Text

Part 1 of 2, Prairie Island Nuclear Generating Plant, Supplement to License Amendment Request Dated December 11, 2000, Conversion to Improved Technical Specifications (ITS)
	ML020570030
Person / Time
Site:	Prairie Island
Issue date:	02/15/2002
From:	Nazar M; Nuclear Management Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	TSTF-153
	Download: ML020570030 (225)
	v • d • e

NMC-7 Committed to NuclearExcellence Nuclear Management Company, LLC Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. East Welch MN 55089 February 15, 2002 10 CFR Part 50 Section 50.90 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Supplement to License Amendment Request dated December 11, 2000 Conversion to Improved Technical Specifications (ITS)

By letter dated, December 11, 2000, Prairie Island submitted a License Amendment Request (LAR) to convert the current Technical Specifications (CTS) using the guidance of NUREG-1431, Revision 1 as amended by NRC and industry Technical Specification Task Force (TSTF) documents. This letter supplements the subject LAR.

By letter dated December 19, 2001, the NRC Staff sent NMC requests for additional information (RAIs) regarding our LAR dated December 11, 2000 to convert to Improved Technical Specifications. Attachment 1 to this letter contains the NRC RAIs for ITS Section 3.4, "Reactor Coolant System", and the Nuclear Management Company (NMC) answers to these RAIs.

NMC also proposes to make the review changes and corrections, identified as E22, E24, E27, E32 and E33. Changes designated E22 clarify the dual role of containment vacuum breakers as requested by the NRC staff. Changes designated as E24 restore the clause, "to perform preplanned work activities" to Specification 3.4.5 LCO Note. Changes designated E27 delete TSTF-1 53 from Section 3.4 since this TSTF was withdrawn by the NRC. Changes designated as E32 provide the resolution of all remaining issues on Section 3.6, including RAIs 3.6.8-02, 3.6.8-03, 3.6.8-05 and 3.6.8-06. Changes designated as E33 provide resolution of Specification 3.4.12 and 3.4.13 issues raised in meetings with the NRC staff on January 16, 2002.

USNRC NUCLEAR MANAGEMENT COMPANY February 15, 2002 Page 2 of 3 , Page List by RAI Q, provides a cross-reference of RAIs and other sources of page changes to the pages that they changed. to this letter contains Revision 9 change pages which implement answers to Section 3.4 RAIs and the Review Change/Errata changes designated as E22, E24, E27, E32 and E33. Changes to the Revision 9 pages are sidelined in the right margin beside the line(s) which have been revised. Change Pages from Parts A, B, D, F, G or Cross-References are dated 1/2/02. Change Pages from Parts C and E are marked as Revision 9 with a small textbox below the revision sideline which contains "R-9".

The Significant Hazards Determinations and Environmental Assessments, as presented in the original December 11, 2000 submittal and as supplemented March 6, 2001, July 3, 2001, August 13, 2001, November 12, 2001, December 12, 2001, January 25, 2002, January 31, 2002, February 14, 2002, February 15, 2002 (Section 1.0, 2.0, 3.0, 3.1, 3.2, 4.0, E23, E25, E26 and E29 changes) and by the Part G change pages in Attachment 3 of this letter, bound the proposed license amendment.

NMC is notifying the State of Minnesota of this LAR supplement by transmitting a copy of this letter and attachments to the designated State Official.

To the best of my knowledge and belief, the statements contained in this document are true and correct. In some respects these statements are not based on my personal knowledge, but on information furnished by other Prairie Island Nuclear Generating Plant (PINGP) and NMC employees, contractor employees, and/or consultants. Such information has been reviewed in accordance with company practice, and I believe it to be reliable.

In this letter NMC has not made any new or revised any Nuclear Regulatory Commission commitments. Please address any comments or questions regarding this matter to myself or Mr. Dale Vincent at 1-651-388-1121.

Mano K. Nazar Site Vice Presid t Prairie Island N clear Generating Plant (Copies and attachments listed on Page 3)

USNRC NUCLEAR MANAGEMENT COMPANY February 15, 2002 Page 3 of 3 C: Regional Administrator - Region III, NRC Senior Resident Inspector, NRC NRR Project Manager, NRC James Bernstein, State of Minnesota J E Silberg Attachments:

Affidavit

1. NRC RAIs Section 3.4, "Reactor Coolant System", and NMC Responses.

2. Page List by RAI Q

3. Revision 9 Change Pages

UNITED STATES NUCLEAR REGULATORY COMMISSION NUCLEAR MANAGEMENT COMPANY, LLC PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET NO. 50-282 50-306 REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 & DPR-60 SUPPLEMENT TO LICENSE AMENDMENT REQUEST DATED DECEMBER 11, 2000 CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS (ITS)

By letter dated February 15, 2002, Nuclear Management Company, LLC, a Wisconsin corporation, is submitting additional information in support of the License Amendment Request originally submitted December 11, 2000.

This letter contains no restricted or other defense information.

NUCLEAR MANAGEMENT COMPANY, LLC Mano K. Nazar ,//

Site Vice Presid/'t Prairie Island Vclear Generating Plant State of 1,r 'nA ,,/Q County of _____ _____

On this IS A day of f C before me a notary public acting in said County, personally appeared M o K. Nazar, Site Vice President, Prairie Island Nuclear Generating Plant, and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Nuclear Management Company, LLC, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the statements made in it are true.

o r. t,20 IN I MARLYS E.DAVIS NOTARY PUBUG.MINNNESOTA

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ITS Submittal Copies Recipient Letter NRC CD NMC CD Insert Copies NRC DCD 1 1 Tae Kim 1 2 1 3 NRC Ril! 1 1 NRC RI 1 1 Bernstein 1 1 Bruemmer 1 1 Silberg 1 1 Pearce 1 1 Swigart 1 Jantosik 1 1 Northard 1 1 Weinkam 1 1 R. Anderson 1 D Hoffman 1 1 Alders 1 1 PB Lic Mgr 1 1 K Lic Mgr 1 1 Beach 1 1 Fujimoto 1 1 Gillispie 1 1 Larry Davis (Duke Eng.) 1 1 Mike Johnson 1 1 Chris Mundt 1 1 Reddemann 1 1 Solymossy 1 1 Cutter 1 1 Nazar 1 Werner 1 Lingle 1 Albrecht 1 Amundson 1 Williams 1 Jefferson 1 Larimer 1 Eckholt 1 1 Kivi 1 Leveille 1 Vincent 1 1 2 Frost 1 1 1 VanTassell 1 2 Marty (Manifest only)

Hall 1 20 2 PITC 1 Eng Libr 1 Lic Libr 1 NL File 1 TS History 1 1 PI Records 1 1 Betty Underwood (OSRC) I Totals 44 4 46 16

Prairie Island Nuclear Generating Plant Attachment 1 to Supplement dated February 15, 2002 to License Amendment Request dated December 11, 2000 Conversion to Improved Technical Specifications (ITS)

NRC RAIs Section 3.4, "Reactor Coolant System", and NMC Responses

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

1. ITS SR 3.4.1.3 Note X3.4-104 STS SR 3.4.1.4 is modified by a Note that allows entry into MODE 1, without having performed the SR, and placement of the unit in the best condition for performing the SR. The Note states that the SR is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after _>[90%] RTP. ITS SR 3.4.1.3 Note is modified to state "Required to be performed within 7 days after >_ 90% RTP." X3.4-104 states that 7 days "is sufficient time to perform the necessary calculations and allow any potential RCS fluctuations following the startup to stabilize and provide more accurate determinations." Seven days appears to be an extremely long period of time to be operating in MODE 1 without verifying the total RCS flow rate, especially since CL3.4-103 stated that the control board flow meters do not provide sufficient resolution to measure the specified values.

Comment: Maintain the STS wording or provide historical plant data and any risk assessment insights to support the proposed 7 day allowance for meeting the surveillance requirement. As proposed, this is a generic change to NUREG-1431 which requires the Westinghouse Owners Group approval.

NMC Response:

Parts affected by this change:

Part B: Final ITS pages Part C: CTS markup Part D: DOC A3.4-100 Part E: ISTS markup Part F: JFD CL3.4-104 Part G: NSHD ISTS markup for SR 3.4.1.3 has been revised to delete the subject Note. The Note is not in PI CTS and is not consistent with PI CLB or CTS. Not accepting this Note is consistent with the agreements made between the NRC and the Industry. That agreement allows a plant to maintain their CLB and CTS requirements. This applies in this circumstance. PI CTS does require that a verification of the RCS total flow rate be performed after each refueling outage, however there are no specific CTS requirements to perform this verification within a specific time after reaching a specific RTP. Performance of this test is controlled by plant procedures. The CTS has been revised to closer reflect the ISTS markup. As a result, DOC A3.4-1 00 was generated. In exercising the industry agreement, PI is not incorporating the subject Note.

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Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

2. ITS SR 3.4.1.3 PA3.4-106 STS SR 3.4.1.4 requires the verification "by precision heat balance that RCS total flow rate is >_

[284,000] gpm and greater than or equal to the limit specified in the COLR." ITS SR 3.4.1.3 would not specify the method by which the verification of the RCS total flow rate would be performed. PA3.4-106 states that the phrase "by precision heat balance" is not included because the CTS does not specify a particular method for performing RCS flow test. The method of performing the RCS flow test should be included in the ITS SR 3.4.1.3.

Comment: Maintain the STS wording or provide the method of performing the RCS total flow test in ITS SR 3.4.1.3 that is acceptable to the staff.

NMC Response:

Parts affected by this change:

None ISTS SR 3.4.1.3 requires that the RCS total flow rate be determined specifically by a "precision heat balance test." PI did not incorporate this specific method in ITS SR 3.4.1.3. PI uses other methods of verifying RCS total flow rate and does not use the precision heat balance test. This is also true with other utilities. One of the reasons that this specific test is not being incorporated is that if it is specifically stated per the ISTS, then the precision heat balance test would be the only method that could be used in order to meet this specific SR. PI intends to maintain our current testing procedures and practices such as elbow flow taps.

3. CTS 3.10.J CTS 3.10.J title has been changed from "DNB Parameters" to "RCS Pressure, Temperature and Flow DNB Limits." No discussion of change was provided for the change in title. Also, the new title is missing a '-' that was added to ITS 3.4.1.

Comment: Provide a discussion of change for the CTS 3.1 0.J title.

NMC Response:

Parts affected by this change:

Part C: CTS markup The CTS markup has been revised to add a '-' in the following title, "RCS Pressure, Temperature and Flow - DNB Limits." P1 changed the title of CTS Specification 3.10. j from "DNB Parameters" to "RCS Pressure, Temperature and Flow - DNB Limits" to be consistent with 2

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

NUREG-1431. The title change is also consistent with the actions contained in this section.

This section contains RCS pressure, temperature and flow which are DNB limits as the title states. This change is covered under DOC A3.4-00.

4. ITS LCO 3.4.2 ITS SR 3.4.2.1 ITS LCO 3.4.2 and SR 3.4.2.1 have replaced the bracketed 541 OF with 5400F. No JFD was provided for the change.

Comment: Provide the JFD for the proposed change.

NMC Response:

Parts affected by this change:

None Due to rules of conversion, NUREG 1431 values or information contained in brackets can be retained if applicable to the plant or plant specific information may be substituted. This is additionally discussed in Appendix A of the PI submittal. In this particular case, the NUREG uses a RCS loop average temperature of [541] degrees F. PI has replaced this bracketed value with a plant specific value of 540 degrees F. Operation at this temperature (540 degrees F) provides reasonable margin above the LCO 3.1.8 Technical Specification limit of 535 degrees F for the RCS lowest Tavg. Operation with the reactor critical and with the temperature below 535 degrees F could violate the assumptions for accidents analyzed in the safety analyses.

5. CTS 3.1.B.1.b M3.4-62 ITS 3.4.3 Action A.2 CTS 3.1.B.1.b states that if the conditions cannot be satisfied, the following must be done: 1) restore the temperature and/or pressure to within the limits within 30 minutes, 2) perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the RCS, and 3) determine that the RCS remains acceptable for continued operation.

ITS 3.4.3 Actions A.1 and A.2 require that the parameters are restored within limits in 30 minutes and determine that the RCS is acceptable for continued operation. ITS 3.4.3 Actions A.1 and A.2 does not account for the engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the RCS. M3.4-62 discusses the addition of the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months time limit to the CTS for evaluating the structural integrity of the RCS. This does not appear to be correct. M3.4-62 should discuss the addition of the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months time limit being added to the CTS for determining that the RCS structural integrity remains acceptable for continued operation 3

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS) which would be consistent with ITS 3.4.3 Action A.2. The CTS description of performing an engineering evaluation should be added to the ITS 3.4.3 Bases as added detail.

Comment: Correct M3.4-62 to include discussion to reflect the time allowance is added for the purpose of determining whether the RCS structural integrity remains acceptable for continued operation. Add the description of the engineering evaluation to the ITS 3.4.3 Bases and mark up the CTS to show where this requirement is going to be located.

NMC Response:

Parts affected by this change:

Part B: Final ITS pages Part C: CTS markup Part D: DOC M3.4-62 and LR3.4-101 Part E: ISTS markup Part F: JFD CL3.4-329 Part G: NSHD ITS Bases 3.4.3, Required Action A.1 and A.2, has been revised to include the CTS statement that an engineering evaluation may be one of the methods used to determine the effects of the out-of-limit condition on the structural integrity of the RCS. In addition, DOC M3.4-62 has been revised to clarify that the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time for the evaluation is sufficient to determine if the RCS structural integrity remains acceptable for continued operation.

6. CTS 3.1.A.1.a(1)

ITS LCO 3.4.4 CTS 3.1 .A.1 .a(1) states that the reactor shall not be made or maintained critical unless both reactor coolant loops (with their associated steam generator and reactor coolant pump) are in operation, except 1) during low power PHYSICS TESTS. The exception to CTS 3.1.A.1.a(1) does not appear in ITS LCO 3.4.4 although it is depicted this way on the marked up CTS.

Comment: Deletion of the CTS phrase "except during low power PHYSICS TESTS" needs to be evaluated in a discussion of change for ITS.

NMC Response:

Parts affected by this change:

Part C: CTS markup The CTS markup has been revised to more accurately reflect the ITS by putting the phrase "During low power PHYSICS TESTS" with its associated LCO 3.4.18 designator. CTS 4

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS) 3.1.A.1.a(1) was marked up to identify those requirements associated with LCO 3.4.4 and 3.4.18. The CTS, as was annotated, could cause confusion such that it appeared LCO 3.4.4 contained an exception that a reactor shall not be critical unless both reactor coolant loops (with their associated steam generator and reactor coolant pump) are in operation, except during low power PHYSICS TESTS. LCO 3.4.18 is for the RCS Loop - Test Exceptions. Therefore, the CTS has been revised to place the CTS exception statement "during low power PHYSICS TEST" in LCO 3.4.18, APPLICABILITY Statement.

7. ITS SR 3.4.5.2 X3.4-121 STS SR 3.4.5.2 requires the verification of steam generator secondary side water levels to be>

[17]% for required RCS loops. ITS SR 3.4.5.2 would require verification of steam generator secondary side water levels to be > 60% (Wide Range) for both RCS loops. ITS SR 3.4.5.2 is not consistent with the CTS (new SR) or STS.

Comment: ITS SR 3.4.5.2 is beyond scope.

NMC Response:

Parts affected by this change:

Part B: Final ITS Part C: CTS markup Part D: DOC Part E: ISTS markup Part F: JFD CL3.4-121 The ISTS SR 3.4.5.2 requires verification of the steam generator secondary side water levels to be greater than a bracketed value (17%). Bracketed values are suggested or examples with the plant substituting their plant specific values in their place. No justification is required for values that are already in the plants CLB or CTS. In this case, P1 does not have any specific steam generator level secondary side levels in the CTS. The suggested value in the ISTS is for the steam generator narrow range level instrumentation. PI proposed the wide range instrumentation since the level transmitters and channels are installed as class 1 E. Also, the wide range level indication is displayed using recorders in the control room and on indicators on the hot shutdown panels. Upon further evaluation, as a result of the RAI, PI has chosen to implement the NRC guidance as recommended in NRC Information Notice 95-35 as discussed below.

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Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

ITS has been revised to be consistent with the intent of the CTS which requires at least two methods of decay heat removal shall be OPERABLE with one in operation. Acceptable methods for removing decay heat are at least one reactor coolant pump and its associated SG. PI CTS does not require any specific level to be maintained. PI does require the SG to be OPERABLE and able to perform its intended function which is to provide a heat sink for decay heat removal.

In addition, Information Notice (IN) 95-35 and WOG-155, Rev. 0 which is being proposed as a TSTF. In addition, TSTF-114, Rev. 0 was approved by the NRC on Dec 31, 1996. TSTF-114 modified the Bases for ITS 3.4.7, RCS Loops - MODE 5, Loops Filled" to reference IN 95-35, "Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation" in the Bases whenever removal of decay heat by the SGs was discussed.

IN 95-35 discussed two incidents when the SGs were being relied upon for decay heat removal in MODE 5 with loops filled per the TS, but the SGs were not capable of performing that function because the RCS could not be pressurized. The IN provided additional details, however, stated that ifthe RCS pressure at the top of the SG tubes is allowed to fall below the primary fluid saturation temperature, flashing and steam voiding may occur, interrupting or degrading the natural circulation flow path. Additionally, when system pressure is dropped with elevated water temperatures, gases may come out of solution. The IN concluded that when relying on the SGs for decay heat removal, the following items must be considered: 1) the ability to pressurize and control pressure in the RCS, 2) the secondary side water level in the SGs relied on for decay heat removal, 3) the availability of a supply of feedwater; and 4) the availability of an auxiliary feedwater pump capable of injecting into the relied on SG.

The current ISTS is incomplete and misleading. TSTF-1 14 revised the Bases for LCO 3.4.7 and incorporated a reference to IN 95-35, but did not include sufficient information for an operator to recognize the additional requirements discussed in the IN. The ITS LCO 3.4.7 requirement that the secondary side water level of at least one SG be > 17% is insufficient to ensure the SG can be relied upon to remove heat from the RCS in the applicable conditions. The wording of the LCO and the referencing of the IN create a condition in which the document referenced in the Bases contains additional requirements necessary to meet the intent of the LCO. In addition, the concern raised in IN 95-35 does not apply in MODE 5 - Loops Filled. The concern also applies to MODE 3 and MODE 4. As a result, LCO 3.4.7, SR 3.4.5.1 and SR.3.4.6.1 have been revised to only require verification of SG secondary side water level and removes the specific level values.

8. CTS 3.1.A.11.b(3)(a)

ITS 3.4.5 Action D.1 CTS 3.1.A.1.b(3)(a) requires the immediate de-energization of all control rod drive mechanisms when both RCPs are inoperable or not in operation. ITS 3.4.5 Action D.1 requires that the Rod 6

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

Control System be placed in a condition incapable of rod withdrawal, immediately. Since there are a variety ways to make the rod control system incapable of rod withdrawal, the requirements of CTS 3.1 .A.1 .b(3)(a) and ITS 3.4.5 Action D.1 are not the same. The requirements of ITS 3.4.5 Action D.1 are a less restrictive change. No discussion of differences was provided for this change.

Comment: Provide DOC for ITS 3.4.5 Action D.1 as a less restrictive change.

NMC Response:

Parts affected by this change:

Part C: CTS markup Part D: DOC L3.4-118 Part G: NSHD CTS 3.1 .A.1 .b(3)(a) has been revised or more closely reflect ITS LCO 3.4.5, Required Action D which allows additional methods for placing the control rod drive system in a condition incapable of rod withdrawal. In accepting the allowed flexibility by ISTS LCO 3.4.5, Required Action D, DOC L3.4-118 was generated with its associated NSHD.

9. CTS 3.1.A.1.b(3)*

L3.4-23 ITS 3.4.5 Note CTS 3.1 .A.1 .b(3)* allows both RCPs not in operation for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for preplanned work activities. This was approved by the staff in Licence Amendments 152/143 for Prairie Island Units I and 2. This allowance is also proposed in ITS 3.4.5 Note. L3.4-23 proposes to add the statement "De-energizing control rods is not required for preplanned work activities" to allow for additional tests in this 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months time period. However, the staff credited the de-energization of all control rod drive mechanisms as part of its approval of CTS 3.1 .A.1 .b(3)*. Adding the phrase proposed by L3.4-23 invalidates part of the staff's approval in its SER.

Comment: This is a beyond scope issue.

NMC Response:

Parts affected by this change:

Part C: CTS markup Part D: DOC L3.4-23 Part G: NSHD 7

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

The CTS has been revised deleting the subject phrase. As a result, DOC L3.4-23 and its associated NSHD have been deleted.

10. ITS 3.4.6 Note 1 PA3.4-120 ITS 3.4.6 Note 1 states that all RCPs and RHR pumps may be not in operation for < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. PA3.4-120 adds an additional comment to the note which states that all RCPs and RHR pumps may be not in operation for !<1 hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period to perform tests. This additional phrase is not in the CTS or the STS. However, the additional phrase is discussed in the ITS 3.4.6 Bases. Requirements cannot be stated in the Bases. Additionally, the Bases Control Program controls future Bases changes to ensure the meaning of LCOs are not changed as a result of Bases changes.

Comment: Retain the STS 3.4.6 Note 1 wording.

NMC Response:

Parts affected by this change:

Part B: Final ITS pages Part E: ISTS markup Part F: JFD PA3.4-120 ISTS LCO 3.4.6 Note 1 has been revised by deleting the statement "to perform tests". As a result, JFD PA3.4-120 was also deleted.

11. ITS 3.4.6 Actions A, B, and C CTS 3.1.A.1.e(2) and (3)

TSTF-263 R3 CL3.4-113 ITS 3.4.6 Actions A, B, and C provide the required actions when two loops (RCS or RHR) are not operable and one loop is not in operation in MODE 4. ITS 3.4.6 Actions A, B, and C are similar to CTS 3.1.A.1.e(2) and e(3). However, ITS 3.4.6 Actions A, B, and C are not consistent with STS 3.4.6 Actions A and B since TSTF-263 Rev. 3 was not adopted. CL3.4-113 stated that TSTF-263 Rev. 3 was not included since PI is a two loop plant. CL3.4-113 does not justify why STS 3.4.6 Actions A and B were not adopted.

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Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

Comment: Adoption of STS 3.4.6 Actions A and B is conservative for a two loop plant and should be considered. Justify not adopting STS 3.4.6 Actions A and B for ITS.

NMC Response:

Parts affected by this change:

Part B: Final ITS pages Part C: CTS markup Part D: DOCs A3.4-104, A3.4-105, and A3.4-106 Part E: ISTS markup Part F: JFD3.4-124 PI incorporated TSTF 263, Rev. 3, as revised for PI design. PI is a two loop plant; therefore, some of the TSTF was editorially changed to comply with our design.

12. ITS 3.4.6 Action B Note X3.4-124 X3.4-124 adds a note stating that "required action B.1 is not applicable if all RCS and RHR loops are inoperable and Condition C is entered." This note is not in the CTS or the STS. Additionally, this added instruction would not be needed if the STS 3.4.6 Actions A and B were adopted (see RAI 11). Based on the wording of ITS 3.4.6 Action B, it is not clear why Action B Note is necessary.

Comment: Provide further justification for the addition of the plant specific Note to ITS 3.4.6 Action B.

NMC Response:

Parts affected by this change:

Part B: Final ITS pages Part E: ISTS markup Based on incorporating TSTF-263, Required Action B was deleted. Therefore, there are no changes associated with this RAI. Reference RAI 3.4-11 above.

