator pressure and temperature ensure that the pressure induced stress in the steam generators do not exceed the maximum allowable fracture toughness stress limits and thus prevent brittle fracture of the steam generator shell.

Pressurizer Limits Although the pressurizer operates at temperature ranges above those for which

'there is reason for concern about brittle fracture, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with ASME Code requirements.

I Prairie Island Unit 1 Amendment No. j(, I M, 135 Prairie Island Unit 2 54 jF, 127

B.3.3-2 3.3 ENCINErarD SAFETY FEATURES 1A111 continued (1) Assuring with high reliability that the safety system will function properly if required to do so.

(2) Allowance of sufficient time to complete required repairs and testing using safe and proper procedures.

Assuming the reactor has been operating at full RATED THERMAL POWER for at least 100 days, the magnitude of the decay heat decreases as follows after initiating HOT SHUTDollN.

Time Af ter Shutdown Decav Hest. t of RATED POWER 1 min.

4.5 30 min.

2. 0' 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1.62 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0.96 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 0.62 3

Thus, the requirement for core cooling in case of a postulated loss-of coolant accident while in the shutdown condition is significantly l

reduced below the requirements for a postulated loss of-coolant acci-dent during POWER OPERATION.

Putting the reactor in the HOT SHUTDOWN condition significantly reduced the potential consequences of a loss-of coolant accident, and also allows more free access to some of the engineered safeguards components in order to effect repairs.

The sceumulator and refueling water tank conditions specified are consistent with those assumed in the LOCA analysis (Reference 2).

Specification 3.3.A.3 allows use of an SI pump to perform operations required at low RCS temperatures; e.g., raising accumulator levels in order to meet the level requirement of Specification 3.3.A.1.b(2) or ASME Section XI tests of the SI system check valves.

Specification 3.3.A.3 also allows use of both SI pumps at low tempera-tures for conduct of the integrated SI test and other SI system tests and operations providing the pumps run for less than i hour. In this case, pressurizer level is maintained at less than 50% and a positive means of isolation is provided between the SI pumps and the RCS to prevent fluid injection into the RCS. This isolation is accomplished by using either a closed manual valve or a closed motor operated valve with the power removed. This combination of conditions under strict administrative control assure that overpressurization cannot occur. The option of having the reactor vessel head removed is allowed since in

. r;nis case RCS overpressurization cannot occur.

Maintaining the safety injection pumps incapable of injecting into the RCS, as specified in 3.3.A.3 and 3.3.A.4, and isolating the accumulators, as specified in 3.3.A.5, will provide assurance that the

. plant operating conditions will be bounded by the assumptions applied to the determination of the OPPS setpoints in the mass injection transient analysis. These setpoints will actuate the PORVs upon an RCS pressure increase to maintain RCS pressure within the acceptable operating region of the pressure / temperature (brittle fracture) limit curves in the PTLR.

The provisions of these specifications are not applicable when the Prairie Island Unit 1 Amendment No. K. 127, 135 Prairie Island Unit 2 84'. Jagr.127

4 3.3.3-2n l

l reactor vessel head is removed since in that condition, RCS overpressurization can not occur.

The safety injection pumps are rendered incapable of injecting into che l

RCS by employing at least two independent means to prevent a pump start such that a single action will not result in an injection into the RCS.

l This may be accomplished through the pump control switch being placed in pullout with a blocking device installed over the control switch that would prevent an unplanned pump start.

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l Prairie Island Unit 1 Amendment No. g% W. 127

,127,135 Prairie Island Unit 2 I

I

oMt2 g"

t UNITED STATES s

j NUCLEAR REGULATORY COMMISSION I

WASHINGTON, D.C. 30866-0001

\\.....f NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING Pt ANT. UNIT ?

AMENDMENT TO FACILITY OPERATING LICENSE i

Amendment No.127 License No. DPR-60 1.

The Nuclear Regulatory Commission (the Commission) has found that:

l A.

The application for amendment by Northern States Power Company (the l

licensee) dated March 6,1998, as supplemented March 30,31, and April 13, 1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; I

D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended to approve the relocation of certain Technical Specification requirements to licensee-controlled documents, as described in the licensee's application dated March 6,1998, as supplemented March 30,31, and April 13,1998, and evaluated in the staff's safety evaluation dated May 4,1998.

This license is also hereby amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-60 is hereby amended to read as follows:

i

2-Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 127

, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of issuance, with full implementation within 30 days.

l FOR THE NUCLEAR REGULATORY COMMISSION Beth A. Wetzel, Senior roject Manager Project Directorate Ill-1 Division of Reactor Projects - I!!/IV Office of Nucleat Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance:

May 4, 1998 i

i ATTACHMENT TO LICENSE AMENDMENT NO. -127 FACILITY OPERATING LICENSE NO. DPR-60 DOCKET NO. 50-306 1

Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

REMOVE INSERT TS-xiii TS-xiii TS.1-4 TS.1-4 TS.3.1-2 TS.3.1-2 TS.3.1-4 TS.3.1-4 TS.3.1-5 TS.3.1-5 TS.3.1-6 TS.3.1-6 Figure TS.3.1-1 Figure TS.3.1-2 TS.3.3-1 TS.3.3-1 TS.3.3-3 TS.3.3-3 Table TS.4.1-1c (p. 4 of 4)

Table TS.4.1-1c (p. 4 of 4) i TS.6.7-4 TS.6.7-4 i

B.3.1-3 B.3.1-3 B.3.1-5 B.3.1-5 B.3.1-6 B.3.1-6 B.3.3-2 B.3.3-2 B.3.3-2a i

l

. _ - - -.