13. ITS SR 3.4.6.2 9

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

X3.4-121 STS SR 3.4.6.2 requires the verification of steam generator secondary side water levels to be>

[17]% for required RCS loops. ITS SR 3.4.6.2 would require verification of steam generator secondary side water levels to be > 60% (Wide Range) for each required RCS loops. ITS SR 3.4.6.2 is not consistent with the CTS (new SR) or STS.

Comment: ITS SR 3.4.6.2 is beyond scope.

NMC Response:

Parts affected by this change:

Part B: Final ITS Part C: CTS markup Part D: DOC M3.4-31 Part E: ISTS markup Part F: JFD PA3.4-121 Reference response to RAI 3.4-7.

14. ITS LCO 3.4.7.b X3.4-121 STS LCO 3.4.7.b requires that the steam generator secondary side water levels of two SGs be >

[17]%. ITS LCO 3.4.7.b would require that the steam generator secondary side water level of one SG be > 60% (Wide Range). ITS LCO 3.4.7.b is not consistent with the CTS or STS.

Comment: ITS LCO 3.4.7.b is beyond scope.

NMC Response:

Parts affected by this change:

Part B: Final ITS Part C: CTS markup Part D: DOC Part E: ISTS markup Part F: JFD PA3.4-121 Reference RAI response 3.4-7.

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Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

15. ITS 3.4.7 Note 1 PA3.4-120 ITS 3.4.7 Note 1 states that the RHR pump of the loop in operation may be not in operation for <

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. PA3.4-120 adds an additional comment to the note which states that the RHR pump of the loop in operation may be not in operation for _<1 hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period to perform tests. This additional phrase is not in the CTS or the STS. However, the additional phrase is discussed in the ITS 3.4.7 Bases and therefore should not be added to the ITS 3.4.7 Note 1.

Comment: Retain the STS 3.4.7 Note 1 wording.

NMC Response:

Parts affected by this change:

Part B: Final ITS pages Part E: ISTS markup Part F: JFD PA3.4-120 Reference response to RAI 3.4-10.

16. ITS 3.4.7 Actions A and B CL3.4-113 TSTF-263 Rev. 3 ITS 3.4.7 Actions A, and B provide the required actions when one RHR is not operable and one RHR loop is not in operation in MODE 5, Loops Filled. ITS 3.4.6 Actions A and B are not consistent with STS 3.4.6 Actions A and B since TSTF-263 Rev. 3 was not adopted. CL3.4-113 stated that TSTF-263 Rev. 3 was not included since PI is a two loop plant. CL3.4-113 does not justify why STS 3.4.7 Actions A, B, and C were not adopted. Additionally, the required actions proposed in ITS 3.4.7 do not address the situation where one RHR is inoperable while the other RHR is operable, as specified in ITS 3.4.7.a. ITS actions should address all conditions of the LCO.

Comment: Correct the ITS actions by adopting TSTF-263 Rev. 3 and STS 3.4.7 Actions A, B, and C.

NMC Response:

Parts affected by this change:

Part B: Final ITS pages 11

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

Part C: CTS markup Part D: DOCs A3.4-120, A3.4-121, A3.4-122 Part E: ISTS markup PI incorporated TSTF 263, Rev. 3, as revised for PI design. PI is a two loop plant; therefore, some of the TSTF was editorially changed to comply with our design.

17. ITS 3.4.7 Action A The AND statement of ITS 3.4.7 Condition A states that "both SGs secondary side water levels not within limits." This is not consistent with ITS LCO 3.4.7.b which states that "the secondary side water level of at least one steam generator (SG) shall be _>60% (Wide Range)."

Comment: Correct the ITS LCO 3.4.7.b or ITS 3.4.7 Action A such that they are consistent with one another.

NMC Response:

Parts affected by this change:

Part B: Final ITS Part C: CTS markup Part D: DOC M3.4-26 Part E: ISTS markup Part F: JFD PA3.4-121 Reference response to RAI 3.4-7.

18. ITS SR 3.4.7.2 X3.4-121 STS SR 3.4.7.2 requires the verification of steam generator secondary side water levels to be

[17]% in required SGs. ITS SR 3.4.7.2 would require verification of steam generator secondary side water levels to be > 60% (Wide Range) in the required SGs. ITS SR 3.4.7.2 is not consistent with the CTS (new SR) or STS.

12

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

Comment: ITS SR 3.4.7.2 is beyond scope.

NMC Response:

Parts affected by this change:

Part B: Final ITS Part C: CTS markup Part D: DOC M3.4-32 Part E: ISTS markup Part F: JFD PA3.4-121 Reference response to RAI 3.4-7.

19. CTS 3.1.A.l.c(2)

ITS 3.4.7 Action A The AND statement of ITS 3.4.7 Condition A states that "both SGs secondary side water levels not within limits." CTS 3.1 .A.1 .c(2) does not have this condition. The addition of this condition to the ITS was not discussed in the mark up of the CTS, i.e, this change was not included in the discussion of changes for CTS 3.1 .A.1 .c(2).

Comment: Correct the CTS mark up and provide the appropriate discussion of change for the AND statement of ITS 3.4.7 Condition A.

NMC Response:

Parts affected by this change:

None This item is deleted based on incorporating TSTF 263,

20. ITS 3.4.8 Note 1 CL3.4-131 M3.4-34 STS 3.4.8 Note 1 states that "all RHR pumps may be not in operation for _*15 minutes when switching from one loop to another provided:..." CL3.4-131 proposed not to include the phrase

'switching from one loop to another,' since the Prairie Island CTS allows both RHR pumps to be inoperable up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months without restrictions on the nature of the required operability. M3.4-34 discusses the change to the CTS which would limit the time both RHR pumps can be inoperable 13

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

(from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 15 minutes). This more restrictive change should also include the STS language of 'switching from one loop to another,' since the CTS is being changed.

Comment: Adopt the STS wording for STS 3.4.8 Note 1 or provide further plant specific justification as to why the note is acceptable without the STS wording.

NMC Response:

Parts affected by this change:

Part B: Final ITS pages Part C: CTS markup Part D: DOC M3.4-34 Part E: ISTS markup Part F: JFD CL3.4-131 PI has re-evaluated our response and revised the subject Note to allow the RHR pumps to be shutdown for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period while in Mode 5 with loops not filled. PI CLB and CTS, as approved by the NRC, allows both pumps to be shutdown for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , as long as the restrictions, as stated in LCO 3.4.8 Notes, are maintained. Based on the conversion agreements between the NRC and industry, plants are allowed to maintain their CLB and CTS.

21. CTS 3.1.A.2.a(3)

ITS 3.4.9 Action A M3.4-43 ITS 3.4.9 Action A requires if the pressurizer water level is not within limit, 'be in MODE 3, fully insert all rods, and place rod control system in a condition incapable of rod withdrawal,' within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . This is consistent with STS 3.4.9 Action A. The mark up of CTS 3.1 .A.2.a(3) implies that the completion time is 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to be in MODE 3, insert all rods within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and place rod control system in a condition incapable of rod withdrawal within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The CTS mark up is confusing and does not reflect the STS or ITS.

Comment: Correct the CTS markup to adequately reflect the ITS 3.4.9 Action A statements.

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Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

NMC Response:

Parts affected by this change:

Part C: CTS markup CTS 3.1 .A.2.a (3) has been revised to properly state that if the Pressurizer water level is not within limits then all Required Actions A.1, A.2, and A.3 are to be completed in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months from the time the Condition is entered. This is consistent with NUREG-1431.

22. CTS 3.1.A.2.a(3)

ITS SR 3.4.9.1 M3.4-44 M3.4-44 discusses the changes to the CTS which includes the addition of ITS SR 3.4.9.1. M3.4 44 states that this SR will require periodic verification of a steam bubble in the SG. The CTS mark up states the new SR will verify steam bubble in the pressurizer. However, the actual wording of ITS SR 3.4.9.1 is verify pressurizer water level is _<90%.

Comment: Correct the documentation to support the new SR, ITS SR 3.4.9.1.

NMC Response:

Parts affected by this change:

Part C: CTS markup Part D: DOC M3.4-44 The CTS and associated DOC have been revised by replacing the phrase of verification of the steam bubble in the pressurizer with verify the water level in the pressurizer is-< 90%. This is consistent with NUREG-1431, SR 3.4.9.1.

23. CTS 3.1 .A.2.b(1)

A3.4-46 ITS LCO 3.4.10 ITS SR 3.4.10.1 CTS Table TS 4.1-2A Item 3 CTS 3.1 .A.2.b(1) requires that a reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 350'F unless two pressurizer safety valves are OPERABLE, with lift settings of 2485 psig, +/-1%. ITS LCO 3.4.10 states that two pressurizer safety valves shall be OPERABLE with lift settings >_2410 psig and _<2560 psig. These lift 15

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS) settings are 2485 psig +/-3%. The discussion of changes for this proposed change was A3.4-46 which states that License Amendment 123 approved the use of +/-3% for operability testing. The staff agrees that License Amendment 123 approved the use of +/-3 for operability testing for CTS Table TS 4.1-2A Item 3 but CTS 3.1.A.2.b.(1) was not change with this amendment. Therefore, changing +/-1% to +/-3% in the CTS mark up and ITS LCO 3.4.10 is not an administrative change.

Additionally, the proposed ITS LCO 3.4.10 lift settings (at +/-3%) are inconsistent with the lift settings in ITS SR 3.4.10.1 (at +/-1%). The lift settings as stated in the LCO should be consistent with the lift settings stated in the SR. For PI, the lift settings for the LCO and the SR should be

+/-1% unless the licensing basis is changed. This is not in conflict with the conclusions of license amendment 123 due to the following:

Although the pressurizer safety valves must be set to +/-1% during the surveillance, the pressurizer safety valves satisfy safety analysis assumptions and meet ASME Code requirements if the setpoint is determined to be +/-3% at the end of the surveillance interval.

Therefore, the pressurizer safety valve setpoint is +/-3% for OPERABILITY; however, the valves must be reset to +/-1% during the surveillance to allow for drift.

Comment: Maintain the CTS requirements and correct ITS LCO 3.4.10 to show the lift settings at +/-1%.

NMC Response:

Parts affected by this change:

None The LCO lift setting is based on PI CTS, NRC approved SER License Amendment 123/116 dated May 21, 1996, and Current Licensing Basis for the plant. NRC staff concurred with the proposed ITS LCO statement by telephone call on January 23, 2002. No changes have been made to the ITS submittal based on this RAI.

24. ITS SR 3.4.10.1 PA3.4-143 STS SR 3.4.10.1 requires the verification that each pressurizer safety valve is OPERABLE in accordance with the Inservice Testing Program. Following testing, lift settings shall be within

+/-1%. ITS SR 3.4.10.1 requires that following testing, lift settings shall be within 2460 to 2510 psig. PA3.4-143 stated that the proposed change was provided for clarity. This proposed change is plant specific and not consistent with the CTS or STS.

Comment: Maintain the CTS and STS wording +/-1%.

16

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

NMC Response:

Parts affected by this change:

Part B: Final ITS pages Part E: ISTS markup ITS SR 4.5.10.1 has been revised to include the + 1% pressurizer safety valve lift setting. In addition, for consistency the actual range of 2460 to 2510 psig has also been retained.

25. CTS 3.1.A.2.b(1)

M3.4-45 ITS 3.4.10 Action B CTS 3.1.A.2.b(1) requires the reactor coolant system average temperature be below 350°F within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months total) if the conditions cannot be satisfied. ITS 3.4.10 Action B allows 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for this mode change. M3.4-45 describes the change from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months as a more restrictive change since the plant has to be cooled down further than the CTS. Changing the allowed completion time from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a less restrictive change and should be documented as such.

Comment: Provide the proper discussion of changes for changing the completion time from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

NMC Response:

Parts affected by this change:

Part C: CTS markup Part D: DOC M3.4-45 and L3.4-109 The CTS has been revised to document that going from 12 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a less restrictive change. In addition, the associated DOCs have been revised or added.

26. CTS 3.1.A.2.c.(1).(b).4 ITS 3.4.11 Action C Note ITS 3.4.11 Action C has a Note which states that Required Actions C.1 and C.2 do not apply when block valve is inoperable solely as a result of complying with Required Action B.2 or E.2.

The CTS does not appear to have this note. No discussion of change was included on the CTS mark up to discuss the addition of this note.

Comment: Provide discussion of change for the addition of ITS 3.4.11 Action C Note.

17

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

NMC Response:

Parts affected by this change:

Part C: CTS markup Part D: DOC A3.4-110 The CTS has been revised to specifically show LCO 3.4.11, Condition C Note. In addition, DOC A.3.4-1 10 is provided documenting this administrative change.

27. CTS 3.1.A.2.c.(1).(b).5 ITS 3.4.11 Action F Note ITS 3.4.11 Action F has a Note which states that Required Action F.1 does not apply when block valve is inoperable solely as a result of complying with Required Action B.2 or E.2. The CTS does not appear to have this note. No discussion of change was included on the CTS mark up to discuss the addition of this note.

Comment: Provide discussion of change for the addition of ITS 3.4.11 Action F Note.

NMC Response:

Parts affected by this change:

Part C: CTS markup Part D: DOC A3.4-1 11 The CTS has been revised to specifically show LCO 3.4.11, Condition F Note. In addition, DOC A.3.4-1 11 is provided documenting this administrative change.

28. CTS 3.1.A.2.c.(1).(b).5 A3.4-49 ITS 3.4.11 Action F CTS 3.1 .A.2.c.(1).(b).5 requires that with both block valves inoperable, within one hour either restore the block valves to OPERABLE status or place the PORVs in manual control.

Additionally, restore at least one block valve to OPERABLE status within the next hour. ITS 3.4.11 Action F requires the restoration of one block valve to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

18

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

A3.4-49 does not discuss the deletion of the action to place the PORVs in manual control if both block valves are inoperable.

Comment: Provide further justification for deleting the CTS requirement to place the PORVs in manual control if both block valves are inoperable.

NMC Response:

Parts affected by this change:

Part C: CTS markup Part D: DOC M3.4-123 Part G: NSHD The CTS has been revised adding DOC M3.4-123 and associated NSHD providing additional justification for deleting the CTS requirement of placing the PORVs in manual control when both block valves are inoperable.

29. CTS Table TS 4.1-2A ITS SR 3.4.11.1 Note 2 ITS SR 3.4.11.1 has a Note 2 which states that SR 3.4.11.1 is only required to be performed in MODES 1 and 2. The CTS does not appear to have this note. No discussion of change was included on the CTS mark up to discuss the addition of this note.

Comment: Provide discussion of change for the addition of ITS SR 3.4.11.1 Note 2.

NMC Response:

Parts affected by this change:

Part C: CTS markup Part D: DOC A3.4-113 The CTS has been revised adding DOC A3.4-113 providing additional justification for the subject Note which only requires the SR to be performed in Modes 1 and 2.

30. CTS Table TS 4.1-2A ITS SR 3.4.11.2 Note 19

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

ITS SR 3.4.11.2 has a Note which states that SR 3.4.11.2 is only required to be performed in MODES 1 and 2. The CTS does not appear to have this note. No discussion of change was included on the CTS mark up to discuss the addition of this note.

Comment: Provide discussion of change for the addition of ITS SR 3.4.11.2 Note.

NMC Response:

Parts affected by this change:

Part C: CTS markup Part D: DOC A3.4-114 The CTS has been revised adding DOC A3.4-114 providing additional justification for the subject Note which only requires the SR to be performed in Modes 1 and 2.

31. ITS 3.4.12 Title and LCO ITS 3.4.13 Title and LCO CTS 3.1.A.2.c.(2)

TA3.4-119 TA3.4-119 incorporates TSTF-233 but modifies the inserted phrase from LTOP to Over Pressure Protection System (OPPS) which is the PI specific terminology. CTS 3.1 .A.2.c.(2) uses the OPPS terminology. The following ITS specifications used the OPPS terminology instead of LTOP: ITS 3.4.6 Note 2, ITS 3.4.7 Note 3, ITS 3.4.10 Applicability, ITS 3.4.10 Required Action B.2 and ITS 3.4.12 Applicability. However, the title and LCO for ITS 3.4.12 and ITS 3.4.13 uses the phrase LTOP. This inconsistency is unacceptable.

Comment: Correct the ITS sections to use either the OPPS or LTOP terminology. If the LTOP terminology is used, justification is required since the CTS uses the OPPS terminology.

NMC Response:

Parts affected by this change:

Part B: Final ITS pages Part E: ITST Markup The ISTS LCO 3.4.12 has been revised to change the title to LTOP - RCSCLT > SI Pump Disable Temperature. RCSCLT stands for Reactor Coolant System Cold Leg Temperature.

LCO 3.4.13 title has also been changed to LTOP - RCSCLT < Sl Pump Disable Temperature.

This change is made to be consistent with the PTLR and other associated specifications. The use of OPPS would not be correct for LCO 3.4.12 and 3.4.13. The OPPS is a circuit that 20

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS) controls the PORVs. RCS low temperature overpressure protection includes control of the PORVs through the OPPS, limiting the number of SI pumps capable of injecting into the RCS and limiting the ability of the ECCS accumulator to inject into the RCS.

32. ITS 3.4.12 entire spec ITS 3.4.13 entire spec CTS 3.1.A.2.c.(2)

CTS 3.1.A.2.c.(3)

ITS 3.4.12 and ITS 3.4.13 provide the requirements for LTOP > SI pump disable temperature and LTOP _<Sl pump disable temperature. STS 3.4.12 has the LTQP requirements in one LCO which requires that an LTOP System be OPERABLE with a maximum of [one] [high pressure injection (HPI)] pump [and one charging pump] capable of injecting into the RCS and the accumulators isolated and one pressure relief capabilities. The proposed ITS 3.4.12 and ITS 3.4.13 are not acceptable and do not appear to be consistent with the CTS.

Comment: Review the ITS 3.4.12 for Ginna and work with the staff to produce one acceptable LTOP (or OPPS - once a single terminology is selected) LCO which incorporates the CTS.

NMC Response:

Parts affected by this change:

None Based on a meeting with the NRC on December 19, 2001, PI explained our design and reasoning for writing the ITS 3.4.12. The NRC stated that no further action was required of PI and that the NRC would further review ITS 3.4.12 based on the discussed information and another plant with the same system as Pl. Therefore, no further information is being provided at this time.

33. CTS Table 4.1-2A Item 9 ITS SR 3.4.14.1 L3.4-89 ITS SR 3.4.14.1 requires the verification of RCS operational leakage is within limits by performance of RCS water inventory balance every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . CTS Table 4.1-2A Item 9 requires the verification every day. L3.4-89 states that this increase surveillance interval is considered acceptable based on the leakage detection systems required to be operable by LCO 3.4.16.

However, ITS 3.4.14 Action statements retain the CTS required actions. The current required actions are less restrictive that the required actions in STS 3.4.13 (different numbering same 21

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS) topic). Based on retaining the current required actions, the current frequency of the CTS should also be retained.

Comment: Revise ITS SR 3.4.14.1 frequency such that it is consistent with the CTS Table 4.1 2A Item 9.

NMC Response:

Parts affected by this change:

Part B: Final ITS pages Part C: CTS markup Part D: DOC L3.4-89 Part E: ISTS markup Part F: JFD CL3.4-331 CTS Table 4.1-2A, Item 9 has been revised to retain the CTS Frequency of "daily". As a result, DOC L3.4-89 was deleted, ISTS SR 3.4.13.1 (ITS SR 3.4.14.1) Frequency was changed from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with JFD CL3.4-331 written to justify the Frequency change.

34. CTS Table 4.1-2A ITS SR 3.4.14.1 NOTE ITS SR 3.4.14.1 has a Note which states that SR 3.4.14.1 is not required to be performed until 12 after establishment of steady state operation. The CTS does not appear to have this note.

No discussion of change was included on the CTS mark up to discuss the addition of this note.

Comment: Provide discussion of change for the addition of ITS SR 3.4.14.1 Note.

NMC Response:

Parts affected by this change:

Part C: CTS markup Part D: DOC A3.4-112 The CTS has been revised to annotate ITS SR 3.4.14.1 Note. In addition, DOC A3.4-112 was generated to provide documentation of this administrative change.

22

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

35. ITS 3.4.17 LCO ITS 3.4.17 Action A PA3.4-197 STS LCO 3.4.16 states that the specific activity of the reactor coolant shall be within limits. ITS LCO 3.4.17 proposes to add the specific limits to the LCO in 3.4.17.a and 3.4.17.b. PA3.4-197 states that the LCO and Condition A wording is revised to be consistent with each other and with Condition B. ITS 3.4.17 Action A is revised to state "Dose Equivalent 1-131 specific activity not within limit." These changes are not consistent with STS 3.4.16 and are generic in nature.

Additionally, ITS SR 3.4.17.1 and ITS SR 3.4.17.2 specify the required limits and therefore, the limits should not be specified in the LCO.

Comment: Retain the STS wording for 3.4.16 LCO and Action A.

NMC Response:

Parts affected by this change:

Part B: Final ITS pages Part C: CTS markup Part E: ISTS markup Part F: JFD PA3.4-197 The CTS has been revised identifying corresponding ITS SR 3.4.17.1 and 3.4.17.2. In addition, the wording for the ISTS LCO, Action A, and associated Bases have been retained. As a result, PA3.4-197 has been deleted.

36. ITS 3.4.17 Action C PA3.4-202 The second condition of ITS 3.4.17 Action C is modified to include the phrase "specific activity."

PA3.4-202 states that Condition C is revised to clarify the terminology. This proposed change is generic in nature and not justified as a plant specific change.

Comment: Retain the STS wording for ITS 3.4.17 Action C or provide plant specific justification for the proposed change.

NMC Response:

Parts affected by this change:

Part B: Final ITS pages Part E: ISTS markup 23

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

Part F: JFD PA3.4-202 The ITS has been revised to retain the wording of ISTS 3.4.17 Action C. As a result, JFD PA3.4-202 has been deleted.

37. ITS SR 3.4.17.3 PA3.4-203 ITS SR 3.4.17.3 is modified to state "Determine E from a reactor coolant sample. PA3.4-203 states that SR 3.4.17.3 has been modified to eliminate redundancy and clarify intent. This proposed change is generic in nature and not justified as a plant specific change.

Comment: Retain the STS wording for ITS SR 3.4.17.3 or provide a plant specific justification for the proposed change.

NMC Response:

Parts affected by this change:

Part B: Final ITS pages Part C: CTS markup Part D: DOC A3.4-124 Part E: ISTS markup Part F: JFD PA3.4-203 The ISTS wording for SR 3.4.17.3 has been retained. As a result, the CTS has been revised to reflect the change and the associated DOC A3.4-124 generated to discuss the subject change.

38. CTS Table TS.4.1-2B Item 1 ITS SR 3.4.17.1 L3.4-88 CTS Table TS.4.1-2B Item 1 requires the RCS Gross Activity determination to be performed 5/week. ITS SR 3.4.17.1 would require the RCS Gross Activity determination to be performed every 7 days. L3.4-88 states that the surveillance interval for RCS gross activity determination would be increased to once per week by this change in conformance with the guidance of NUREG-1431. L3.4-88 incorrectly states that the frequency is being increased versus the actual proposed decrease in frequency of the surveillance.

Comment: Correct L3.4-88 to be representative of the proposed change.

24

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

NMC Response:

Parts affected by this change:

None DOC L3.4-88 is correct as stated. CTS Table 4.1-2B Item 1 requires the RCS Gross Activity determination to be performed 5/week (5 times per week). This would result in this SR being performed 20 times during a 4 week period. The ITS requires this SR to be performed every 7 days or 4 times during the same period. Going from performing this SR from 20 to 4 times is a decrease in Frequency. Inversely, going from performing this SR from 20 times to 4 times, during the same period of time interval, would increase the time between performing the SR.

The interval in the CTS is essentially daily, whereas, the ITS interval is increased to performing the SR once in 7 days.

39. CTS Table TS.4.1-2B Item 2 ITS SR 3.4.17.2 Note ITS SR 3.4.17.2 has a Note which states that SR 3.4.17.2 is only required to be performed in MODE 1. The CTS does not appear to have this note. No discussion of change was included on the CTS mark up to discuss the addition of this note.

Comment: Provide discussion of change for the addition of ITS SR 3.4.17.2 Note.

NMC Response:

Parts affected by this change:

Part C: CTS markup Part D: DOC A3.4-125 The CTS has been revised to add DOC A.3.4-125 discussing the addition of ITS SR 3.4.17.2 Note.

40. CTS Table TS.4.1-2B Note 1 ITS SR 3.4.17.3 Note ITS SR 3.4.17.3 has a Note which states that SR 3.4.17.3 is not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . CTS Table TS.4,1-2B Note 1 states that samples to be taken after a minimum of 2 EFPD and 20 days of Power Operation have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer. ITS Sr 3.4.17.3 Note and 25

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

CTS Table TS.4.1-2B Note are not the same. No discussion of change was included on the CTS mark up to discuss the differences between the two notes.

Comment: Provide discussion of change for the addition of ITS SR 3.4.17.3 Note.

NMC Response:

Parts affected by this change:

Part C: CTS markup Part D: DOC M3.4-117 The CTS has been revised to add DOC M3.4-117 discussing the addition of SR 3.4.17.3 Note.

41. CTS 3.1.A.1.b(2)

A3.4-08 CTS 3.1 .A.1 .b(2) is modified to state be in MODE 4 versus reduce reactor coolant system average temperature below 350 0 F. This change is described as A3.4-08. CTS 3.1 .A.!.b(2) does not appear on the list for A3.4-08 as one of the specifications altered by this change.

Comment: Add CTS 3.1 .A.1 .b(2) to the list for A3.4-08.

NMC Response:

Parts affected by this change:

Part D: DOC A3.4-08 DOC A3.4-08 has been revised to add CTS 3.1.A.1.b (2).

42. CTS 3.1.A.3 TS 4.18 R3.4-56 CTS 3.1 .A.3 is proposed to be relocated to the TRM. R3.4-56 states that this relocation is acceptable since CTS 3.1.A.3 and the associated surveillance requirements in TS 4.18 do not meet the criteria of 10 CFR 50.36. However, each criteria of 10 CFR 50.36 was not addressed in order to demonstrate that this system does not meet the 10 CFR 50.36 criteria.

26

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

Comment: Provide further documentation which addresses each criteria of 10 CFR 50.36 to demonstrate that this system is not required to be in TS.

NMC Response:

Parts affected by this change:

Part D: DOC R3.4-56 The RAI has been revised to address each criteria of 10CFR50.36. Based on this evaluation, the reactor head vent system does not meet any of the criteria and is therefore being relocated.

43. CTS 3.1.B.2 R3.4-66 CTS 3.1 .B.2 is proposed to be relocated to the PTLR. R3.4-66 states that this change is acceptable since the Bases for Specification 3.4.3 (STS Bases?) state that the reactor pressure vessel is the most limiting component for brittle fracture; thus the requirements for the pressurizer have not been included in the ITS. This is not adequate justification for relocating technical specifications.

Comment: Provide further documentation which addresses each criteria of 10 CFR 50.36 to demonstrate that the pressurizer heatup and cooldown specifications are not required to be in TS.

NMC Response:

Parts affected by this change:

Part D: DOC R3.4-66 The RAI has been revised to address each criteria of 10CFR50.36. Based on this evaluation, the Pressurizer Pressure/Temperature limits do not meet any of the criteria and are therefore being relocated.

27

Prairie Island Nuclear Generating Plant Improved TS Review Comments ITS Section 3.4, Reactor Coolant System (RCS)

44. CTS 3.1..B.3 R3.4-67 CTS 3.1 .B.3 is proposed to be relocated to the PTLR. R3.4-67 states that this relocation is acceptable since CTS 3.1 .B.3 and the associated surveillance requirements do not meet the criteria of 10 CFR 50.36. However, each criteria of 10 CFR 50.36 was not addressed in order to demonstrate that this system does not meet the 10 CFR 50.36 criteria.

Comment: Provide further documentation which addresses each criteria of 10 CFR 50.36 to demonstrate that this system is not required to be in TS.

NMC Response:

Parts affected by this change:

Part D: DOC R3.4-67 The RAI has been revised to address each criteria of 10CFR50.36. Based on this evaluation, the Steam Generator Pressure/Temperature Limits do not meet any of the criteria and are therefore being relocated.

28

Prairie Island Nuclear Generating Plant Attachment 2 to Supplement dated February 15, 2002 to License Amendment Request dated December 11, 2000 Conversion to Improved Technical Specifications (ITS)

Page List by RAI Q (Revision 9)

Page List by RAI Q 2/15/2002 RAI Q # Package# Part Page#

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Prairie Island Nuclear Generating Plant Attachment 3 to Supplement dated February 15, 2002 to License Amendment Request dated December 11, 2000 Conversion to Improved Technical Specifications (ITS)

Revision 9 Change Pages

Improved Technical Specifications Supplement dated 2/15/02 Revision 9 Change Page List UPDATING INSTRUCTIONS Remove Insert Chapter/ Revision/ Chapter/ Revision/

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B 3.4.5-7 3.4 B 3.4 B B 3.4.5-8 i ....... . ............................................

3.4 B iB 3.4,6-2 3.4 B B 3.4.6-3 3.4 B B 3.4.6-4 3.4 B B 3.4.6-5 3.4 B B 3.4.6-6 3..4 B 3.4 B 3.4 B B 3.4.7-2 3.4 B B 3.4.7-4 3.4 .......

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Section Part Page Date Section Part Page Date 3.4 B B 3.4.13-3 12/11/00 B... 13- 4 3 ..... S--12/11/00 3.4 B B 3.4.13-5 12/11/00 3.4 B B.3.4. 13..

3.4 B 3.4.13-5 12/11/00 B

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C 13 of 28 3.4 3 of 13

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Section Part Page Date Section Part Page Date 3.4 E B 3.4.7-4 B 3.4.7-5 3.4 E 3.4 B 3.4.7-6 E

3.4 ...........................

B 3.4.7-7 3.4 E .........

3.4 B 3.4.8-2 E

3.4 B 3.4.8-3 E

3.4 B 3.4.8-4 3.4 E B 3.4.8-5 E

3.4 B 3.4.12-1 E

3.4 E B 3.4.12-2 E

B 3.4.12-3 3.4 E B 3.4.12-4 3.4 E B 3.4.12-5 3.4 B 3.4.12-6 3.4 B 3.4.12-7 3.4 E B 3.4.12-8 3.4 E E B 3.4.12-9 3.4 E B 3.4.12-10 3.4 E . 3. ..............................................

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9 F

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..=..............

3.4 F 41 3.4 F 42 3.4 F 43 3.4 F 44 3.4 F 45 3.4 F F.............

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Section Part Page Date Section Part Page Date 3.6 B B 3.6.3-2 12/1/01 B 3.6.3-6 3.6 B 5/1/01 3.6 B 3.6.3-7 B 12/1/01 3.6 B 3.6.3-8 12/11/00 B 12/11/00 3.6 B 3.6.3-9 i12/1/01 B

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3.6 E B 3.6.3-8 2

6 E B 3.6.3-9 6

3.6 E B 3.6.3-11 -1 12 of 13

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Section Part Page Date B 3.6.8-1 3.6 E B 3.6.8-4 3.6 E 2 3.6 B 3 6.8-5 2.....................

E 2 3.6 B 3.6.8-6 E 2 3.6 i2 E B 3.6.8-7 .. ..

3.6 12/1/01 F 5 3.6 5/1/01 F 20 3.6 F 5/1/01 21 3.6 F 5/1/01 23 3.6 F 5/1/01 29 3.6 0 5/1/01 3.6 G 5/1/01 3

3.6 G 5/1/01 34 3.6 G 5/1/01 35 13 of 13

RCS Pressure, Temperature, and Flow - DNB Limits 3.4.1 CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is greater than or equal 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to the limit specified in the COLR.

SR 3.4.1.2 Verify RCS average temperature is less than or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months equal to the limit specified in the COLR.

SR 3.4.1.3 Verify RCS total flow rate is within the limit 24 months specified in the COLR.

Prairie Island Units 1 and 2 3.4.1-2 1/2/02

RCS Loops - MODE 3 3.4.5 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.5 RCS Loops-MODE 3 LCO 3.4.5 Two RCS loops shall be OPERABLE, and either:

a. Two RCS loops shall be in operation when the Rod Control System is capable of rod withdrawal; or

b. One RCS loop shall be in operation when the Rod Control System is not capable of rod withdrawal.

NOTE Both reactor coolant pumps may be de-energized for _*12 hours to perform preplanned work activities provided:

a. No operations are permitted that would cause introduction into the RCS, coolant with boron concentration less than required to meet the SDM of LCO 3.1.1; and

b. Core outlet temperature is maintained at least 10°F below saturation temperature.

APPLICABILITY: MODE 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RCS loop A.1 Restore inoperable RCS 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable, loop to OPERABLE status.

Prairie Island Units 1 and 2 3.4.5-1 1/2/02

RCS Loops - MODE 3 3.4.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required RCS loops are in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.5.2 Verify required steam generator capable of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months removing decay heat.

SR 3.4.5.3 ------------------- NOTE -------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

Verify correct breaker alignment and indicated 7 days power are available to each required pump.

Prairie Island Units 1 and 2 3.4.5-4 i/2/02

RCS Loops - MODE 4 3.4.6 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Loops- MODE 4 LCO 3.4.6 Two loops consisting of any combination of RCS loops and residual heat removal (RHR) loops shall be OPERABLE, and one loop shall be in operation.

NOTES ---------------------

1. All reactor coolant pumps (RCPs) and RHR pumps may be de energized for _*1 hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a. No operations are permitted that would cause introduction into the RCS, coolant with boron concentration less than required to meet the SDM of LCO 3.1.1; and

b. Core outlet temperature is maintained at least 10°F below saturation temperature.

2. No RCP shall be started with any RCS cold leg temperature _*the Over Pressure Protection System (OPPS) enable temperature specified in the PTLR unless:

a. The secondary side water temperature of each steam generator (SG) is _*50'F above each of the RCS cold leg temperatures; or

b. There is a steam or gas bubble in the pressurizer.

APPLICABILTY: MODE 4.

Prairie Island Units 1 and 2 3.4.6-1 1/2/02

RCS Loops - MODE 4 3.4.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required loop A. 1 Initiate action to restore a Immediately inoperable. second loop to OPERABLE status.

AND A.2 --------- NOTE------

Only required if RHR loop is OPERABLE.

Be in MODE 5. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months B. Both loops inoperable. B.1 Suspend operations that Immediately OR would cause introduction into the RCS, coolant with boron concentration less operationr than required to meet SDM of LCO 3.1.1.

AND B.2 Initiate action to restore one Immediately loop to OPERABLE status and operation.

Prairie Island Units 1 and 2 3.4.6-2 1/2/02

RCS Loops - MODE 4 3.4.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify required RHR or RCS loop is in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months operation.

SR 3.4.6.2 Verify required SG capable of removing decay 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months heat.

SR 3.4.6.3 ------------------- NOTE --------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

Verify correct breaker alignment and indicated 7 days power are available to each required pump.

Prairie Island Units 1 and 2 3.4.6-3 1/2/02

RCS Loops - MODE 5, Loops Filled 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Loops - MODE 5, Loops Filled LCO 3.4.7 One residual heat removal (RHR) loop shall be OPERABLE and in operation, and either:

a. One additional RHR loop shall be OPERABLE; or

b. One steam generator (SG) shall be capable of removing decay heat.

NOTES --------------------

1. The RHR pump of the loop in operation may be de-energized for *< 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a. No operations are permitted that would cause introduction into the RCS, coolant with boron concentration less than required to meet the SDM of LCO 3.1.1; and

b. Core outlet temperature is maintained at least 10°F below saturation temperature.

2. One required RHR loop may be inoperable for _ 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided that the other RHR loop is OPERABLE and in operation.

3. No reactor coolant pump shall be started with one or more RCS cold leg temperatures _*the Over Pressure Protection System (OPPS) enable temperature specified in the PTLR unless:

a. The secondary side water temperature of each SG is _ 50'F above each of the RCS cold leg temperatures; or

b. There is a steam or gas bubble in the pressurizer.

4. Both RHR loops may be removed from operation during planned heatup to MODE 4 when at least one RCS loop is in operation.

Prairie Island Units 1 and 2 3.4.7-1 1/2/02

RCS Loops- MODE 5, Loops Filled

3.4.7 APPLICABILITY

MODE 5 with RCS loops filled.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required RHR loop A. 1 Initiate action to restore a Immediately inoperable, second RHR loop to OPERABLE status.

AND OR One RHR loop OPERABLE. A.2 Initiate action to restore Immediately required SG capable to remove decay heat.

B. One or more SGs not B.1 Initiate action to restore a Immediately capable of decay heat second RHR loop to removal. OPERABLE status.

AND OR One RHR loop B.2 Initiate action to restore a Immediately OPERABLE. required SG capable to remove decay heat.

Prairie Island Units 1 and 2 3.4.7-2 1/2/02

RCS Loops - MODE 5, Loops Filled 3.4.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. No required RHR loops C.1 Suspend operations that Immediately OPERABLE. would cause introduction into the RCS, coolant with AND boron concentration less than required to meet SDM Required RHR loop not of LCO 3.1.1.

in operation.

AND C.2 Initiate action to restore Immediately one RHR loop to OPERABLE status and operation.

Prairie Island Units 1 and 2 3.4.7-3 1/2/02

RCS Loops - MODE 5, Loops Filled 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1 Verify required RHR loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.7.2 Verify required SG capable of removing decay 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months heat.

SR 3.4.7.3 ------------------- NOTE --------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

Verify correct breaker alignment and indicated 7 days power are available to each required RHR pump.

Prairie Island Units 1 and 2 3.4.7-4 1/2/02

RCS Loops - MODE 5, Loops Not Filled 3.4.8 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.8 RCS Loops - MODE 5, Loops Not Filled LCO 3.4.8 Two residual heat removal (RHR) loops shall be OPERABLE and one RHR loop shall be in operation.

-NOTES.

1. All RHR pumps may be de-energized for _<1 hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a. No operations are permitted that would cause introduction into the RCS, coolant with boron concentration less than required to meet the SDM of LCO 3.1.1;

b. The core outlet temperature is maintained > 10°F below saturation temperature; and

c. No draining operations to further reduce the RCS water volume are permitted.

2. One RHR loop may be inoperable for _<2 hours for surveillance testing provided that the other RHR loop is OPERABLE and in operation.

APPLICABILITY: MODE 5 with RCS loops not filled.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required RHR loop A. 1 Initiate action to restore Immediately inoperable. RHR loop to OPERABLE status.

Prairie Island Units 1 and 2 3.4.8-1 1/2/02

RCS Loops - MODE 5, Loops Not Filled 3.4.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. No required RHR loop B.1 Suspend operations that Immediately OPERABLE. would cause introduction into the RCS, coolant with OR boron concentration less than required to meet SDM Required RHR loop not in of LCO 3.1.1.

operation.

AND NOTE --------

A Safety Injection pump may be run as required to maintain adequate core cooling and RCS inventory.

B.2 Initiate action to restore one Immediately RHR loop to OPERABLE status and operation.

Prairie Island Units 1 and 2 3.4.8-2 1/2/02

RCS Loops - MODE 5, Loops Not Filled 3.4.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify required RHR loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.8.2 ------------------ NOTE ---------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

Verify correct breaker alignment and indicated 7 days power are available to each required RHR pump.

Prairie Island Units 1 and 2 3.4.8-3 1/2/02

Pressurizer Safety Valves 3.4.10 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND OR B.2 Be in MODE 4 with any 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months RCS cold leg temperature Both pressurizer safety _ the OPPS enable valves inoperable, temperature specified in the PTLR.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each pressurizer safety valve is In accordance OPERABLE in accordance with the Inservice with the Testing Program. Following testing, lift settings Inservice Testing shall be within 1% (2460 to 2510 psig). Program Prairie Island Units 1 and 2 3.4.10-2 1/2/02

LTOP - RCSCLT > SI Pump Disable Temperature 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Low Temperature Overpressure Protection (LTOP) - Reactor Coolant System Cold Leg Temperature (RCSCLT) > Safety Injection (SI) Pump Disable Temperature LCO 3.4.12 LTOP shall be provided with:

a. a maximum of one SI pump capable of injecting into the RCS;

b. the emergency core cooling system (ECCS) accumulators isolated;

c. an OPERABLE Over Pressure Protection System (OPPS); and

d. two OPERABLE pressurizer power operated relief valves (PORVs) with lift settings within the limits specified in the PTLR.

NOTES

1. Both SI pumps may be run for _<1 hour while conducting SI system testing providing there is a steam or gas bubble in the pressurizer and at least one isolation valve between the SI pump and the RCS is shut.

2. ECCS accumulator may be unisolated when accumulator pressure is less than the maximum RCS pressure for the existing RCS cold leg temperature allowed by the P/T limit curves provided in the PTLR.

APPLICABILITY: MODE 4 when any RCS cold leg temperature is < the OPPS enable temperature specified in the PTLR and > the SI pump disable temperature specified in the PTLR.

Prairie Island Units 1 and 2 3.4.12-1 1/2/02

LTOP - RCSCLT > SI Pump Disable Temperature 3.4.12 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Two SI pumps capable of A.1 Initiate action to assure a Immediately injecting into the RCS. maximum of one SI pump is capable of injecting into the RCS.

B. An ECCS accumulator B.1 Isolate affected ECCS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months not isolated when the accumulator.

ECCS accumulator pressure is greater than or equal to the maximum RCS pressure for existing cold leg temperature allowed in the PTLR.

C. Required Action and C.1 Increase RCS cold leg 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion temperature to > the OPPS Time of Condition B not enable temperature met. specified in the PTLR.

OR C.2 Depressurize affected 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ECCS accumulator to less than the maximum RCS pressure for existing cold leg temperature allowed in the PTLR.

D. One required PORV D.1 Restore required PORV to 7 days inoperable. OPERABLE status.

Prairie Island Units 1 and 2 3.4.12-2 1/2/02

LTOP - RCSCLT > SI Pump Disable Temperature 3.4.12 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Two PORVs inoperable. E.1 Be in MODE 5. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months OR AND Required Action and E.2 Depressurize RCS and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion establish RCS vent of> 3 Time of Condition A, C, square inches.

or D not met.

OR OPPS inoperable.

Prairie Island Units 1 and 2 3-.4.12-3 1/2/02

LTOP - RCSCLT > SI Pump Disable Temperature 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Verify a maximum of one SI pump is capable of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months injecting into the RCS.

SR 3.4.12.2 ------------------ NOTE --------------

Only required to be performed when ECCS accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR.

Verify each ECCS accumulator is isolated. Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter SR 3.4.12.3 Verify PORV block valve is open for each 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months required PORV.

SR 3.4.12.4 ------------------ NOTE --------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after decreasing RCS cold leg temperature to _*the OPPS enable temperature specified in the PTLR.

Perform a COT on OPPS. 31 days SR 3.4.12.5 Perform CHANNEL CALIBRATION for each 24 months OPPS actuation channel.

Prairie Island Units 1 and 2 3.4.12-4 1!/2/02

LTOP - RCSCLT _*SI Pump Disable Temperature 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 Low Temperature Overpressure Protection (LTOP) - Reactor Coolant System Cold Leg Temperature (RCSCLT) _<Safety Injection (SI) Pump Disable Temperature LCO 3.4.13 LTOP shall be provided with: 1) no SI Pumps capable of injecting into the RCS; 2) the emergency core cooling system (ECCS) accumulators isolated; and 3) one of the following pressure relief capabilities:

a. An Over Pressure Protection System (OPPS) shall be OPERABLE with two pressurizer power operated relief valves (PORVs) with lift settings within the limits specified in the PTLR; or

b. The RCS depressurized and an RCS vent of >_3 square inches.

NOTES

1. Both safety injection (SI) pumps may be run for *< 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months while conducting SI system testing provided there is a steam or gas bubble in the pressurizer, the reactor vessel head is on, and at least one isolation valve between the SI pump and the RCS is shut.

2. During reduced inventory conditions an SI pump may be run as required to maintain adequate core cooling and RCS inventory.

3. ECCS accumulator may be unisolated when ECCS accumulator pressure is less than the maximum RCS pressure for the existing RCS cold leg temperature allowed by the P/T limit curves provided in the PTLR.

APPLICABILITY: MODE 4 when any RCS cold leg temperature is _*the SI Pump disable temperature specified in the PTLR, MODE 5 when the steam generator (SG) primary system manway and pressurizer manway are closed and secured in position, MODE 6 when the reactor vessel head is on and the SG primary system manway and pressurizer manways are closed and secured in position.

Prairie Island Units 1 and 2 3.4.13-1 1/2/02

LTOP - RCSCLT _!SI Pump Disable Temperature 3.4.13 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or both SI pump(s) A. 1 Initiate action to verify no SI Immediately capable of injecting into pump is capable of injecting the RCS. into the RCS.

B. An ECCS accumulator B.l Isolate affected ECCS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months not isolated when the accumulator.

ECCS accumulator pressure is greater than or equal to the maximum RCS pressure for existing cold leg temperature allowed in the PTLR.

C. Required Action and C.l Increase RCS cold leg 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion temperature to > the OPPS Time of Condition B not enable temperature specified met. in the PTLR.

OR C.2 Depressurize affected ECCS 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months accumulator to less than the maximum RCS pressure for existing cold leg temperature allowed in the PTLR.

Prairie Island Units 1 and 2 3.4.13-2 1/2/02

LTOP - RCSCLT _<SI Pump Disable Temperature 3.4.13 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. ------- NOTE --------- D.1 Restore required PORV to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Only applicable in LCO OPERABLE status.

3.4.13.a.

One required PORV inoperable.

E. Two required PORVs E.1 Depressurize RCS and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months inoperable for LCO establish RCS vent of > 3 3.4.13.a. square inches.

OR Required Action and associated Completion Time of Condition A, C, or D not met.

OR OPPS inoperable.

Prairie Island Units 1 and 2 3.4.13-3 1/2/02

LTOP - RCSCLT < SI Pump Disable Temperature 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 Verify no SI pumps are capable of injecting into 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months the RCS.

SR 3.4.13.2 ------------------ NOTE --------------

Only required to be performed when ECCS accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR.

Verify each ECCS accumulator is isolated. Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter SR 3.4.13.3 ----------------- NOTE ------------------ 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for Only required to be performed when complying unlocked open with LCO 3.4.13.b.Verify required RCS vent _>3 vent valve(s) square inches open.

. AN D Verify required RCS vent _> 3 square inches open. 31 days for other vent path(s)

SR 3.4.13.4 Verify PORV block valve is open for each 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months required PORV.

Prairie Island Units 1 and 2 3.4.13-4 1/2/02

LTOP - RCSCLT _ SI Pump Disable Temperature 3.4.13 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.13.5 ----------------- NOTE ---------------

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

< the OPPS enable temperature specified in the PTLR.

2. Only required to be performed when complying with LCO 3.4.13.a.

Perform a COT on OPPS. 31 days SR 3.4.13.6 Perform CHANNEL CALIBRATION for OPPS 24 months actuation channel.

Prairie Island Units 1 and 2 3.4.13-5 1/2/02

RCS Operational LEAKAGE 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 ---------------------------- NOTE ---------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify RCS operational leakage within limits by 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months performance of RCS water inventory balance.

SR 3.4.14.2 Verify steam generator tube integrity is in In accordance accordance with the Steam Generator Program. with the Steam Generator Program Prairie Island Units 1 and 2 3.4.14-3 1/2/02

RCS Specific Activity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 RCS Specific Activity LCO 3.4.17 The specific activity of the reactor coolant shall be within limits:

APPLICABILITY: MODES 1 and 2, MODE 3 with RCS average temperature (Tavg) >-500'F.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT ------------ Note ---------

1-131 > 1.0 /Ci/gm. LCO 3.0.4 is not applicable.

A. 1 Verify DOSE Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months EQUIVALENT 1-131 within the acceptable region of Figure 3.4.17-1.

AND A.2 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT 1-131 to within limit.

Prairie Island Units 1 and 2 3.4.17-1 1/2/02

RCS Specific Activity 3.4.17 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Gross specific activity B.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months of the reactor coolant Tavg < 500 F.

not within limit.

C. Required Action and C.l Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Tavg < 500 0F.

Time of Condition A not met.

OR DOSE EQUIVALENT 1-131 in the unacceptable region of Figure 3.4.17-1.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify reactor coolant gross specific activity 7 days S100/I .uCi/gm.

Prairie Island Units 1 and 2 3.4.17-2 1/2/02

RCS Specific Activity 3.4.17 SURVEILLANCE REQUIREMENTS (continued) I SURVEILLANCE FREQUENCY i

SR 3.4.17.2 NOTE ---------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 14 days 1-131 specific activity _*1.0 /,Ci/gm.

AND Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change of >_15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period

+/-

SR 3.4.17.3 NOTE ---------------

Not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for _>48 hours.

Determine - from a sample taken in MODE 1 after 184 days a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for _>48 hours.

Prairie Island Units 1 and 2 3.4.17-3 1/2/02

RCS Pressure, Temperature, and Flow - DNB Limits B 3.4.1 BASES SURVEILLANCE SR 3.4.1.2 REQUIREMENTS (continued) Since Required Action A. 1 allows a Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to restore parameters that are not within limits, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Surveillance Frequency for RCS average temperature is sufficient to ensure the temperature can be restored to a normal operation, steady state condition following load changes and other expected transient operations. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions.

SR 3.4.1.3 Measurement of RCS total flow rate once every 24 months allows the installed RCS flow instrumentation to be calibrated and verifies the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate. This verification may be performed via a precision calorimetric heat balance or other means.

The Frequency of 24 months reflects the importance of verifying flow after a refueling outage when the core has been altered, which may have caused an alteration of flow resistance.

REFERENCES 1. USAR, Section 14.

Prairie Island Units 1 and 2 B 3.4.1-5 1/2/02

RCS P/T Limits B 3.4.3 BASES (continued)

ACTIONS A.1 and A.2 Operation outside the P/T limits during MODE 1, 2, 3, or 4 must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses.

The 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.

Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify the RCPB integrity remains acceptable and must be completed before continuing operation. Several methods may be used, including an engineering evaluation to determine effects of the out of-limit condition on the structural integrity of the RCS, a comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.

ASME Code,Section XI, Appendix E, may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable to accomplish the evaluation. The evaluation for a mild violation is possible within this time, but more severe violations may require special, event specific stress analyses or inspections. A favorable evaluation must be completed before continuing to operate.

Condition A is modified by a Note requiring that Required Action A.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action A. 1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

Prairie Island Units 1 and 2 B 3.4.3-5 1/2/02

RCS Loops - MODE 3 B 3.4.5 BASES LCO The Note permits both RCPs to be de-energized for (continued) _*12 hours to perform preplanned work activities.

One purpose of the Note is to allow performance of tests that are designed to validate various accident analyses values. One of these tests is validation of the pump coastdown curve used as input to a number of accident analyses including a loss of flow accident. This test was performed during the initial startup testing program, and would normally only be performed once. If, however, changes are made to the RCS that would cause a change to the flow characteristics of the RCS, the input values of the coastdown curve must be revalidated by conducting the test again. Another test performed during the startup testing program was the validation of rod drop times, both with and without flow. Any future no flow test may be performed in MODE 3, 4, or 5 and requires that the pumps be stopped for a short period of time. The Note permits stopping the pumps in order to perform this test and validate the assumed analysis values.

Another purpose of the Note is to allow stopping of both RCP's for a sufficient time to perform station electrical lineup changes without transition to MODE 4. During these evolutions both RCP's may be inoperable. Transition to MODE 4 would put the plant through unnecessary cooldown and heatup transients. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months time period specified is adequate to perform the necessary load shedding, switching and load restoration activities and restart an RCP without requiring transition to MODE 4.

Utilization of the Note is permitted provided the following conditions are met:

Prairie Island Units 1 and 2 B 3.4.5-3 1/2/02

RCS Loops - MODE 3 B 3.4.5 BASES LCO a. No operations are permitted that would dilute the RCS boron (continued) concentration with coolant with boron concentration less than required to meet SDM of LCO 3.1.1, thereby maintaining the margin to criticality. Boron reduction with coolant at boron concentrations less than required to assure SDM is maintained is prohibited to preclude the need for a boration, due to the time required to achieve a uniform distribution when in natural circulation (Ref. 1); and

b. Core outlet temperature is maintained at least 10°F below saturation temperature, so that no vapor bubble may form and possibly cause a natural circulation flow obstruction.

An OPERABLE RCS loop consists of one OPERABLE RCP and one OPERABLE SG which is capable of removing decay heat as specified in SR 3.4.5.2. An RCP is OPERABLE if it is capable of being powered and is able to provide forced flow if required.

APPLICABILITY In MODE 3, this LCO ensures forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. The most stringent condition of the LCO, that is, two RCS loops OPERABLE and two RCS loops in operation, applies to MODE 3 with the Rod Control System capable of rod withdrawal.

The least stringent condition, that is, two RCS loops OPERABLE and one RCS loop in operation, applies to Mode 3 with the Rod Control System not capable of rod withdrawal.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops-MODES 1 and 2";

LCO 3.4.6, "RCS Loops-MODE 4";

LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled";

LCO 3.4.8, "RCS Loops-MODE 5, Loops Not Filled";

Prairie Island Units 1 and 2 B 3.4.5-4 1/2/02

RCS Loops - MODE 3 B 3.4.5 BASES ACTIONS D.1, D.2, and D.3 (continued) coolant could be introduced to the core; however, coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations. The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.

SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required loops are in operation. Verification may include flow rate, temperature, or pump status monitoring, which helps ensure that forced flow is providing heat removal. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS loop performance.

SR 3.4.5.2 SR 3.4.5.2 requires verification that the SG has the capability to remove decay heat. The ability to remove decay heat requires the ability to pressurize and control pressure in the RCS, sufficient secondary side water level in the SG relied on for decay heat removal, and an available supply of feedwater (Ref. 2). The ability of the SG to provide an adequate heat sink for decay heat removal further ensures that the SG tubes remain covered.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of the other indications available in the control room to alert the operator to a loss of the SG to remove decay heat.

Prairie Island Units 1 and 2 B 3.4.5-7 1/2/02

RCS Loops - MODE 3 B 3.4.5 BASES SURVEILLANCE SR 3.4.5.3 REQUIREMENTS (continued) Verification that each required RCP is OPERABLE ensures that an additional RCP can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power availability to each required RCP. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability.

This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a pump is not in operation.

REFERENCES 1. License Amendment Request Dated November 19, 1999.

(Approved by License Amendment 152/143, July 14, 2000.)

2. NRC Information Notice 95-35, "Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation."

Prairie Island Units 1 and 2 B 3.4.5-8 1/2/02

RCS Loops - MODE 4 B 3.4.6 BASES (continued)

LCO The purpose of this LCO is to require that at least two loops be OPERABLE in MODE 4 and that one of these loops be in operation.

The LCO allows the two loops that are required to be OPERABLE to consist of any combination of RCS loops and RHR loops. Any one loop in operation provides enough flow to remove the decay heat from the core with forced circulation. An additional loop is required to be OPERABLE to provide redundancy for heat removal.

Note 1 permits all RCPs or RIR pumps to be de-energized for

_* 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of the Note is to permit tests that are designed to validate various accident analyses values.

One of the LCO tests performed during the startup testing program was validation of rod drop times during cold conditions, both with and without flow. If changes are made to the RCS that would cause a change in flow characteristics of the RCS, the input values must be revalidated by conducting the test again. Any future no flow test may be performed in MODE 3, 4, or 5 and requires that the pumps be stopped for a short period of time. The Note permits stopping the pumps in order to perform this test and validate the assumed analysis values. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time period is adequate to perform the test, and operating experience has shown that boron stratification is not a problem during this short period with no forced flow.

Utilization of Note 1 is permitted provided the following conditions are met along with any other conditions imposed by startup test procedures:

a. No operations are permitted that would dilute the RCS boron concentration with coolant with boron concentration less than required to meet SDM of LCO 3.1.1, therefore maintaining the margin to criticality. Boron reduction with coolant at boron concentrations less than required to assure SDM in maintained is prohibited to preclude the need for a boration, due to the time required to achieve a uniform distribution when in natural circulation (Ref. 1); and Prairie Island Units 1 and 2 B 3.4.6-2 1/2/02

RCS Loops - MODE 4 B 3.4.6 BASES LCO b. Core outlet temperature is maintained at least 10°F below (continued) saturation temperature, so that no vapor bubble may form and possibly cause a natural circulation flow obstruction.

Note 2 requires a steam or gas bubble in the pressurizer or that the secondary side water temperature of each SG be _*50'F above each of the RCS cold leg temperatures before the start of an RCP with any RCS cold leg temperature _ the OPPS enable temperature specified in the PTLR. A steam or gas bubble ensures that the pressurizer will accommodate the swell resulting from an RCP start. Either of these restraints prevents a low temperature overpressure event due to a thermal transient when an RCP is started.

An OPERABLE RCS loop consists of an OPERABLE RCP and an OPERABLE SG which is capable of removing decay heat as specified in SR 3.4.6.2.

Similarly for the RHR System, an OPERABLE RHR loop consists of an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat exchanger. RCPs and RHR pumps are OPERABLE if they are capable of being powered and are able to provide forced flow if required.

APPLICABIrITY In MODE 4, this LCO ensures forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops- MODES 1 and 2";

LCO 3.4.5, "RCS Loops- MODE 3";

LCO 3.4.7, "RCS Loops- MODE 5, Loops Filled";

LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled";

LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level" (MODE 6); and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level" (MODE 6).

Prairie Island Units 1 and 2 B 3.4.6-3 1/2/02

RCS Loops - MODE 4 B 3.4.6 BASES (continued)

ACTIONS A.1 If one required loop is inoperable, redundancy for heat removal is lost. Action must be initiated to restore a second RCS or RHR loop to OPERABLE status. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal. Entry to a reduced MODE (MODE 5 or 6) requires RIR availability for long term decay heat removal. Remaining in MODE 4, with RCS loop operation, is conservative.

If restoration is not accomplished and an RHR Loop is OPERABLE, the unit must be brought to MODE 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Bringing the unit to MODE 5 is a conservative action with regard to decay heat removal. With only one RHR loop OPERABLE, redundancy for decay heat removal is lost and, in the event of a loss of the remaining RHR loop, it would be safer to initiate that loss from MODE 5 rather than MODE 4. The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable time, based on operating experience, to reach MODE 5 from MODE 4 in an orderly manner and without challenging plant systems.

The Required Action is modified by a Note which indicates that the unit must be placed in MODE 5 only if a RHR loop is OPERABLE.

With no RHR loop OPERABLE, the unit is in a condition with only limited cooldown capabilities. Therefore, the actions are to be concentrated on the restoration of a RHR loop, rather than a cooldown of extended duration.

Prairie Island Units 1 and 2 B 3.4.6-4 1/2/02

RCS Loops - MODE 4 B 3.4.6 BASES ACTIONS B.1 and B.2 (continued)

If both loops are inoperable or a required loop not in operation, except during conditions permitted by Note 1 in the LCO section, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 must be suspended and action to restore one RCS or RHR loop to OPERABLE status and operation must be initiated. The margin to criticality must not be reduced in this type of operation. Suspending the introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core; however, coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations. The immediate Completion Times reflect the importance of maintaining operation for decay heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.

SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required RCS or RHR loop is in operation. Verification may include flow rate, temperature, or pump status monitoring, which helps ensure that forced flow is providing heat removal. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS and RHR loop performance.

Prairie Island Units 1 and 2 B 3.4.6-5 1/2/02

RCS Loops - MODE 4 B 3.4.6 BASES SURVEILLANCE SR 3.4.6.2 REQUIREMENTS (continued) SR 3.4.6.2 requires verification that the required SG has the capability to remove decay heat. The ability to remove decay heat requires the ability to pressurize and control pressure in the RCS, sufficient secondary side water level in the SG relied on for decay heat removal, and an available supply of feedwater (Ref. 2). The ability of the SG to provide an adequate heat sink for decay heat removal further ensures that the SG tubes remain covered. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of the other indications available in the control room to alert the operator to a loss of capability of the SG to remove decay heat.

SR 3.4.6.3 Verification that each required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.

Verification is performed by verifying proper breaker alignment and power available to each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.

This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a pump is not in operation.

Prairie Island Units 1 and 2 B 3.4.6-6 1/2/02

RCS Loops - MODE 4 B 3.4.6 BASES (continued)

REFERENCES 1. License Amendment Request Dated November 19, 1999.

(Approved by License Amendment 152/143, July 14, 2000.)

2. NRC Information Notice 95-35, "Degraded Ability of Steam Generator to Remove Decay.Heat by Natural Circulation."

Prairie Island Units 1 and 2 B 3.4.6-7 1/2/02

RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES BACKGROUND OPERABLE RHR loop or maintaining a SG capable of removing (continued) decay heat to provide an alternate method for decay heat removal via natural circulation.

APPLICABLE In MODE 5, RCS circulation increases the time available for SAFETY mitigation of an accidental boron dilution event. The RHR ANALYSES loops provide this circulation.

RCS Loops - MODE 5 (Loops Filled) satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO The purpose of this LCO is to require that at least one RHR loop be OPERABLE and in operation with an additional RHR loop OPERABLE or a SG capable of removing decay heat via natural circulation. One RHR loop provides sufficient forced circulation to perform the safety functions of the reactor coolant under these conditions. An additional RHR loop is required to be OPERABLE to provide redundancy. However, if the standby RHR loop is not OPERABLE, an acceptable alternate method is a SG. Should the operating RHR loop fail, the SG could be used to remove decay heat via natural circulation.

Note 1 permits all RHR pumps to be de-energized < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of the Note is to permit tests designed to validate various accident analyses values. One of the tests performed during the startup testing program was validation of rod drop times during cold conditions, both with and without flow. If changes are made to the RCS that would cause a change in flow characteristics of the RCS, the input values must be revalidated by conducting the test again. Any future no flow test may be performed in MODE 3, 4, or 5 and requires that the pumps be stopped for a short period of time. The Note permits stopping the pumps in order to perform this test and validate the assumed analysis values. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time period is adequate to perform the test, and operating Prairie Island Units 1 and 2 B 3.4.7-2 1/2/02

RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES LCO Note 4 provides for an orderly transition from MODE 5 to MODE 4 (continued) during a planned heatup by permitting removal of RHR loops from operation when at least one RCS loop is in operation. This Note provides for the transition to MODE 4 where an RCS loop is permitted to be in operation and replaces the RCS circulation function provided by the RHR loops.

RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required. A SG is capable of removing decay heat via natural circulation when: 1) there is the ability to pressurize and control pressure in the RCS; 2) there is sufficient secondary side water level in the SG relied on for decay heat removal; and 3) there is an available supply of feedwater (Ref.

1). An OPERABLE SG can perform as a heat sink via natural circulation when it has the capability to remove decay heat as specified in SR 3.4.7.2.

APPLICABILITY In MODE 5 with RCS loops filled, this LCO requires forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. One loop of RHR provides sufficient circulation for these purposes. However, one additional RHR loop is required to be OPERABLE, or a SG is capable of removing decay heat.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops-MODES 1 and 2";

LCO 3.4.5, "RCS Loops-MODE 3";

LCO 3.4.6, "RCS Loops-MODE 4";

LCO 3.4.8, "RCS Loops-MODE 5, Loops Not Filled";

LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (MODE 6).

Prairie Island Units 1 and 2 B 3.4.7-4 1/2/02

RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES (continued)

ACTIONS A. 1, A.2, B. 1 and B.2 If one RHR loop is OPERABLE and the SGs are not capable of removing decay heat , redundancy, for heat removal is lost. Action must be initiated immediately to restore a second RHR loop to OPERABLE status or to restore the required SG capability to remove decay heat. Either Required Action will restore redundant heat removal paths. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

C.1 and C.2 If a required RHR loop is not in operation, except during conditions permitted by Note 1, or if no loop is OPERABLE, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 must be suspended and action to restore one RHR loop to OPERABLE status and operation must be initiated. Suspending the introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core; however, coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations. The immediate Completion Times reflect the importance of maintaining operation for heat removal.

Prairie Island Units 1 and 2 B 3.4.7-5 1/2/02

RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES (continued)

SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required loop is in operation. Verification may include flow rate, temperature, or pump status monitoring, which helps ensure that forced flow is providing heat removal. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RHR loop performance.

SR 3.4.7.2 SR 3.4.7.2 requires verification that the required SG has the capability to remove decay heat via natural circulation. This provides an alternate decay heat removal method in the event that the second RHR loop is not OPERABLE. The ability to remove decay heat requires the ability to pressurize and control pressure in the RCS, sufficient secondary side water level in the SG relied on for decay heat removal, and an available supply of feedwater(Ref. 1).

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available in the control room to alert the operator to a loss of capability of the SG to remove decay heat.

SR 3.4.7.3 Verification that each required RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.

Verification is performed by verifying proper breaker alignment and power available to each required RHR pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. If at least one SG is capable of decay heat removal, this Surveillance is not needed. The Frequency Prairie Island Units 1 and 2 B 3.4.7-6 1/2/02

RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE SR 3.4.7.3 (continued)

REQUIRMENTS of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.

This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a pump is not in operation.

REFERENCES 1. NRC Information Notice 95-35, "Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation".

2. License Amendment Request Dated November 19, 1999.

(Approved by License Amendment 152/143, July 14, 2000.)

Prairie Island Units 1 and 2 B 3.4.7-7 1/2/02

RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES LCO Note 1 permits all RHR pumps to be de-energized _*1 hour per 8 (continued) hour period. The circumstances for stopping both RHR pumps are to be limited to situations when the outage time is short and core outlet temperature is maintained > 10 F below saturation temperature. The Note prohibits boron dilution with coolant at boron concentrations less than required to assure SDM is maintained or draining operations when RHR forced flow is stopped.

Note 2 allows one RHR loop to be inoperable for a period of

ý 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , provided that the other loop is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable loop during the only time when these tests are safe and possible.

An OPERABLE RHR loop is comprised of an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat exchanger. RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required.

requires core heat In MODE 5 with loops not filled, this LCO APPLICABILITY In MODE 5 with loops not filled, this LCO requires core heat removal and coolant circulation by the RHR System.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops-MODES 1 and 2";

LCO 3.4.5, "RCS Loops-MODE 3";

LCO 3.4.6, "RCS Loops-MODE 4";

LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled";

LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (MODE 6).

Prairie Island Units 1 and 2 B 3.4.8-2 1/2/02

RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES (continued)

ACTIONS A.1 If one required RHR loop is inoperable, redundancy for RHR is lost.

Action must be initiated to restore a second loop to OPERABLE status. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

B. 1 and B.2 If no required loop is OPERABLE or the required loop is not in operation, except during conditions permitted by Note 1, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 must be suspended and action must be initiated immediately to restore an RHR loop to OPERABLE status and operation. The margin to criticality must not be reduced in this type of operation.

Suspending the introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core; however, coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations. The immediate Completion Time reflects the importance of maintaining operation for heat removal.

The action to restore must continue until one loop is restored to OPERABLE status and operation.

The Note in Required Action B.2 allows the use of one safety injection pump to provide heat removal in the event of a loss of RHR system cooling during reduced RCS inventory conditions.

Prairie Island Units 1 and 2 B 3.4.8-3 1/2/02

RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES (continued)

SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required loop is in operation. Verification may include flow rate, temperature, or pump status monitoring, which helps ensure that forced flow is providing heat removal. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RHR loop performance.

SR 3.4.8.2 Verification that each required pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.

This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a pump is not in operation.

REFERENCES None.

Prairie Island Units 1 and 2 B 3.4.8-4 1/2/02

LTOP - RCSCLT > SI Pump Disable Temperature B 3.4.12 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.12 Low Temperature Overpressure Protection (LTOP) - Reactor Coolant System Cold Leg Temperature (RCSCLT) > Safety Injection (SI) Pump Disable Temperature BASES BACKGROUND The LTOP function limits RCS pressure at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is not compromised by violating the pressure and temperature (P/T) limits of 10 CFR 50, Appendix G (Ref. 1). The reactor vessel is the limiting RCPB component for demonstrating such protection. The Over Pressure Protection System (OPPS) and the pressurizer power operated relief valves (PORVs) provide the LTOP function (Ref. 2).

The PTLR provides the maximum allowable OPPS actuation setpoints for the PORVs and the maximum RCS pressure for the existing RCS cold leg temperature during cooldown, shutdown, and heatup to meet the Reference 1 requirements during the LTOP MODES. The LTOP MODES are the MODES as defined in the Applicability statement of LCO 3.4.12 and LCO 3.4.13.

The pressurizer safety valves and PORVs at their normal setpoints do not provide overpressure protection for certain low temperature operational transients. Inadvertent pressurization of the RCS at temperatures below the OPPS enable temperature specified in the PTLR could result in exceeding the ASME Appendix G (Ref. 3) brittle fracture P/T limits. Exceeding the RCS P/T limits by a significant amount could cause brittle cracking of the reactor vessel.

LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," requires administrative control of RCS pressure and temperature during heatup and cooldown to prevent exceeding the PTLR limits.

This LCO provides RCS overpressure protection by restricting coolant input capability and ensuring adequate pressure relief capacity. In MODE 4, above the safety injection (SI) pump disable temperature, limiting coolant input capability requires one (SI) pump incapable of injection into the RCS and isolating the emergency core cooling system (ECCS) accumulators. In MODE 4, above the SI Prairie Island Units 1 and 2 B 3.4.12-1 1/2/02

LTOP - RCSCLT > SI Pump Disable Temperature B 3.4.12 BASES BACKGROUND pump disable temperature, one PORV is the overpressure protection (continued) device that acts to terminate an increasing pressure event.

Limiting coolant input capability reduces the ability to provide core coolant addition. The LCO does not require the makeup control system deactivated or the SI actuation circuits blocked. Due to the lower pressures in the LTOP MODES and the expected core decay heat levels, the charging system can provide adequate flow. If conditions require the use of more than one SI pump for makeup in the event of loss of inventory, then pumps can be made available through manual actions.

In MODE 4, above the SI pump disable temperature, pressure relief consists of two PORVs with reduced lift settings. Two. PORVs are required for redundancy. One PORV has adequate relieving capability to prevent overpressurization for the required coolant input capability.

As designed for the LTOP function, each PORV is signaled to open by OPPS if the RCS pressure approaches the lift setpoint provided when OPPS is enabled. The OPPS monitors both RCS temperature and RCS pressure and indicates when a condition not acceptable in the PTLR limits is approached. The wide range RCS temperature setpoints indicate conditions requiring enabling OPPS.

The PTLR presents the OPPS setpoints for LTOP.

APPLICABLE Safety analyses (Ref. 2) demonstrate that the reactor vessel is SAFETY adequately protected against exceeding the Reference 1 P/T limits.

ANALYSES In MODES 1, 2, and 3, and in MODE 4 with RCS cold leg temperature exceeding the OPPS enable temperature specified in the PTLR, the pressurizer safety valves will prevent RCS pressure from exceeding the Reference 1 limits. At about the OPPS enable temperature specified in the PTLR and below, overpressure prevention falls to two OPERABLE PORVs or to a depressurized Prairie Island Units 1 and 2 B 3.4.12-2 1/2/02

LTOP - RCSCLT > SI Pump Disable Temperature B 3.4.12 BASES APPLICABLE RCS and a sufficiently sized RCS vent. Each of these means has a SAFETY limited overpressure relief capability. LCO 3.4.13, "LTOP *< SI ANALYSES Pump Disable Temperature," provides the requirements for (continued) overpressure prevention at the lower temperatures.

The actual temperature at which the pressure in the P/T limit curve falls below the pressurizer safety valve setpoint increases as the reactor vessel material toughness decreases due to neutron embrittlement. Each time the PTLR curves are revised, LTOP must be re-evaluated to ensure its functional requirements can still be met using the PORV method.

The PTLR contains the acceptance limits that define the LTOP requirements. Any change to the RCS must be evaluated against the Reference 2 analyses to determine the impact of the change on the LTOP acceptance limits.

Transients that are capable of overpressurizing the RCS are categorized as either mass or heat input transients. The bounding mass input transient is inadvertent safety injection with injection from one SI pump and three charging pumps, and letdown isolated.

The bounding heat input transient is reactor coolant pump (RCP) startup with temperature asymmetry within the RCS or between the RCS and steam generators.

The following limitations are required during the Applicability of this specification to ensure that mass and heat input transients in excess of analysis assumptions do not occur:

a. Rendering one SI pump incapable of injection;

b. Deactivating the ECCS accumulator discharge isolation valves in their closed positions; and

c. Disallowing start of an RCP if secondary temperature is more than 50'F above primary temperature in any one loop.

LCO 3.4.6, "RCS Loops - MODE 4," provides this protection.

Prairie Island Units 1 and 2 B 3.4.12-3 1/2/02

LTOP - RCSCLT > SI Pump Disable Temperature B 3.4.12 BASES APPLICABLE The Reference 2 analyses demonstrate that one PORV can maintain SAFETY RCS pressure below limits when only one SI pump and all charging ANALYSES pumps are actuated. Thus, the LCO allows only one SI pump (continued) OPERABLE during the Applicability of this specification.

Since one PORV cannot handle the pressure transient resulting from ECCS accumulator injection, when RCS temperature is low, the LCO also requires ECCS accumulator isolation when ECCS accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR.

The isolated ECCS accumulators must have their discharge valves closed and the valve power supply breakers fixed in their open positions.

Fracture mechanics analyses established the temperature of LTOP Applicability at the OPPS enable temperature specified in the PTLR.

The consequences of a small break loss of coolant accident (LOCA) in LTOP MODE 4 above the SI Pump disable temperature conform to 10 CFR 50.46 and 10 CFR 50, Appendix K, requirements by having a maximum of one SI pump OPERABLE and SI actuation enabled.

The fracture mechanics analyses show that the vessel is protected when the PORVs are set to open at or below the limit shown in the PTLR. The OPPS setpoints are derived by analyses that model the performance of the system, assuming the limiting LTOP transient of one SI pump and all charging pumps injecting into the RCS. These analyses consider pressure overshoot and undershoot beyond the PORV opening and closing, resulting from signal processing and valve stroke times. The OPPS setpoints at or below the derived limit ensures the Reference 1 P/T limits will be met.

The OPPS setpoints in the PTLR will be updated when the revised P/T limits conflict with the LTOP analysis limits. The P/T limits are Prairie Island Units 1 and 2 B 3.4.12-4 1/2/02

LTOP - RCSCLT > SI Pump Disable Temperature B 3.4.12 BASES APPLICABLE periodically modified as the reactor vessel material toughness SAFETY decreases due to neutron embrittlement caused by neutron ANALYSES irradiation. Revised limits are determined using neutron fluence (continued) projections and the results of examinations of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," discuss these examinations.

The LTOP function satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO This LCO requires that LTOP be provided, by limiting coolant input capability and by OPERABLE pressure relief capability. Violation of this LCO could lead to the loss of low temperature overpressure mitigation and violation of the Reference 1 limits as a result of an operational transient.

To limit the coolant input capability, the LCO requires that a maximum of one SI pump be capable of injecting into the RCS, and all ECCS accumulator discharge isolation valves be closed and deenergized (when ECCS accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR).

The LCO is modified by two Notes. Note 1 allows operation of both SI pumps for _ 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for conducting SI system testing providing there is a steam or gas bubble in the pressurizer and at least one isolation valve between the SI pump and the RCS is shut. The purpose of this note is to permit the conduct of the integrated SI test and other SI system tests and operations that may be performed in MODE 4. 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 to the RCS. This isolation is accomplished by either a closed manual valve or motor operated valve with the power removed. This combination of conditions under strict administrative control assure that overpressurization cannot occur.

Prairie Island Units 1 and 2 B 3.4.12-5 1/2/02

LTOP - RCSCLT > SI Pump Disable Temperature B 3.4.12 BASES LCO Note 2 states that ECCS accumulator isolation is only required (continued) when the ECCS accumulator pressure is more than or at the maximum RCS pressure for the existing RCS cold leg temperature allowed by the P/T limit curves provided in the PTLR (less allowance for instrument uncertainty). This Note permits the ECCS accumulator discharge isolation valve Surveillance to be performed only under these pressure and temperature conditions.

To provide low temperature overpressure mitigation through pressure relief, the LCO requires an OPERABLE OPPS with two pressurizer PORVs. A PORV is OPERABLE for LTOP when its block valve is open, its low pressure lift setpoint has been selected (OPPS enabled), and the backup air supply is charged.

APPLICABILITY This LCO is applicable in MODE 4 when any RCS cold leg temperature is *_ the OPPS enable temperature specified in the PTLR and > the SI Pump disable temperature specified in the PTLR. The pressurizer safety valves provide overpressure protection that meets the Reference 1 P/T limits above the OPPS enable temperature specified in the PTLR.

LCO 3.4.3 provides the operational P/T limits for all MODES.

LCO 3.4.10, "Pressurizer Safety Valves," requires the OPERABILITY of the pressurizer safety valves that provide overpressure protection during MODES 1, 2, and 3, and MODE 4 above the OPPS enable temperature specified in the PTLR. LCO 3.4.13 provides the LTOP-requirements in MODE 4 _<SI pump disable temperature and in MODES 5 and 6.

Low temperature overpressure prevention is most critical during shutdown when the RCS is water solid, and a mass or heat input transient can cause a very rapid increase in RCS pressure when little or no time allows operator action to mitigate the event.

Prairie Island Units 1 and 2 B 3.4.12-6 1/2/02

LTOP - RCSCLT > SI Pump Disable Temperature B 3.4.12 BASES (continued)

ACTIONS A.1 With two SI pumps capable of injecting into the RCS, RCS overpressurization is possible.

To immediately initiate action to restore restricted coolant input capability to the RCS reflects the urgency of removing the RCS from this condition.

B.1, C.1, and C.2 An unisolated ECCS accumulator requires isolation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

This is only required when the ECCS accumulator pressure is at or more than the maximum RCS pressure for the existing temperature allowed by the P/T limit curves.

If isolation is needed and cannot be accomplished in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Required Action C. 1 and Required Action C.2 provide two options, either of which must be performed in the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . By increasing the RCS temperature to > the OPPS enable temperature specified in the PTLR, an accumulator pressure of 800 psig cannot exceed the LTOP analysis limits if the ECCS accumulators are fully injected. Depressurizing the ECCS accumulators below the LTOP limit from the PTLR also gives this protection.

The Completion Times are based on operating experience that these activities can be accomplished in these time periods and on engineering evaluations indicating that an event requiring LTOP is not likely in the allowed times.

D.1 In MODE 4 when any RCS cold leg temperature is --the OPPS enable temperature specified in the PTLR, with one required PORV Prairie Island Units 1 and 2 B 3.4.12-7 1/2/02

LTOP - RCSCLT > SI Pump Disable Temperature B 3.4.12 BASES ACTIONS D. 1 (continued) inoperable, the PORV must be restored to OPERABLE status within a Completion Time of 7 days. Two PORVs are required to provide low temperature overpressure mitigation while withstanding a single failure of an active component.

The Completion Time considers the facts that only one of the PORVs is required to mitigate an overpressure transient and that the likelihood of an active failure of the remaining valve path during this time period is very low.

E.1 MODE 5 must be entered, the RCS must be depressurized and a vent must be established within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when:

a. Both PORVs are inoperable; or

b. A Required Action and associated Completion Time of Condition A, C, or D is not met; or

c. The OPPS is inoperable.

The vent must be sized >_3 square inches to ensure that the flow capacity is greater than that required for the worst case mass input transient reasonable during the applicable MODES. The vent opening is based on the cross sectional flow area of a PORV. A PORV maintained in the open position satisfies the vent requirement. This action is needed to protect the RCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel.

Prairie Island Units I and 2 B 3.4.12-8 1/2/02

LTOP - RCSCLT > SI Pump Disable Temperature B 3.4.12 BASES ACTIONS E.1 (continued)

The Completion Time considers the time required to place the plant in this Condition and the relatively low probability of an overpressure event during this time period due to increased operator awareness of administrative control requirements.

SURVEILLANCE SR 3.4.12.1 and SR 3.4.12.2 REQUREMENTS To minimize the potential for a low temperature overpressure event by limiting the mass input capability, one SI pump is verified incapable of injecting into the RCS and the ECCS accumulator discharge isolation valves are verified closed and deenergized.

The SI pump is rendered incapable of injecting into the RCS by employing at least two independent means to prevent a pump start such that a single failure or single action will not result in an injection into the RCS. 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.

The ECCS accumulator motor operated isolation valves can be verified closed and deenergized by use of control board indication.

SR 3.4.12.2 is modified by a Note specifying that this verification is only required when the ECCS accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR. If ECCS accumulator pressure is less than this limit, no verification is required since the ECCS accumulator cannot pressurize the RCS to or above the OPPS setpoint.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering other indications and alarms available to the operator in the control room, to verify the required status of the equipment.

Prairie Island Units 1 and 2 B 3.4.12-9 1/2/02

LTOP - RCSCLT > SI Pump Disable Temperature B 3.4.12 BASES SURVEILLANCE SR 3.4.12.3 REQUIREMENTS (continued) The PORV block valve must be verified open every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to provide the flow path for each required PORV to perform its function when actuated. The valve may be remotely verified open in the main control room.

The block valve is a remotely controlled, motor operated valve. The power to the valve operator is not required to be removed, and the manual operator is not required to be locked in the inactive position.

Thus, the block valve can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure situation.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is considered adequate in view of other administrative controls available to the operator in the control room, such as valve position indication, that verify that the PORV block valve remains open.

SR 3.4.12.4 Performance of a COT is required every 31 days on OPPS to verify and, as necessary, adjust the PORV lift setpoints. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay.

This clarifies what is an acceptable CHANNEL OPERATIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The COT will verify the setpoints are within the PTLR allowed maximum limits in the PTLR.

PORV actuation during this testing could depressurize the RCS and is not required.

Prairie Island Units 1 and 2 B 3.4.12-10 1/2/02

LTOP - RCSCLT > SI Pump Disable Temperature B 3.4.12 BASES SURVEILLANCE SR 3.4.12.4 (continued)

REQUIREMENTS A Note has been added indicating that this SR is required to be performed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after decreasing RCS cold leg temperature to

_*the OPPS enable temperature specified in the PTLR. The COT may not have been performed before entry into the LTOP MODES.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months initial time considers the unlikehood of a low temperature overpressure event during this time.

SR 3.4.12.5 Performance of a CHANNEL CALIBRATION on OPPS is required every 24 months to adjust the whole channel so that it responds and the valve opens within the required range and accuracy to known input.

REFERENCES 1. 10 CFR 50, Appendix G.

2. USAR, Section 4.4.

3. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix G, with ASME Code Case N-514.

Prairie Island Units 1 and 2 B 3.4.12-11 1/2/02

LTOP - RCSCLT *_ SI Pump Disable Temperature B 3.4.13 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.13 Low Temperature Overpressure Protection (LTOP) Reactor Coolant System Cold Leg Temperature (RCSCLT) _*Safety Injection (SI) Pump Disable Temperature BASES BACKGROUND The LTOP function limits RCS pressure at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is not compromised by violating the pressure and temperature (P/T) limits of 10 CFR 50, Appendix G (Ref. 1). The reactor vessel is the limiting RCPB component for demonstrating such protection. The Over Pressure Protection System (OPPS) provides the actuation setpoints for the pressurizer power operated relief valves (PORVs) for the LTOP function (Ref. 2). The PTLR provides the maximum allowable OPPS actuation setpoints and the maximum RCS pressure for the existing RCS cold leg temperature during cooldown, shutdown, and heatup to meet the Reference 1 requirements during the LTOP MODES. The LTOP MODES are the MODES as defined in the Applicability statement of LCO 3.4.12 and LCO 3.4.13.

The pressurizer safety valves and PORVs at their normal setpoints do not provide overpressure protection for certain low temperature operational transients. Inadvertent pressurization of the RCS at temperatures below the OPPS enable temperature specified in the PTLR could result in exceeding the ASME Appendix G (Ref. 3) brittle fracture P/T limits. Exceeding the RCS P/T limits by a significant amount could cause brittle cracking of the reactor vessel.

LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," requires administrative control of RCS pressure and temperature during heatup and cooldown to prevent exceeding the PTLR limits.

This LCO provides RCS overpressure protection by restricting coolant input capability and ensuring adequate pressure relief capacity. In MODE 4, at or below the safety injection (SI) pump disable temperature, limiting coolant input capability requires both SI pumps incapable of injection into the RCS and isolating the emergency core cooling system (ECCS) accumulators. The pressure Prairie Island Units 1 and 2 B 3.4.13-1 1/2/02

LTOP - RCSCLT *5 SI Pump Disable Temperature B 3.4.13 BASES BACKGROUND relief capacity requires either two redundant RCS relief valves or a (continued) depressurized RCS and an RCS vent of sufficient size. One PORV or the open RCS vent is the overpressure protection device that acts to terminate an increasing pressure event.

Limiting coolant input capability reduces the ability to provide core coolant addition. The LCO does not require the makeup control system deactivated or the safety injection SI actuation circuits blocked. Due to the lower pressures in the LTOP MODES and the expected core decay heat levels, the charging system can provide adequate flow. If conditions require the use of an SI pump for makeup in the event of loss of inventory, the pump can be made available through manual actions.

The LTOP pressure relief consists of two PORVs with reduced lift settings or a depressurized RCS and an RCS vent of sufficient size.

Two PORVs are required for redundancy. One PORV has adequate relieving capability to prevent overpressurization for the required coolant input capability.

OPPS and PORV Requirements As designed for the LTOP function, each PORV is signaled to open by OPPS if the RCS pressure approaches the lift setpoint provided when OPPS is enabled. The OPPS monitors both RCS temperature and RCS pressure and indicates when a condition not acceptable in the PTLR limits is approached. The wide range RCS temperature setpoints indicate conditions requiring enabling OPPS. The PTLR presents the OPPS setpoints for LTOP.

RCS Vent Requirements Once the RCS is depressurized, a vent exposed to the containment atmosphere will maintain the RCS at containment ambient pressure Prairie Island Units 1 and 2 B 3.4.13-2 1/2/02

LTOP - RCSCLT < SI Pump Disable Temperature B 3.4.13 BASES BACKGROUND RCS Vent Requirements (continued) in an RCS overpressure transient, if the relieving requirements of the transient do not exceed the capabilities of the vent. Thus, the vent path must be capable of relieving. the flow resulting from the limiting LTOP mass or heat input transient, and maintaining pressure below the P/T limits. The required vent capacity may be provided by one or more vent paths.

APPLICABLE Safety analyses (Ref. 2) demonstrate that the reactor vessel is SAFETY adequately protected against exceeding the Reference 1 P/T limits.

ANALYSES In MODES 1, 2, and 3, and in MODE 4 with RCS cold leg temperature exceeding the OPPS enable temperature specified in the PTLR, the pressurizer safety valves will prevent RCS pressure from exceeding the Reference 1 limits. At about the OPPS enable temperature specified in the PTLR and below, overpressure prevention falls to two OPERABLE PORVs or to a depressurized RCS and a sufficiently sized RCS vent. Each of these means has a limited overpressure relief capability. LCO 3.4.12, "LTOP > SI Pump Disable Temperature," provides the requirements for overpressure prevention at temperatures above the SIP disable temperature.

The actual temperature at which the pressure in the P/T limit curve falls below the pressurizer safety valve setpoint increases as the reactor vessel material toughness decreases due to neutron embrittlement. Each time the PTLR curves are revised, LTOP must be re-evaluated to ensure its functional requirements can still be met using the PORV method or the depressurized and vented RCS condition.

The PTLR contains the acceptance limits that define the LTOP requirements. Any change to the RCS must be evaluated against the Reference 2 analyses to determine the impact of the change on the LTOP acceptance limits.

Prairie Island Units 1 and 2 B 3.4.13-3 1/2/02

LTOP - RCSCLT _<SI Pump Disable Temperature B 3.4.13 BASES APPLICABLE Transients that are capable of overpressurizing the RCS are SAFETY categorized as either mass or heat input transients. The bounding ANALYSES mass input transient is inadvertent safety injection with injection (continued) from one SI pump and three charging pumps, and letdown isolated.

The bounding heat input transient is reactor coolant pump (RCP) startup with temperature asymmetry within the RCS or between the RCS and steam generators.

The following limitations are required during the Applicability of this specification to ensure that mass and heat input transients in excess of analysis assumptions do not occur:

a. Rendering both SI pumps incapable of injection;

b. Deactivating the ECCS accumulator discharge isolation valves in their closed positions; and

c. Disallowing start of an RCP if secondary temperature is more than 50 'F above primary temperature in any one loop.

LCO 3.4.6, "RCS Loops - MODE 4," provides this protection.

The Reference 2 analyses demonstrate that either one PORV or the depressurized RCS and RCS vent can maintain RCS pressure below limits when all charging pumps are actuated. Neither one PORV nor the RCS vent can handle the pressure transient resulting from inadvertent SI pump or ECCS accumulator injection when the RCS is below the SI Pump disable temperature. Thus, the LCO requires both SI pumps to be disabled below the temperature specified in the PTLR.

The LCO also requires ECCS accumulator isolation when ECCS accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR. The isolated ECCS accumulators must have their discharge valves closed and the valve power supply breakers fixed in their open positions.

Prairie Island Units 1 and 2 B 3.4.13-4 1/2/02

LTOP - RCSCLT - SI Pump Disable Temperature B 3.4.13 BASES APPLICABLE Fracture mechanics analyses established the temperature of LTOP SAFETY Applicability at the OPPS enable temperature specified in the PTLR.

ANALYSES The fracture mechanics analyses show that the vessel is protected (continued) when the PORVs are set to open at or below the limit shown in the PTLR. The OPPS setpoints are derived by analyses that model the performance of the system, assuming the limiting LTOP transient of all charging pumps injecting into the RCS. These analyses consider pressure overshoot and undershoot beyond the PORV opening and closing, resulting from signal processing and valve stroke times.

The OPPS setpoints at or below the derived limit ensures the Reference 1 P/T limits will be met.

The OPPS setpoints in the PTLR will be updated when the revised P/T limits conflict with the LTOP analysis limits. The P/T limits are periodically modified as the reactor vessel material toughness decreases due to neutron embrittlement caused by neutron irradiation. Revised limits are determined using neutron fluence projections and the results of examinations of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," discuss these examinations.

With the RCS depressurized, analyses show a vent size equivalent to the cross sectional flow area of a PORV is capable of mitigating the allowed LTOP overpressure transient. The capacity of a vent this size is greater than the flow of the limiting transient for the LTOP configuration, both SI pumps disabled and all charging pumps OPERABLE when the RCS is below the SI Pump disable temperature, maintaining RCS pressure less than the maximum pressure on the P/T limit curve.

The RCS vent is passive and is not subject to active failure.

The LTOP function satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

Prairie Island Units 1 and 2 B 3.4.13-5 1/2/02

LTOP - RCSCLT *< SI Pump Disable Temperature B 3.4.13 BASES (continued)

LCO This LCO requires that LTOP be provided, by limiting coolant input capability and by OPERABLE pressure relief capability. Violation of this LCO could lead to the loss of low temperature overpressure mitigation and violation of the Reference 1 limits as a result of an operational transient.

To limit the coolant input capability, the LCO requires both SI pumps be incapable of injecting into the RCS, and all ECCS accumulator discharge isolation valves be closed and deenergized (when ECCS accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR).

The LCO is modified by three Notes. Note 1 allows operation of both SI pumps for _*1 hour for conducting SI system testing providing there is a steam or gas bubble in the pressurizer and at least one isolation valve between the SI pump and the RCS is shut.

The purpose of this note is to permit the conduct of the integrated SI test and other SI system tests and operations that may be performed in MODES 4, 5 or 6. 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 to the RCS.

This isolation is accomplished by either a closed manual valve or motor operated valve with the power removed. This combination of conditions under strict administrative control assure that overpressurization cannot occur.

Note 2 allows operation of an SI pump during reduced inventory conditions as required to maintain adequate core cooling and RCS inventory. The purpose of this note is to allow use of an SI pump in the event of a loss of other injection capability (e.g., loss of Residual Heat Removal System cooling while in reduced inventory conditions). The operation of an SI pump under such conditions would be controlled by an approved emergency operating procedure.

Prairie Island Units 1 and 2 B 3.4.13-6 1/2/02

LTOP - RCSCLT ! SI Pump Disable Temperature B 3.4.13 BASES LCO Note 3 states that ECCS accumulator isolation is only required when (continued) ECCS accumulator pressure is more than or at the maximum RCS pressure for the existing RCS cold leg temperature allowed by the P/T limit curves provided in the PTLR (less allowance for instrument uncertainty). This Note permits the ECCS accumulator discharge isolation valve Surveillance to be performed only under these pressure and temperature conditions.

The elements of the LCO that provide low temperature overpressure mitigation through pressure relief are:

a. An OPERABLE OPPS with two PORVs; or A PORV is OPERABLE for LTOP when its block.valve is open, its low pressure lift setpoint has been selected (OPPS enabled), and the backup air supply is charged.

b. A depressurized RCS and an RCS vent.

An RCS vent is OPERABLE when open with an area of

Ž 3.0 square inches. 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 vent requirement.

Each of these methods of overpressure prevention is capable of mitigating the limiting LTOP transient.

APPLICABILITY This LCO is applicable in MODE 4 when any RCS cold leg temperature is

the SI Pump disable temperature specified in the PTLR, in MODE 5, and in MODE 6 when the reactor vessel head is on and the SG primary system manways and pressurizer manway are closed and secured. The pressurizer safety valves provide overpressure protection that meets the Reference 1 P/T limits above the OPPS enable temperature specified in the PTLR. When the reactor vessel head is off, overpressurization cannot occur.

Prairie Island Units 1 and 2 B 3.4.13-7 1/2/02

LTOP - RCSCLT *< SI Pump Disable Temperature B 3.4.13 BASES APPLICABILITY LCO 3.4.3 provides the operational P/T limits for all MODES.

(continued) LCO 3.4.10, "Pressurizer Safety Valves," requires the OPERABILITY of the pressurizer safety valves that provide overpressure protection during MODES 1, 2, and 3, and MODE 4 above the OPPS enable temperature specified in the PTLR. LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) :! Safety Injection Pump (SI) Pump Disable Temperature," provides the requirements for MODE 4 below the OPPS enable temperature and above the SI Pump disable temperature.

Low temperature overpressure prevention is most critical during shutdown when the RCS is water solid, and a mass or heat input transient can cause a very rapid increase in RCS pressure when little or no time allows operator action to mitigate the event.

ACTIONS A.1 With one or more SI pumps capable of injecting into the RCS, RCS overpressurization is possible.

To immediately initiate action to restore restricted coolant input capability to the RCS reflects the urgency of removing the RCS from this condition.

B.1, C.1, and C.2 An unisolated ECCS accumulator requires isolation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

This is only required when the ECCS accumulator pressure is at or more than the maximum RCS pressure for the existing temperature allowed by the P/T limit curves.

If isolation is needed and cannot be accomplished in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Required Action C. 1 and Required Action C.2 provide two options, either of which must be performed in the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . By Prairie Island Units 1 and 2 B 3.4.13-8 1/2/02

LTOP - RCSCLT *< SI Pump Disable Temperature B 3.4.13 BASES ACTIONS B.1, C.1, and C.2 (continued) increasing the RCS temperature to > the OPPS enable temperature specified in the PTLR, an ECCS accumulator pressure of 800 psig cannot exceed the LTOP analysis limits if the ECCS accumulators are fully injected. Depressurizing the ECCS accumulators below the LTOP limit from the PTLR also gives this protection.

The Completion Times are based on operating experience that these activities can be accomplished in these time periods and on engineering evaluations indicating that an event requiring LTOP is not likely in the allowed times.

D.l The consequences of operational events that will overpressurize the RCS are more severe at lower temperature. Thus, with one PORV inoperable in MODE 4 when any RCS cold leg temperature is _< the SI Pump disable temperature specified in the PTLR, MODE 5 or in MODE 6 with the head on, the Completion Time to restore two valves to OPERABLE status is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . A Note clarifies that Condition D is only applicable when the OPPS and PORVs are being used to satisfy the pressure relief requirements of LCO 3.4.13.a.

The Completion Time represents a reasonable time to investigate and repair several types of relief valve failures without exposure to a lengthy period with only one OPERABLE PORV to protect against overpressure events.

E.1 The RCS must be depressurized and a vent must be established within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when:

Prairie Island Units 1 and 2 B 3.4.13-9 1/2/02

LTOP - RCSCLT --SI Pump Disable Temperature B 3.4.13 BASES ACTIONS E. 1 (continued)

a. Both required PORVs are inoperable; or

b. A Required Action and associated Completion Time of Condition A, C, or D is not met; or

c. The OPPS is inoperable.

The vent must be sized Ž_ 3 square inches to ensure that the flow capacity is greater than that required for the worst case mass input transient reasonable during the applicable MODES. The vent opening is based on the cross sectional flow area of a PORV. A PORV maintained in the open position satisfies the vent requirement. This action is needed to protect the RCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel.

The Completion Time considers the time required to place the plant in this Condition and the relatively low probability of an overpressure event during this time period due to increased operator awareness of administrative control requirements.

SURVEILLANCE SR 3.4.13.1 and SR 3.4.13.2 REQUIREMENTS To minimize the potential for a low temperature overpressure event by limiting the mass input capability, both SI pumps are verified incapable of injecting into the RCS and the ECCS accumulator discharge isolation valves are verified closed and deenergized.

The SI pumps are rendered incapable of injecting into the RCS by employing at least two independent means to prevent a pump start such that a single failure or single action will not result in an injection into the RCS. This may be accomplished through the Prairie Island Units 1 and 2 B 3.4.13-10 1/2/02

LTOP - RCSCLT ___ SI Pump Disable Temperature B 3.4.13 BASES SURVEILLANCE SR 3.4.13.1 and SR 3.4.13.2 (continued)

REQUIREMENTS pump control switch being placed in pullout with a blocking device installed over the control switch that would prevent an unplanned pump start.

The ECCS accumulator motor operated isolation valves can be verified closed and deenergized by use of control board indication.

SR 3.4.13.2 is modified by a Note specifying that this verification is only required when the ECCS accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR. If ECCS accumulator pressure is less than this limit, no verification is required since the ECCS accumulator cannot pressurize the RCS to or above the OPPS setpoint.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering other indications and alarms available to the operator in the control room, to verify the required status of the equipment.

SR 3.4.13.3 The RCS vent of __3 square inches is proven OPERABLE by verifying its open condition either:

a. Once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for a valve that is not locked, sealed, or secured in the open position.

b. Once every 31 days for other vent path(s) (e.g., a vent valve that is locked, sealed, or secured in position). A removed pressurizer safety valve or open manway also fits this category.

The passive vent path arrangement must only be open when required to be OPERABLE. This Surveillance is required if the vent is being used to satisfy the pressure relief requirements of LCO 3.4.13b.

Prairie Island Units 1 and 2 B 3.4.13-11 1/2/02

LTOP - RCSCLT _*SI Pump Disable Temperature B 3.4.13 BASES SURVEILLANCE SR 3.4.13.4 REQUIREMENBS (continued) The PORV block valve must be verified open every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to provide the flow path for each required PORV to perform its function when actuated. The valve may be remotely verified open in the main control room. This Surveillance is performed if the PORV satisfies the LCO.

The block valve is a remotely controlled, motor operated valve. The power to the valve operator is not required to be removed, and the manual operator is not required to be locked in the inactive position.

Thus, the block valve can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure situation.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is considered adequate in view of other administrative controls available to the operator in the control room, such as valve position indication, that verify that the PORV block valve remains open.

SR 3.4.13.5 Performance of a COT is required every 31 days on OPPS to verify and, as necessary, adjust the PORV lift setpoints. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay.

This clarifies what is an acceptable CHANNEL OPERATIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The COT will verify the setpoints are within the PTLR allowed maximum limits in the PTLR.

PORV actuation during this testing could depressurize the RCS and is not required.

Prairie Island Units 1 and 2 B 3.4.13-12 1/2/02

LTOP - RCSCLT _<SI Pump Disable Temperature B 3.4.13 BASES SURVEILLANCE SR 3.4.13.5 (continued)

REQUIREMENTS Note 1 has been added indicating that this SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after decreasing RCS cold leg temperature to _<the OPPS enable temperature specified in the PTLR. The COT may not have been performed before entry into the LTOP MODES.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months initial time considers the unlikehood of a low temperature overpressure event during this time.

Note 2 has been added to specify that this SR is only required to be performed when OPPS and PORVs are providing the LTOP function per LCO 3.4.13.a.

SR 3.4.13.6 Performance of a CHANNEL CALIBRATION on OPPS is required every 24 months to adjust the whole channel so that it responds and the valve opens within the required range and accuracy to known input.

REFERENCES 1. 10 CFR 50, Appendix G.

2. USAR, Section 4.4.

3. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix G, with ASME Code Case N-514.

Prairie Island Units 1 and 2 B 3.4.13-13 1/2/02

RCS Operational LEAKAGE B 3.4.14 BASES SURVEILLANCE SR 3.4.14.1 (continued)

REQUIREMENTS The RCS water inventory balance must be met with the reactor at steady state operating condition (stable temperature, power level, equilibrium xenon, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows). Therefore, a Note is added allowing that this SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Steady state operation is required to perform a proper inventory balance since calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by monitoring containment atmosphere radioactivity. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in LCO 3.4.16, "RCS Leakage Detection Instrumentation."

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents.

SR 3.4.14.2 This SR provides the means necessary to determine SG OPERABILITY in an operational MODE. The requirement to demonstrate SG tube integrity in accordance with the Steam Prairie Island Units 1 and 2 B 3.4.14-7 1/2/02

RCS Specific Activity B 3.4.17 BASES APPLICABILITY For operation in MODE 3 with RCS average temperature < 500°F, (continued) and in MODES 4 and 5, the release of radioactivity in the event of a SGTR is unlikely since the saturation pressure of the reactor coolant is below the lift pressure settings of the main steam safety valves.

ACTIONS A.1 and A.2 With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at intervals of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months must be taken to demonstrate that the limits of Figure 3.4.17-1 are not exceeded. The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is required to obtain and analyze a sample. Sampling is done to continue to provide a trend.

The DOSE EQUIVALENT 1-131 must be restored to within limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is required, if the limit violation resulted from normal iodine spiking.

Permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 u.Ci/gm DOSE EQUIVALENT 1-13 1, but within the allowable limit shown on Figure 3.4.17-1, accommodates the possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1.0 pCi/gm DOSE EQUIVALENT 1-131 but within the limits shown on Figure 3.4.17-1 should be minimized since the activity levels allowed by the figure increase the dose at the site boundary following a postulated steam generator tube rupture.

A Note to the ACTIONS excludes the MODE change restriction of LCO 3.0.4. This exception allows entry into the applicable MODE(S) while relying on the ACTIONS even though the ACTIONS may eventually require plant shutdown. This exception is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is Prairie Island Units 1 and 2 B 3.4.17-3 1/2/02

RCS Specific Activity B 3.4.17 BASES ACTIONS A. 1 and A.2 (continued) limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operation.

B.1 With the gross specific activity in excess of the allowed limit, the reactor must be placed in a MODE in which the requirement does not apply. The change within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to MODE 3 and RCS average temperature < 500'F lowers the saturation pressure of the reactor coolant below the setpoints of the main steam safety valves and prevents venting the SG to the environment in a SGTR event. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 below 500'F from full power conditions in an orderly manner without challenging plant systems.

C.1 If a Required Action and the associated Completion Time of Condition A is not met or if the DOSE EQUIVALENT 1-131 is in the unacceptable region of Figure 3.4.17-1, the reactor must be brought to MODE 3 with RCS average temperature < 500'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 below 500 'F from full power conditions in an orderly manner without challenging plant systems.

Prairie Island Units 1 and 2 B 3.4.17-4 1/2/02

RCS Specific Activity B 3.4.17 BASES (continued)

SURVEILLANCE SR 3.4.17.1 REQUIREMENTS SR 3.4.17.1 requires performing a gamma isotopic analysis as a measure of the gross specific activity of the reactor coolant at least once every 7 days. While basically a quantitative measure of radionuclides with half lives longer than 15 minutes, excluding iodines, this measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken.

This Surveillance provides an indication of any increase in gross specific activity.

Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The Surveillance is applicable in MODES 1 and 2, and in MODE 3 with Tavg at least 500SF. The 7 day Frequency considers the unlikelihood of a gross fuel failure during the time.

SR 3.4.17.2 This Surveillance is performed in MODE 1 only to ensure iodine remains within limit during normal operation and following fast power changes when fuel failure is more apt to occur. The 14 day Frequency is adequate to trend changes in the iodine activity level, considering gross activity is monitored every 7 days. The Frequency, between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change >_15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, is established because the iodine levels peak during this time following fuel failure; samples at other times would provide inaccurate results.

Prairie Island Units 1 and 2 B 3.4.17-5 1/2/02

PACKAGE 3.4 REACTOR COOLANT SYSTEM (RCS)

PART C MARKUP OF PRAIRIE ISLAND CURRENT TECHNICAL SPECIFICATIONS List of Pages Part Current Technical Part Current Technical Specifications Page C Page Specifications Page C Page 1 TS.3-10-8 18 TS.3.1-9 TS.3.1 -1 19 TS.3.1 -10 2

20 Figure TS.3.1-3 3 TS.3.1-1 Overflow 21 TS.3.3-3 4 TS.3.1-2 22 Table TS.4.1-1C(Page 3 of 4) 5 TS.3.1-2 Overflow 23 Table TS.4.1-1C(Page 4 of 4) 6 TS.3.1-2 Overflow (Cont.)

24 Table TS.4.1-2A(Page 1 of 2) 7 TS.3.1-3 25 Table TS.4.1-2A(Page 2 of 2) 8 TS.3.1-3 Overflow 26 Table TS.4.1-2B(Page 1 of 2) 9 TS.3.1-4 27 Table TS.4.1-2B(Page 2 of 2) 10 TS.3.1-4 Overflow 28 TS.4.3-1 11 TS.3.1-4 Overflow (Cont.)

29 TS.4.6-3 12 TS.3.1-5 30 TS.4.18-1 13 TS.3.1-5 Overflow 14 TS.3.1-6 15 TS.3.1-7 16 TS.3.1-8 17 TS.3.1-8 Overflow PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNITS 1 AND 2 Improved Technical Specifications Conversion Submittal

IA3-4-00 1 TS.3 058 REV 92 3/13/90 R-2 Addressed IElsewhere J--RCS Pressure? Temperature and Flow - DNB Limits Parametors

.4.1 IThe following DNB related parameters limits shall be maintained R-9 I--

during POWER OPERATION: m-01

a. Reactor Coolant System Tavg *<-limit specified in the COLR5644SF

b. Pressurizer Pressure >limit specified in the COLR 2-2 ps4ia*-

c. Reactor Coolant Flow Žthe value specified in the CORE OPERATING LIMITS REPORT LC03.4.l1 With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to MODE 2 less than 5% of PATED 408 THERMAL POWER within the next ýý4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . I 05 SR 3 .4 *. 1 . tCompliance with a. and b. is demonstrated by verifying that each of he parameters is within its limits at least once each 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SR3.4 :.I.Zj Compliance with c. is d.m..*tr.at. d by verify4-nq ROS total flow rate SR3.4 z1.3L 1 that the parameter is within f limit secified in the COLE after I each refueling cycle.

L R-9 03.4.1 LC Jn3.-0-° I

Limit not applicable during either a THERMAL POWER ramp inerease in excess of (5%) RATED THERMAL POWER per minute or a THERMAL POWER step inerease in excess of (10%) RATED THERMAL POWER PI Current TS 1 of 30 Markup for PI ITS Part C

T-S.3.ll:-

REV 157 7,114,10 3.41 REACTOR COOLANT SYSTEM A3. 4 -00 I..

S....i.........

Applies to the operating status . f the r... .. .. nt s. tem when irradiated fu*l is in the.ontaminnt .

A3.4-03 ObjectI ve To specify these limiting conditions for ep.r.tion of the r.a.t.r eeolant system ;hieh maust be met to assure safe reaetor epearotioen.

Specification A. Operational Components

1. Reactor Coolant Loops and Coolant Circulation

a. Reactor Critical LCO3.4 (1) A reactor shall not be in MODE 2-made or maintainad eritiea! unless both reactor coolant loops (with their I

associated steam generator and reactor coolant pump) are in operation, .eeep.. 1-with thermal power <P-7 (if power >P-7, open Reactor S4. I8 Trip Breaker) during low power PHYSICS TESTS, New SR, Verify power <P-7 and perform COT on P-7, R-9 SR3 4.18 power(low setpoint) and intermediate range neutron I 7 flux channels.

or 2) as specified in 3.l.A.l.a. (2) below.

LCO3.4.4 (2) With less than the above required reactor coolant loops in operation, be in at least MODE 3110T SITUTD... within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

New SR, Verify each RCS loop in operation

b. Reactor Coolant System MODE :3Av_ __rage

'oempralýt.re A eve - - . .--. 4 -2 LC03.4.5 1 (1) Reactor coolant system average temperature shall not be in MODE 3eieeed 9--02-F-unless both reactor coolant loops (with their associated steam generator and reactor coolant pump) are OPERABLE and both RCS loops shall be in operation when the Rod Control l System is capable of rod withdrawal; or PI Current TS 2 of 30 Markup for PI ITS Part C

Overflow with-at least one reactor coolant loop in 13 operation when the Rod Control System is n-ot capable of rod withdrawal, (except as specified in 3.l.A.l.b(2) and 3.l.A.l.b. (3) below).

(2) A reactor coolant loop may be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months previded STARTUP OPERATION is diseentinued until OPEPABILITY is restored. If OPERABILITY is not restored Co nd A:.5 LCO3.4 within the time specified, be in MODE 4r*duce reactor A48 Cond B3 0 coolant system average temperature below 3SO F within the next 12 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . 1L 6 (3) with both reactor coolant numos inoperable or not in operat :ion immediately: IL3"4-118 Cond D (a) De energize all Place the control rod drive system in a condition incapable of rod withdrawal meehanisma, (b) Suspend all operations involving a reduction of RCSý boron concentration, ,

rLCO3.4.5 I (c) Establish and maintain the core outlet temperature at LCO Note least 10 'F below saturation temperature, and LCn (c) Initiate action to restore one reactor coolant pump to Cond D OPERABLE status and operation.*

if at least ene reaetor eeelant pump is net roesterod to TTl GPERAEI1.1TY and operation w.nthn :7 neurs, izeuu t= '-- ~ e b le-ý -990 next 12 h.urs. While applics bin. this snecification superccdes 3.1 .A.I.b(2).

If the RCP shutdown or inoperability was due to preplanned work activities such as testing, switching, or maintenance, immediate not required, but if atandleast one reactor NCOT INOTE . coolant pump action restoration is not isrestored to operability operation within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , R-9 rociu.. reactor coolant system average eompcracuro s I i LC03.4.5 If one RCS loop is not in operation with Rod Control Cond C System capable of rod withdrawal, within one hour I

restore the loop to operation, or render the Rod Control System inoperable.

New SRs, Verify RCS loops operating, verify SG capable of321 ISR3.4.5.12 SR3.4.5. 2 SR3.4.5.3 removing decay heat, verify breaker alignment for non R-9 PI Current TS 3 of 30 Markup for PI ITS Part C

REV 1R5C/1.'98 r A3vTm 3.1.A1.c. Reactor Coolant System Mode 4, loops fille Pverage Temperature Bolew

ý_ A -4 A P # T.PVP -1 e v e tHe ftea A.4-:; r vesse e-ý-

1O2 IT n A .(1)Whenever the reactor coolant system average tefmerature is Oclow 0

3SO F, excopt during REFUELING, at leastMode 4, two loops consisting of any combination of methods f c removinrg d..ay hea shall be OPERABLE with one in operation* (ýep sseiidi 1-0 beloew).

(2) Acmeptable mcthed*f.r rem.ving de.ay heat are at least one reV . I-24 coolant system loop pump and it-s (3) asseociated steam generator; or and residual heat removal loop indec*ing a pump ang +ts asseeoclaace uetnger hat IA3.4-I03 I 71.tA..e.. Reactor Coolant System Mode 5, loops filledAverage Temperature Belo.w A O.4-1o I JLC03.4.7 1 Whq~npivPer #Ihe rea#ctr- ceeelant svstemn averaco ternoerature is below 25~,except during REFUE1.+/-LJU, at- lasat two metanoea ror emoi d-e.eay hea,-Mode 5 with loops filled, one RHP loop shall be OPERABLE wi.th *and e in operation* (except as sp..ified in 3.-.A.*.C. (2) b-'w). '-"-n-table m2fethed for remevina decay heat areand at least one roactor eoolant pump and its a..o.iated steam generator capable of decay heat removal; or a residual heat removal loop A1qPu44Hqn.f a eumBadit seeeedha exe ange Mý ,. iR3.4-24 IL3"4-126 IIA3"4-1°4 IIA3"4-1°5 1 (2) With only one required loop inoperable OPERPABLE method of removing deeay heat* immediately initiate pr~emp- action to restore a second ConLC03. 4. 6 loop to twe OPERABLE metheds ef removing de.ay CondLC3A.

[~TI (2) With enly one required RER loop inoperable and one RHR loop OPERABLE method of r-emoing decay -h-t, immediately initiate premp-.

action to restore a second RER loop to twe OPERABLE or initiate atin LC03 .4 .7 to restore required SG capable to ef-removeing decay heat. LR3.4-24I IA3.4-1o I (2) With enly one or more SGs not capable OPERABLE mcthod of removing decay heat and one RHR loop OPERABLE, immediately initiate pr empt Cond B action to restore a second RER loop to twe OPERABLE or initiate action or restore required SGs capable to remove methods of removing- decay heat. ....... I Only required If the remaining RHR loop is operable method is an Cond A RiiR leep, be in MODE 5COLD SHUTDO.N within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . O Not e R-9 PI Current T* 4 of 30 Markup for PI ITS Part C

T-S .3 .1-2 REV 135 S/'4/98 IA3.4-08 j (Eov er f 1 e )

w3)

With no OPERBPLE mothods of romoving deeayht Both loopsI -°1 CondI inoperable or required loop not in operation, suspend ConLC03. 4. 6 operations that would cause introduction into RCS coolant with involving a rodu tien in boron concentration less than required to meet SDM of LCO 3.1.1 ef-the r-RRcý1=)r R~eoafiE sys.tem and immediately LC03.4.7l initiate prempt action to restore one loop OPERABLE and in operation Cond C J method of removing deoay heat.

(3) With no OPERABLE methods of removing decay heat required RHR loops OPERABLE or required RHR loop not in operation, suspend a1*1.

operations that would cause introduction into RCS coolant with involvinga rdution in boron concentration less than required to meet SDM of LCO 3.1.1 of the ra..t..r... l.ant s ..tem and immediately initiate prempt action to restore one loop OPERABLE and in operation method oef romoving decay heat. J 22

,R-9 (4) A reactor coolant pump may be started at RCS temperature less LC03 .4. G than the Over Pressure Protection System Enable Temperature specified in the PTLR, only if either of the following NOTE 2 conditions is met:

LC03.4. 7 NOTE 3 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.

Both RHR loops may be removed from operation for planned 28

4. heatup to MODE 4 when at least one RCS loop is in NOTELC3 operation.

ISR3.4.6.1 New SRs, Verify one RHR or RCS loop in operation, verify SR3.4.6.2 SR3.4.6.3 SG capable of removing decay heat, verify pump breaker alignment ISR3.4.7.1 New SRs, Verify one RHR loop in operation, verify SG capable of removing decay heat, verify pump breaker alignment R-9

i. . .

PI Current TS 5 of 30 Markup for PI ITS Part C

Ts. 3.i2 Gverflc;w (continuod-)

d-L.Reactor Coolant System, Mode 5, loops not filledLoveol Beo.w or at the Reactor Vessel Flanq (1) Both residual heat removal loops, each oonsisting of a pump an, Lasociated heat ...hang..., shall be OPERABLE with one in operation* (except as specified in 3.1.A.l.d.(2) below).

With one required or beth residual heat removal loop(s) inoperable, LC03.4. 8 Cond A pr-emptimmediately action shall be taken to restore the inoperable residual heat removal loop--s-+ to an OPERABLE status.

LC03.4.8 and, if no RHR loops operable, or required RHR Loop Cond B not in operation, immediately suspend all operations involving reduction in RCS boron concentration and initiate action to restore one RHR Loop to OPERABLE status and in operation.

During reduced inventory conditions, a safety injection R-9 1LC03.4. 8 pump may be run as required to maintain adequate core LCO3.4. 13 Notes 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 per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period LC03.4.6 1provided the reactor is subcritical, no operations are permitted 1-1-.4.7 that would cause dilution of the reactor coolant boron 1concentration and core outlet temperature is main tained at least 10'F below saturation temperature.

03.4.8 *0All pumps may be shutdown for up to one hour per a hour period TB 1 Iprovided the reactor is subcritical, 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 and no ROS draining operations are permitted.

{-9 SI_

LC03.4.7ne required RER loop may be inoperable for < 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for LCO3.4.8 surveillance provided that the other RHR loop is OPERABLE and in MOTE 2 ]operation.

.4.8.1 New SRs, Verify required RER loop in operation, verify breaker

.4.B.2 I alignment for RHR pump not operating ,R-9 P1 Current TS 6 of 30 Markup for P1 ITS Part C

REV 91 10/27,/89

'I Zý ;2_ PpnýRtPr rosAAnt# System Pr-ess'are Control

a. Pressurizer A.8 (1)When A reactor is in MODE 1, 2, and 3, the pressurizer shall be 1LC03.4.9 1 OPERABLE witsh-l Pt bmaeemanandeielnrshl reactor eeolant system average temper atudre ex eed 350OF unlees pressurizer level 90% there t i team bubble in the M 41 pre..uri.Zr and two heater groups A.and.B are operable 24 and capable of being powered from an

.emergency power supply.

(except as specified in 3.1.A.2.a.2 and 3.1.A.2.a.3 below).

AO8 (2)During M4odes 1, 2, and 3STARTUP TPERP :'+/-'UH4 or iU"*LE) u1'blc"+/-u2n, n BC oLC03.4.9 Group72 Ahours for or B one pressurizer heater crroup may be inoperable JA3.4-141 preooided STARTUP OPERATION is di - P'n4;:1 V 4 4114 P Iliq - ý I i, i kI is restered.

If OPERABILITY is not restored within the time specified, be in at least MODE 3HOT SHUTDOWN within the next 6 JA3.4-08 hours and MODE 4reduee reaetor coolant system average temperature below 350 0 F within ,2the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

(3) With the pressurizer otherwise inoperable, within one M 42 heou initiate the action necessary to place the unit in Cond A HOT. SHUTDO..WN, and be in at...east. MODE 34IT. SHUTDON.

wi-thin the next 6 houros, AND fully insert all rods AND render Control Rod system incapable of rod withdrawal in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , R-9 i -

and be in MODE qreauee rea.t. r e...Lant average temperature below 350 0 F within 12the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ISR3.4.9.1 ]New SR, Verify water level is < 90% in pressurizer.

i44 R-9 PI Current TS 7 of 30 Markup for PI ITS Part C

TS.3.1*3 OVerflow

b. Pressurizer Safety Valves

-i)Reactor Coolant System Modes 1, 2, 3, and 4 with jJall cold leg a-veragetemperatureS greater than OPPSý IA3.4-085 IA 8.4-48

... enable temperature specified in th~e PTLRe*-~ra to 25 0oF When A reactor is in MODE 1, 2, 3, and 4 with all shallno pb modo orý mo-i~ntor riio o IM3.4-45 shall4reactor coolant system cold leg e e JLC03.4.10 temperatures >e.eeed OPPS enable temperature specified in the PTLR, .G..

unless- two pressurizer safety valves shall beare OPERABLE, with IA3.4-46 lift settings of 2485 psig +/-i 4:%.

If these conditions cannot be satisfied, diooontinue 3 LC03.4.10 STARTUP OPERATION and within 15 minutes restore valve Cord A to OPERABLE statusinitiate the o.tion necessary t-ot place the unit in HOT SUTDOWN, and be in at least MODE 3rHOT SHUTDOWN within the next "LC03.4.10 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and redue-ýbe in MODE 4 with any reactor coolant system cold leg a;eraqe-temperature below IM3.4-45 the OPPS enable temperature specified in the PTI4R 73.4-109 I 3-58F within 24tho f 6-hours.

'llo;.ng

,R-9 NOTEt The lift settings are not required to be within the LCO limits during MODE 3 for the purpose LC03.4.10 of setting the pressurizer safety valves under NOTE ambient (hot) conditions, This exception is allowed for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months following entry into MODE 3 provided a preliminary cold setting was made prior to heatup.

47 At least .n. pressurizer safety valve shall. be PE. .LE, with a lift sctting of 2485 psig +/-1%, whenever the head is on the reactor vessel, o...pt during hydro-tatio tests.

With nO pressurizer sal fty rPEmLoit,valve promptly plo an OPEPABLE residual heat romovol loop-into operotion.

PI Current TS 8 of 30 Markup for PI ITS Part C

TS. 3. 4:

REV 135 5/4/98 3.1.A.2.c Pressurizer Power Operated Relief Valves oS average LC03 ..4.11 (i) modes 1, 2, and 3 (Reactor Coolant System

... temperature greater than or equal to 350OF (a) in MODES 1, 2, and -)tte a e _eE_ O8 S-3 F'unless two power operated relief valves (PORVs)

4. 1l

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

(b) During STARTUP OPEPRATION or POWER OPERATION, any o f 3.

the following conditions of inoperabi-ity may exist for ILCO .4.11 each unit. -If OPERABILITY is not restored within the time Cond D, G specified or the required action cannot be completed, be in A3 _0 at least MODE 3HGT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and bOlo 35reduc reactor belo *=OF-*

within 12the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . JA3"4-39

1. With one or both PORVs inoperable because of excessive IA3.449 "TL03 4 .11I1 seat leakage, within one hour either restore the PORV(o)

ond A to OPERABLE status or close the associated block valve(s) with power maintained to the block valve(s).

2. With one PORV inoperable due to causes other than 49 rester.A3"4-excessive seat leakage, within one hour either LC03.4.11 the P0RV to OPEBPtBLE status or close and remove power Cond B from the associated block valve. Restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3. With both PORVs inoperable due to causes other than excessive seat leakage, within one hour ither r. tr. A3.4-49 at least one P0RV to OPEPRBLE status or close and remove ILC03.4.1l1 power from the associated block valves and be in at least M.....

ODE 311T SH.TDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 4

_081 reduce reactor coolant system average tompcraturo below 3S 'F*

JA3.4-39 within l2thc fo!!o;-ig--- hours.

4. With cnc block valve inopcrable, within one hour eitherI restere the block valve to OPERABLE sta-tns-e--place its LC03.4.11 associated PORV in manual control. Restore the block Cond C valve to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

ILC03 .4 .11 Required Actions C.1 and C.2 do not apply when block valve is inoperable solely as a result of comply~irng with Required Actions S+J,,.+.4-11.0 Cond C D,2 and E.2.

L Note R-9 P1 Current TS 9 of 30 Markup for P1 ITS Part C

T-S. 3 .1 4 REV 135 5/4/98 E)VerrflQý'

oLC03.4.11 5. With both block valves inoperable, within two hours-.e-e Cond F I hour either roztere the block valves to GPERABLE status lrpaee-t~h-e PGR-V in manual control. Restore at- 4-as one block valve to OPERABLE status ;within the next hour.

IM.-23 LC03 .4.11 Required Action F.1l does nIot apply when block valve is Cond Note F ~inoperable solely as a result of complying with Required,

~l II _____

Actions B.2 and E.2.

R-9 PI Current TS 10 of 30 Markup for PI ITS Part C

Ts. 3. 1 Overf lew eeontinued)

(2) Reactor Coolant System average temperature greater than or equal to the temperature specified in the PTLR for II4.2 disabling both safety injection pumps and below the Over Pressure Protection System Enable Temperature specified in the PTLR 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 JLC03.4. 12 Over Pressure Protection SystemEnable Temperature specified in the PTLR; with a maximum of one SI pump capable of injecting into the RCS and the ECCS 52 accumulators** isolated, both pressurizer power operated relief valves (PORVs) shall be OPERABLE (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 asscciated block valve opcn, and the associated backup air -53 supply chargod LC03.4. 12 If both SI pumps are capable of injecting into the RCS, 5l Cond A prompt action shall be taken to make one incapable of injecting into the RCS.

If an ECCS accumulator is not isolated as required, isolate LC03 . 4 . 12 the affected accumulator within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or increase the RCS Cond B, CI cold leg temperature > LTOP enable temperature specified in the PTLR within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or depressurize the accumulator to less than the maximum RCS pressure for the existing cold legL temperature allowed in the PTLR within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . --

SIMB3.4-52 R-9 ILC03 .4. 12 **Accumulator isolation is only required when accumulator NOTE 2 pressure is greater than or equal to the maximum RCS 2 pres sure for the existing RCS cold leg temperature allowed .4-54 by the P/T limit curves in the PTLR.

1M3 .4 -5 FTE SR3 . 4.12 . 1 New SRs, Verify no more than one SI pump capable of SR3 .4. 12 .2 injecting into RCS every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , verify accumulator isolation SR3.4.12.3 valve isolated once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter, verify PORV block valve open every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

R-9 PI Current TS 11 of 30 Markup for PI ITS Part C

TES.3. i5 REV 141 12/17/98

.1.A..2.c.(2).

LCO3.4. (a) One PORV may be inoperable for 7 days.

Cond D 12I IL3.4-501 ILCO3 .4. 12 If these conditions cannot be met, be in MODE 5 within 8 Cond E I hours and depressurize and vent the reactor coolant system through at least a 3 square inch vent within t-he ne*-t*24 hours. IL'3"4-S° 1 LCO3.4. 12 (b) With both PORVs inoperable, be in MODE 5 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Cond E and complete depressurization and venting of the RCS through at least a 3 square inch vent within 12S hours.

(3) Reactor Coolant System average temperature below the temperature ILCO3-.4.-13I specified in the PTLR for disabling both safety injection pumps With Reactor Coolant System temperature less than the LCO3.4.13 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 O, larger vent; with both pumps incapableM 5 of injecting into the RCS and the ECCS accumulators**

isolatedboth 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 1 open, and the asseciated backup air supply .hargd.

LCO3.4.13 Cond D (a) One PORV may be inoperable for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

LCO3 .4.13 If these conditions cannot be met, depressurize and vent the Cond E reactor coolant system through at least a 3 square inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

LCO3. 4.13 (b) With both PORVs inoperable, complete depressurization and venting of the RCS through at least a 3 square inch vent Cond E within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

ILCO3.4.13 Cond A I If one or bth SI pumps are capable of injecting intol the RCS, prompt action shall be taken to make both incapable of injecting into the RCS.

ILC03.4.13 CondB I If an ECCS accumulator is not isolated as required1 isolate the affected accumulator within I hour, or increase the RCS cold leg temperature > LTOP enable M3 4-521 temperature specified in the PTLR within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or depressurize the accumulator to less than the maximum RCS pressure for the existing cold leg temperature allowed in the PTLR within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SR3 . 4. 13 . 1 N~ew SRs, Verify no SI capable of injecting into RCS every SR3.4.13.2 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , verify accumulator isolation valve isolated once SR3.4.13.3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter, verify RCS SR3.4 .13.4 vent >.3 square inches 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for unlocked open vent val-ve(s) and 31 days for other vent path(s), verify PORV block valve open every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . ____

R-9 PI Current TS 12 of 30 Markup for PI ITS Part C

3..A Roactor Ccolamt vent System

a. A reactor shall not ho mado or maintainod critical nor 0 shall reactor coolant system average temperature exceed 200 F unloss Reactor Coolant Vent Systom paths from both the roactor vossol hood and prossurizer steam space arc .PEP...

and closod (o opt..as spocifiod in .l..A.3.b and 3....A..c

h. ie ew.. . STAR~TUP During . . .. OPER.TION and .. POWER . OPEPMION91, any ono of the to 10;;).. . .

. . . of..... inoporability

. conditions . , may oxEist for oach unit following providod STARTUP OPERATION i icniudutlOE~L is rostorod. if any ono of thoso conditions is not rosooro to an OPEPALBE status within 30 days, ho in at loast HOTI SHUJT-DOWN within tho noiet 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTIDOWN within tho following 30 hour3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months s:

(I) Both of the parallel vent valves in the reactor vessel head vent path inoperable, or (2) Both of the parallel vent valves in tho prossurizor vont path inoporablo, or (a) Tho vont valve to tho prossurizor roliof tank disohorgo lino inoporablo, o~r (4) Tho vont valvo to tho contaminant atmospheric dis.hargo lino inopo-rablo.

c. With no Roeaetr Colant Vot gSysteom path OEL, rostoro.at loast ono vont path to status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or b

.PE..LE in at least HOT- SHUTDOWN within tho next 6 hours and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

PI Current TS 13 of 30 Markup for PI ITS Part C

T. .3."* G REV 135 5/41/98 LC03.4. 2 New Specification 3.4.2, RCS Minimum Temperature for SR3,4.2. I Criticality 3.1.B. Pressure/Temperature Limits

1. Reactor Coolant System

a. The Unit 1 and Unit 2 Reactor Coolant Systems (except the LCO3.4.3 3 pressurizer) temperature, pressure heatup rates, and cooldown rates shall be maintained within the limits specified in the Pressure and Temperature Limits report (PTLR) . A3.4-61

b. If these conditions cannot be satisfied in M.odes 1, 2. I - I

!l01 3, and 4, restore the temperature and/or pressure to within the limits within 30 minutes; perf orm an ongirnooring evaluation to deterio ef f ets of the out ef lii on i i n n to t u tral m 7~2: I Cord A iofnt oho r itRy act r Co lan Syc om; within 712hoursI determine that the Reactor Coolant System remains acceptable ------

for continued operation or be in at least MODE 3R*-G R-9 s...u..n.. within the next 6 hours and MODE Serduee the IA3.4-0 reactor coolant system average temperature and pressure to less than 2002F and 500 psig, respectively, within 36.t-h4. .4-3 LC 3 .4.3 following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

CordC If these conditions cannot be satisfied in other than 63 Mode 1, 2, 3, or 4, promptly initiate action to restore parameters to within limits and determine if the RCS is acceptable for continued operation prior to entering MODE 4.

New SR, Verify pressure, temperature, heatup, cooldown rates within PTLR limits

2. Pressurizer

ýý~-4 ~ ~ r ah~

' nt i rý ~q 1- - 1pim i ta to:

n

1. A maximum hoatup of !00OF in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period.

2. A maximum cooldown of 200'F in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period.

b. The.. r...rizer spray salln not ue used f tho emperaturre

.ffer.n.b

.twonthe prcssurizeran dtý.. spray fli is P-ifac teeenditmflnc ennoto be stisfiedr resoret eto temperatuare to wlthqIn tnc +lmlts w~tAIn zsv ffmj-i~tut-t; s ur z r

... pr.......... tha

... the....

.. pressurizer rcmainna S acceptable fer continued operation or be in at least HOT SHUTIDOWN within the..... 6 hr....... an the...

pressurizer pressure te less than 500 p-ig within the fo!Iswing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

PI Current TS 14 of 30 Markup for PI ITS Part C

TS.3.1 7 pV 91 1 o /27/89 3 . 1.B.3 . Steam Cenerators a The secondary side f the steam. generator must n*t be pressurized above 200 psig if the temperature of the steam 67 generator is below 790 F.

b. If these conditions cannot be satisfied, reduce the steam.

generator pressure to less than or equal t 200 psi-g within 30 minutes; perform. an engineering evaluation to dotermaine the effects of the over-pressurization on the structural intaegrity of tho steam generator, determine.

that the steamf generator remains acceptable for continued operation prior 0

to- increasing its temperature abeve 200 F.

PI Current TS 15 of 30 Markup for PI ITS Part C

TS.*3.

REV 91 10/27/89 3.1.C. REACTOR COOLANT SYSTEM LEAKAGE 1.Leakage Detection LCO 3.0.4 does not apply,

-. rage J 08 In Modes 1, 2, 3 and 4The reaetor coolant sys...

ILCO3.4.16 ] 0 temperature shall not exceed 200 F unless at least two means of reactor coolant system leakage detection shall be OPERABLE, one of which must depend on the detection of radionuclides in the LC03.4.16 containment. The other means is the containment sump A pump run time instrumentation., one means of leakage detection may be inoperable for 30 days provided daily RCS water balance 6 Cond A Cond B inventories are performed and, if radionuclide detection is Cond C inoperable, daily compensatory sampling shall be performed. If these conditions cannot be met, be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ..

ILC03.4. 16 If these conditions cannot be satisfied (all required monitorsI Cond D inoperable), immediately enter LCO 3.0.3 within ene hoiur initiate the action n...ssary to placo the affee t- -fn-4nO SHUT.OWN, and be in at loast !TOT HUTJED-WN viithin the n"xt rC" hours and in COLýD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SR3.4.16. New SRs, Channel Check, COT and Calibration of containment 72 SR3.4.16. radionuclide instrumentation and Calibration of sump pump SR3. 4 .16 run time instrumentation.

2. Leakage Limitations The following leakage limitations are applicable whenever in 08 ELC03.4.14 Modes 1, 2, 3 and 4the reactor coolant systomavorago te F~pe r aE+/-ei--eeiceeeds 200'

a. If the LAzA Ileakage rate, fromo-ether than controlled LC3 .4.1 leakage sertees, such as the reactor coolant pump centr!ele4 Cond A leakage seals, exceeds 1 gpm and the source of the leakage is not identified within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of leak detection, be in at least MODE 340T SHUT.O.. within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

LCO3.4.14 If the source of leakage is not identified within 54aI tional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , be in MODE 5COLD SHUTDOWN within 84teeA3.4-39I following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. If the sources of leakage are identified and the results of the ovaluations are that continued oporation is safe, LCO3.4.14 leakage, other than operation of the reactor with a total leakage from controlled sources, not exceeding 10 gpm shall be permitted (except as specified in 3.1.C.2.c below).

If it is determined that leakage exists through a fault ILC03 . 4 .14 which has developed in a Reactor Coolant System component Cond D I body, pipe wall, vessel wall, or pipe weld, and that the IM3.4-421 fault cannot be isolated, within one hour initiate action to placo the unit in HOT. SHUTDOWN and be in at least MODE 3 IA3.4-081 HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and be in MODE 5 GOLD SHUTDOWN within 3e-the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months and take corrective action prior to resumption of unit operation.

PJ Current TS 16 of 30 Markup for P1 ITS Part C

TC.3.l8 low

d. If the total LEAKAGElokalc, ether than leakago from A3.4 -73 eentrolled souroce, exceeds 10 gpm, within one hour b

initiate action to place the unit in HOT SHUTDOWN and LC03.4.14 Cond C in at least A3.408 MODE 3 S GRUTJTDOWN-within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . If the condition is not corrected within 14an additional 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , be in MODE - SHUT-DGOWI within 44the following A3439 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and. r-main in OLLD siUiLD)Uw' urntil+/- the u dition is 77 eerreeted.

PI Current TS 17 of 30 Markup for PI ITS Part C

TS. 3.1 9

]

e. If the total reactor coolant system to secondary coolant system leakage through any one steam generator of a unit exceeds 150 gallons per day (GPD), within ono heur initiate actioen to plaee the unit in HOT SHUTDOWN and be I

in at least MODE 3 m-HGT S1UGBWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and I~

be inMODE 5 COLD SHUJTDOW.N hours S3.l4.1 and perform an inservice steam generator tube inspection in accordance with Steam Generator Program Tehia JA.-78 speeifieatien 61.12.

3. Pressure Isolation Valve Leakage

[C03.4.15 Leakage through the pressure isolation valves shall not exceed L...............Jthe maximum allowable leakage specified in Specification 4.3 when in MODEtS 1, 2, 3, and 4 re-ae-te-r-eelant system aver-~

tomnoeratur exeds 200 0F.

4-081 LC03 .4 15 If the maximum allowable leakage is exceeded, isolate the high CodA pressure portion of the system in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and restore the PIy to within limits in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , Valves used to isolate the Ply flow I1 path must meet SR 3.4.15.1 leakage limits. within eno hour initiate the actio nccrytplothuit nHO SHTDOW4N,

[ L3..15 ConB a-nd If these requirements can not be met, be in at least MODE 3 HOT- SIUT-DOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 5 COLD SHUTDOWN within 136the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

PI Current TS 18 of 30 P1CrretTS PI ITS Part C o30Markup 8for

REV 147 12/7/98

-3.i.D. RCS Specific L4AXIMUi COOLGNT-ACTIVITY

1. The specific activity of the primary coolant (.xcopt as spocifi**

ILC03.4.17 1 in 3.1.D.2 and 3 bcolw) shall be limited-4te:

ISR3-4.-17.2 a. Less than or equal to 1.0 microcuries per gram DOSE EQUIVALENT 1-131, and SR3.4.17

b. Less than or equal to 100/R microcuries per gram of gross radioactivity. -08r R-9

2. If a reactor is in Modes 1 and 2,or critical or tha ..

Modes 3., with reactor coolant system average temperature 0

i-s--greater than or equal to 500 F:

a. With the specific activity of the primary coolant ILC03.4.17 greater than 1.0 microcurie per gram DOSE EQUIVALENT 1-131, verify once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to'be within the Cond A acceptable region of Figure TS.3.l-3q, for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown on Figure TS.3.1-3, LC03.4.17 Cond C I

I the reactor shall be in MODE 3 with shutdown and reactorl coolant system average temperature cooled to below 500'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the gross specific activity of the primary coolan LC03. 4. 1 greater than 100/2 microcurie per gram, the reactor 1A3 -8 Cond B shall be in MODE 3 with shutdon,ý and reactor coolant system average temperature cooled to below 500'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3. if a r ca ct or is at- v or ab .... G L ...

G;RZ w-itT the... sp eeif .

activity of thc primary coolant greater than 1.0 rmicrocuro par gram DOSE EQUIVALENT 1 131 or groator than 100/E mic.rocuri. per gram, porform the sampling and analysis roquiromonts of itom 4a activity of the primary of Table 4.1 2B until the spcific 4 r*-Cr'rc 4c within its limits.

'\T-t nrcs;:=Picuro

4=p' T3.l 3 and TS.3 .l 12 .

PI Current TS 19 of 30 Markup for PI ITS Part C

Figure TO.3.1 3 REV 80 11/14/86 E '! f ,I ,I f i ll i' 444 II I I44

- l111,o liAI ,,

Ii I ... 8......i..OPERAIO IIN F

Ii ; i il' ! t l

"'i~ ; lI l l-i l i

'*

i, ,I.i '; 4111 liii ti I . t ,! i 4i I i 2W0 II 44 !UNACCEPTABLE

, ;* i ! t\ i 1iiViI OPERATION , li 0 fillil :*-l..rl 1ii'rirl. T, , .FF*

1_

4 I1 44 I 4 4 4 I I I 4 4 i l 0W

),; 1 : 1 1 1 1i1 1 1i1 1i1 !,H4 44111I! Ii 4444 444 7

I *1 12-7I K I i L L I I I I I I C7 81l 1 ' I!

l I '4t ! I I i i I 0

0 0

I .- f

-il ill+

11 4i ii IY III 1 fll 1 Ill,

,;i#

I II I InLnI * . 'III ii i ' t ] 4'i 0n II t i I : S*I I I l

_LJ i Ir .i ,I I;!* ii l T: l'i*-rn i ; i1 11F~ *iI i II' l,, l'4, , , 4 ; i1 l1 11i i i

.i4 44t i 1 i i i I%. 444I 44 II II w ACCEPTABLE -J---4--- 4-4-4-4-4-4--i- i--

I I%

U-Su 4,4 I I I4 I 1 OPERATION 4I 4','

flu - . -9 I !

i I I I T 4- _________

5 444I I1 I4444 w

w

f1 IiI ~t 111414 i 4 4i 8 *"'i i i 09 0

I -, i il-t

~ t li *I ;:

i I - i- l Ii ! i i , .1, - , ii 7+7 -!i t"1 W

20 30 40 5o 60 70 80 90 100 PERCENT OF RATED THERMAL POWER Figure TG3. 3.4.17-1 (Page 1 of 1)

Reactor Coolant DOSE EQUIVALENT 1-131 P"rima*y Gelant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primar*y Coolant Specific A rti ,it ,"> 1 0 , t i ! .ra . D o se E u i'v,,* ,

Rt 1-43 1 PI Current TS 20 of 30 Markup for PI ITS Part C

R-2 Adresed El sewhere

.3.A.2.g. T2fte -valve position mocnitor iighfts or alarms tor motor -operate72 valves specified in 3.3.A.I.g above may be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided the val-ve position is verified onc~e each shift.I

3. A maximum of one safety injection pump shall be capable of injecting into the RCS whenever in MODE 4 with RCS A383 JLC03.4.12 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 JLCO3 LCO3. .4.4.13 12 following conditions is met:

NOTE I (a) There is a steam or gas bubble in the pressurizer and an isolation valve between the SI pump and the RCS is shut, or c.ater vessel head is r..v.d. IA3"4-127 (b) The.

R-9

4. No safety injection pumps*** shall be capable of injecting ILCO3"4"13 1 into the RCS whenever in MODE 4 with RCS temperature is l*ss than the temperature specified in the PTLR for disabling JA3.4-83

T I

both safety injection pumps, MODE S when the SG primary system manways and pressurizer manway are closed and secured in position, MODE 6 when the reactor vessel head is on and the SG primary system manways and pressurizer manlway a~re may be run as specified in 3.3.A.3 and 3.l.A.l.d. (2)). IR-9 ILC03.4.12 II

5. Both reactor coolant system accumulators shall be isolated*

LC03 .4. 13 whenever RCS temperature is less than the Over Pressure Protection System Enable Temperature specified in the PTLR.

LC03 .4 .12 NOTE 2 LCO3 .4. 13 *This specification does not apply whenever the reactor coolant system NOTE 3 accumulators are depressurized or the reactor vessel head is removed.

LC03 .4. 12 LCO3.4. 13 **Other SI system tests and operations may also be conducted under these NOTE 1 conditions.

ILCO3-4.13

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

PI Current TS 21 of 30 Markup for PI ITS Part C

Elsewhere R-2 I ISR3.4.12. 4 SR3 .4. 12. 5 SR3.4. 13.5 SR35.4 13O.s6 Miao

25. Overpressure Mitigation M R N.A. 4(38) 5 N.A. R Addressed Elsewherel

-- T-3 P1 Current TS 22 of 30 Markup for PI ITS Part C

TABLE TS.4.1-1C (Page 4 of 4)

TABLE NOTATIONS FREQUENCY NOTATION NOTATION FREQUENCY S Shift D Daily W Weekly M Monthly Q Quarterly S/U Prior to each startup Y Yearly R Each refueling shutdown N.A. Not applicable TABLE NOTATION Address ed Elsewhere lover Pressure Protection System Enable Temperature specified in the PTLR.

nrrrrym,rrrrnflfltVrtfl(flW*3A&Jflz,4jA 4i ... A$.J P

(38) Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of When the reactor coolant system average temperature is less than the PI Current TS 23 of 30 Markup for PI ITS Part C

FA34-00 1 R-2 Addressed Table TS.4.1 2A (Pago 1 of-2) l s ew h e r e

.E.............................

3. Pressurizer Safety Verify OPERABLE in accordance Per ASME Code, Section XI Inservice Testing with the Inservice Testing Program SR3.4.10.1lValves Program (+ 3%). Following testing, lift settings shall Addressed be within +1% ElsewhereI

6. Pressurizer PORV Functional cavity ~is fl~ooded.

Quarterly, unless the block valve has It Block Valves been closed per Specification SR3.4.1 3.1.A.2.c. (1) . (b) .2 or 3.1.A.2.c. (1) . (b) .3.

Every onths L Kb D

Pressurizer PORVs Functional 386"-..

SR3.4. 11 24 l I Note -Only required to be performed in Modes 1 and 2.

R-9 SR3 .4 . 11 . 1 iiA3.41131 S114 and 3.4.11.2 PI Current TS 24 of 30 Markup for PI ITS Part C

Table TSX1- 2A (Page 2 of 2)

MINIMUM FREQUENCIES FOR EQUIPMENT TESTS FSAR Sect.

Frequency Reference Equipment Test

8. Deleted SR3.4.14.l

9. Primary System Leakage Evaluate Daily A3 .4-:It12]

_R-9

-- - - - J-IAddressed Elsewhere 0

PI Current TS 25 of 30 Markup for PI ITS Part C

Table TS.4.1 2B

~1j(Page

.*-oo~ ~~... .....

.... 1 of;/;,

2)

F3 4-0 TABLE TS.4.1-2B: R-2 REV 147 12/17/98 MINIMUM FREQUENCIES FOR SAMPLING TESTS TEST FREQUENCY IL3.4-88 iSR3.4.17.11 1. RCS Gross I s/week IAl4-25 Activity Determination SR3.4.17.2

2. RCS Isotopic Analysis for DOSE 1/14 days (when at p.w.r) and Note EQUIVALENT 1-131 Concentration only required to be performed in Mode 1 ILC° .17.3 3. RCS Radiochemistry E determination I/h6 menth- 184 days,(l) -(when-a-t peweir) from a sample taken in MODE 1 after a minimum of 2 EFPD and 4-24 20 days of Mode 1 operation have elapsed since the reactor was last subcritical for _> 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

I R-9 i i I

4. RCS Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever JL034.1 the specific activity ex Including 1-131, 1-133, and 1-135 ceeds 1.0 uCi/gram DOSE EQUIVALENT 1-131 or 100/E uCi/gram I (at or above 50o0 F cold shutdown), and b) One sample between 2 and 6 ISR3 .4 .17 .2 hours following THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period (above hot shutdown) r :PC P T ff fAlCBAPP r P1 q #I= PV - z*L349 iSR3.4.17.2I

6. 11CC Tritium Activity Weekly LR3.4-94 I

7. Deleted

8. 11CC Boren Coneentration 2/Week (01) LR3.4-96 AdCaressea aElsewhere PI Current TS 26 of 30 Markup for PI ITS Part C

Table T-8.1.1 2B (Page 2 of 2)

REV 129 6/112/97 TABLE T".4.1 2B Addrse Notes:

Sample to be taken Not required to be performed until 31 days3, Iafter a minimum Note 1 have elapsed since reactor was SR3-4-17"3 1I of 2 EFPD and 20 days of ,

POWER PEPATIGNMode last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer. M3 -7

R-9

2. T. d.torm..in a . of c..rrsion products hav7ing a half

.tivity life greater than 3 minutos. ILR3"4-94 j.During P-EFUELING, the boron concentrationh shall be v~erified by chemical

'I c-meood 5 days. JR3"4-96

4. The mamimum interval between analyses sona+/-: ll(

IAddressed I P1 Current TS 27 of 30 Markup for PI ITS Part C

A3.4 -00 RES.4.3 1 i 116 3/10/95 R-2 4-4-PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES ..............

_r ...... m -- _- *1 ..... -' ......

- -ý +-In,- nF-i-fermed en the primary _eeelant: svs--l" i1.olation valves to veritv ooorabilitv.

A303 To increase the reliability of primary coolant sys Tcm prcussrc usii; thereby 4=P R i icing the potential of an intersystem 1 ess ef e7elant accident.

Specification Periodic leakage testing of each of the following valves shall be power SR3.4.15.1 individually accomplished every 24 months, prior to resuming operation after ash t - plant is placed in the cold shutd"wn th condition for refueling, each time the plant is placed in MODE 5 a-eed IA3.4-08 I s.hutdown condition for 7 days or more if testing has not been accomplished in the preceding 9 months, and prior to returning the valve to service after maintenance, repair, or replacement work is performed:

Valve Number

-Allewable System Unit No. 1 Unit No. 2 Leakagl To- #6)croLIt SI 9 9 2S1 9 6 <5 p .34-971 Upper Plenumf S! 9 4 2S! 9 4 !* S gpmfl i 95S 2S! 95 !5 gpm S 1 93 261 93 *gpmR RHR to Loop -B Accumulator inj Line S! 6 2 2S1 6 2 5gpmfl To satisfy ALAR requirements, leakage may be measurod indireetly (as from the performanec of pressure indieaters) if aeccmplished in aeeordanco with approved procoduroc and supported by eeomputations showing that thoe method is capable o-f dcmenstrating valvoe complianco with the leakage eriteria.

NOTES:

1. Leakage rates less than or equal to 0,5 ee--gpm per nominal inch of valve size-are acceptable_ at an RCS pressure of z 2215 psig and

2255 psig.

SR3.7.15.

2. Leakage rates greater than on*, but loss than or equal to five gpm arc considored cron ..... if the acc-ptable ....

maucraccpal an latetr ... m..asurod

............. rate has not ..... dd the DrevOUSmeasured rate by an amount whieh reueuu 5 9 pm by a X;= r r

=*m*.r Otherwise the leakage T.VcR* iq Rco si-dedred unaceeptable.

. Leakage rates greater than five gpm are considered unacceptable.

- ini~mum elartcrorntial test pressure snai- neE- be less than 150 psid. lLR3"4-97 I PI Current TS 28 of 30 Markup for PI ITS Part C

3

[A3. 4- 0 REV,:1.6 9 10,127/89 Addressed Elsewhere 4.6.

C. Pressurizer Heater Emergency Power Supply ISR3.4.9.2 The emergency pressurizer heater supply shall be demonstrated OPERABLE at least once every 24i-- months by transferring Backup He.ater Croup "El" from its norma 1 bus to its safeur bus and enorgizing the heators.

PI Current TS 29 of 30 Markup for PI ITS Part C

TS. 4.lS8 1 REV 91 10/27/39 T.his REACTOR COOLT2NT VENT SYSTEM PA-THS 36 Appliailt Applies te tho survc:illancc pcrfcrmod on tht Roactor Coolant vono yuzejLLI orify OPERABILITY.

To assure that the capability exists to vent nono*ndcnsiblo gases from the Reaetor- Colant System -tý ht cold inhibit natural circulation eoro cooling.

A. Vent Path Operability Each Racteor Coolant Vcnt System path shall bc dcmonstratcd OPE.LE prier to commencing START-UP OPEPATION after each rofucling by:

1. verifying all manual isolation valves in each vent path arc blooked and tagged in the open position.

2. Cycling ea"oh solenoid operated valve in the vent least o ne somplete cycle of fu 1 1 travel from the centrol reom.,

B. S'ystem F! Tt Flow shall be verified through cach Roactor Coolant Vcnt System fellewing each refucling.

PI Current TS 30 of 30 Markup for PI ITS Part C

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 A 08 CTS 3.1.A.1.a(2), 3.1.A.1.b(1), 3.1.A.1.b (2), 3.1.A.l.c, 3.1.A.l.d, 3.1.A.2.a(1), 3.1 .A.2.a(2), 3.1.A.2.a(3), 3.1.A.2.b(1),

3.1.A.2.c(1), 3.1 .A.2.c(1 )(a), 3.1.A.2.c(1 )(b), 3.1 .A.2.c(1 )(b)3, 3.1.B.1.b, 3.1.C.1,3.1.C.2, 3.1.C.3, 3.1.D.2, 3.1 .D.2.a, 3.1.D.2.b, 3.10.J and 4.3. The CTS contain prose descriptions of the modes of applicability and conditions in which the plant must be placed or remain when equipment is inoperable.

These descriptions have been replaced with the equivalent MODES of applicability for the ITS. Since the modes of applicability and plant conditions in which the plant is placed or remains have not changed this is an administrative change.

9 Not used.

10 Not used.

M 11 A new SR, 3.4.4.1, is included in conformance with ISTS which requires verification of each RCS loop in operation while at power.

This is more restrictive since PI CTS do not require this surveillance. This more restrictive change is included to make the PI ITS complete.

Prairie Island Units 1 and 2 5 1/2/02

Part D Package 3.4 Part D NSHD Change Number Discussion of Change Category 3.4 M 21 Three new SRs, 3.4.5.1, 3.4.5.2 and 3.4.5.3, have been added in conformance with the guidance of NUREG-1431. These SRs will require verification that the required RCS loop is in operation, verify the SG is capable of decay heat removal, and verify power is available to the RCP that is not operating. The SR requirement for verifying the SG is capable of decay heat removal is consistent with WOG-1 55, Information Notice 95-35, TSTF 114, and the intent of the CTS. This is more restrictive since PI CTS does not specifically require these surveillances.

These more restrictive changes are included to make the PI ITS complete and do not introduce any unsafe plant conditions.

A 22 CTS 3.1.A.1 .b(3). CTS requires the plant to shut down to MODE 4 if the Action Statement requirements are not met. These requirements are not included in the ITS, since once the Action Statement requirements are not met, the plant will enter LCO 3.0.3 which essentially requires the same actions as CTS.

Prairie Island Units I and 2 9 1/2/02

Part D Package 3.4 Part D NSHD Change Category Number Discussion of Change 3.4-23 Not used.

LR 24 CTS 3.1.A.l.c(1), 3.1.A.l.c(2), 3.1 .A.l.d(1) and 3.1.A.2.a(1).

The CTS description of equipment required for system operability has been relocated to the Bases which is consistent with the format and guidance of NUREG-1 431. As an example, the CTS refers to a reactor coolant pump (RCP) and associated steam generator (SG) as needing to be OPERABLE during specific plant conditions. The ITS Bases defines that a RCS loop consists of the RCP and associated SG. Therefore, the CTS has been revised to state RCS loop instead of listing specific components that make up the system. The individual components making up the system are considered to be details placed in the Bases Section. This change is acceptable since the system is required to be operable in accordance with the definition of OPERABILITY and details of the specific equipment are unnecessary in the specification. Since the Bases are under licensee control, this is a less restrictive change.

Prairie Island Units I and 2 10 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4-25 Not used.

M 26 CTS 3.1.A.1 .c. CTS has been revised to require that the SG is capable of decay heat removal. This change is a more restrictive change since it is a new requirement by providing specific operability requirements on the SG that were not in the CTS. This change is acceptable since it specifically requires the SG to maintain sufficient secondary water inventory to provide a heat sink for decay heat removal in the event the RHR loop is inoperable and not in operation. This change is consistent with NUREG-1431.

27 Not used.

Prairie Island Units 1 and 2 11 1/2/02

Part D Package 3.4 NSHID Change Category Number Discussion of Change 3.4 A 28 CTS 3.1.A.1.c. CTS has a single specification which requires two methods of cooling the RCS to be operable with one in operation when in MODE 4 and MODE 5, loops filled. This can be any combination of RCS loops or RHR trains. During plant heatup from MODE 5 to MODE 4, one RCS loop may be in operation with one or both RHR loops operable but not operating. NUREG-1431 splits MODE 4 and MODE 5, loops filled into two specifications, LCO 3.4.6 and LCO 3.4.7. Because these MODES or other specified conditions of applicability have been split into two Specifications, LCO Note 4 in 3.4.7 is required to allow one RCS loop in operation with one or both RHR loops operable but not operating. This configuration is required to allow a smooth transition to MODE 4. Since the configuration allowed by Note 4 currently is allowed by CTS, explicit statement of this Note is an administrative change. This change is consistent with the ISTS.

29 Not used.

30 Not used.

Prairie Island Units 1 and 2 12 12/11/00

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 M 31 Three new SRs, 3.4.6.1, 3.4.6.2, 3.4.6.3, have been added in conformance with NUREG-1431. These SRs will require verification that the required RHR or RCS loop is in operation, verify the SG is capable of decay heat removal, and verify power is available to the required RHR or RCP that is not operating. The SR requirements for verifying that the SG is capable of decay heat removal is consistent with WOG-1 55, Information Notice 95-35, TSTF-1 14, and the intent of the CTS.

This is more restrictive since P1 CTS does not specifically require these surveillances. These more restrictive changes are included to make the PI ITS complete and do not introduce any unsafe plant operations.

M 32 Three new SRs, 3.4.7.1, 3.4.7.2, 3.4.7.3, have been added in conformance with NUREG-1431. These SRs will require verification that one RHR loop is in operation, verify the SG is capable of decay heat removal, and verify power is available to the RHR pump that is not operating. The SR requirements for verifying that the SG is capable of decay heat removal is consistent with WOG-1 55, Information Notice 95-35, TSTF 114, and the intent of the CTS. This is more restrictive since PI CTS does not specifically require these surveillances.

These more restrictive changes are included to make the PI ITS complete and they do not introduce any unsafe plant operations.

Prairie Island Units 1 and 2 13 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 M 33 CTS 3.1.A.1.d. An additional required action has been included in the action statement when both RHR loops are inoperable to suspend operations involving reduction in RCS boron concentration. This change is included to make the ITS complete and may improve plant safety. Since this change places additional restrictions on plant operations it is more restrictive. This change is also consistent with the guidance of NUREG-1431.

M 34 CTS 3.1.A.1 .d. CTS allows the pumps to be shutdown for I hour. To be consistent with NUREG-1431, this is further restricted to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. In addition, the CTS Note is modified by addition of a restriction that, in MODE 5 with the loops not filled, no RCS draining operations are permitted when all pumps are shutdown. Restrictions on boron concentration changes have also been included. These changes were made for consistency with the guidance of NUREG-1431 as modified by approved TSTF-286, Revision 2.

These changes are more restrictive since they impose requirements on plant operations that are not in the CTS.

These changes are acceptable since they may improve plant safety.

35 Not used.

Prairie Island Units 1 and 2 14 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4-L 36 CTS 3.1.A.1 .d. One RHR loop may be made inoperable for surveillance testing while the other RHR loop is OPERABLE and in operation. This change is made for consistency with NUREG 1431. This is acceptable since the operating RHR loop provides sufficient heat removal capability and the loss of the operating RHR loop in the two hours allowed for testing is unlikely. This is less restrictive since this provision is not contained in the CTS.

M 37 Two new SRs, 3.4.8.1 and 3.4.8.2, have been added in conformance with NUREG-1431. These SRs will require verification that one RHR loop is in operation and verify power is available to the RHR pump that is not operating. This is more restrictive since PI CTS do not require these surveillances. These more restrictive changes are included to make the PI ITS complete. These changes do not place the plant in any unsafe operating conditions.

M 38 CTS 3.1.A.1.d. CTS allows the pumps to be shutdown for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

To be consistent with NUREG-1431, this is further restricted to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. Since this imposes additional restrictions on plant operations, this is a more restrictive change. This change is acceptable because this change may limit the time that all pumps are not operating and improve plant safety.

Prairie Island Units 1 and 2 15 12/11/00

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 A 39 CTS 3.1.A.2.a, 3.1.A.2.c.(1), 3.1.B.1.b and 3.1.C.2. The format for CTS and ITS fundamentally differ in the presentation of shutdown tracks but the Completion Time requirements are the same. The CTS format has been changed to the ITS format.

Since there is no net change in plant operations, this is an administrative change.

40 Not used.

M 41 CTS 3.1 .A.2.a(1). CTS for pressurizer OPERABILITY require a steam bubble in the pressurizer and group A and B heaters operable. The ITS will require pressurizer level less than or equal to the pressurizer high water level Allowable Value and two heater groups capable of being powered from an emergency power supply. Since the pressurizer high water level Allowable Value is significantly below the level required to assure that there is a steam bubble in the pressurizer, this is a more restrictive change.

This change has been made to be consistent with NUREG-1431 and current NRC guidance. This change is acceptable since it will not introduce any unsafe plant operating or test conditions.

Prairie Island Units 1 and 2 16 12/11/00

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4-M 42 CTS 3.1.A.2.a(3), 3.1.C.2.c, 3.1.C.2.d and 3.1.C.2.e. For consistency with NUREG-1431, the one hour to initiate actions necessary for shutdown has been deleted. This is more restrictive since the plant has one less hour prior to placing the plant in MODE 3. The time provided in the ITS is adequate to plan for and implement an orderly shutdown if required; thus this change is acceptable. This change does not place the plant in any unsafe operating conditions.

M 43 CTS 3.1.A.2.a(3). New Required Actions have been provided for the possibility that the pressurizer is inoperable. These actions incorporate proposed TSTF-87, Rev. 1. This is more restrictive since PI CTS do not require these actions. This more restrictive change is included to make the PI ITS complete.

M 44 A new SR, 3.4.9.1, has been added in conformance with the guidance of NUREG-1431. This SR will require periodic verification that the pressurizer water level is < 90%. This is more restrictive since PI CTS do not require this surveillance.

This more restrictive change is included to make the PI ITS complete.

Prairie Island Units I and 2 17 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 M 45 CTS 3.1.A.2.b(1). CTS 3.1.A.2.b(1) requires two RCS pressurizer safety valves (PSVs) to be operable when the RCS temperature is > 350'F. This change will require two RCS PSVs to be operable whenever both RCS cold leg temperatures are greater than the OPPS enable temperature specified in the PTLR. This change is consistent with the guidance of NUREG-1431 and will provide additional overpressure protection when the RCS is between 3500F and the OPPS enable temperature. Since this change will require more components to be operable for more plant conditions, this is a more restrictive change. This change is acceptable because it does not cause any unsafe plant operating or testing conditions.

A 46 CTS 3.1.A.2.b(1). P1 CTS License Amendment 123 allows pressurizer safety valves to be 2485 psig +/-3% when tested for operability. However, because of the particular wording in Specification 3.1 .A.2.b(1), it was not revised. The intent of Amendment 123 was that if the operators were aware the setting was outside 2485 psig + 1% then they would not continue with their startup. CTS Table 4.1-2A, Item 3 requires verification that the settings are within +/-3. Thus, the change in LCO 3.4.10 is not a technical change since CTS already allow +/-3% tolerance on the valve settings and this is simply an administrative change.

Prairie Island Units 1 and 2 18 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 L 47 CTS 3.1.A.2.b(2). In conformance with NUREG-1431 guidance, this specification which requires one pressurizer safety valve operable when the head is on the reactor vessel has been not been retained. Since the pressurizer safety valves do not provide overpressurization protection when the RCS temperature is below the LTOP enable temperature, this specification serves no purpose. NUREG-1431 requires PSVs to be operable when the RCS temperature exceeds the LTOP enable temperature. Below the LTOP enable temperature, NUREG-1431 requires redundant PORVs to be OPERABLE at the LTOP lift settings or the RCS to be vented. Thus, this specification has not been retained. This change results in less restrictions on plant operation.

Prairie Island Units 1 and 2 19 12/11/00

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 L 48 CTS 3.1.A.2.b(1). CTS requires the pressurizer safety valves (PSVs) to be operable in MODES 1, 2, and 3. A new Applicability Note has been included to be consistent with the guidance of NUREG-1431. This Note does not require the PSVs to meet the LCO lift setting limits in MODE 3, when the plant is started up, to allow adjusting the settings under hot conditions. Because this Note allows the PSVs to be potentially inoperable in MODE 3 until the PSVs can be tested and set, this change is less restrictive.

This change is acceptable since the Note also requires preliminary lift settings under cold conditions prior to heatup which is consistent with current plant practices and the time during which the valves are potentially inoperable for valve setting adjustments is limited to 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months by the Note. The Note may improve plant safety by allowing more accurate settings under hot conditions.

A 49 CTS 3.1.A.2.c(1)(b). CTS required action explicitly requires the operators to "either restore the PORV to OPERABLE status or" take other appropriate actions. The option to restore equipment to operable status is always an available means of exiting the ITS Required Actions and, under the Writers Guide, this option should not be stated unless there are no other actions which should be taken. Thus this statement and similar statements within this CTS paragraph are not included in the ITS. Since it is understood that these are viable operator actions, this is an administrative change.

Prairie Island Units l and 2 20 12/11/00

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4-L 50 CTS 3.1.A.2.c.(2). When the RCS temperature is greater than the SI pump disable temperature and less than the OPPS enable temperature, CTS requires the plant to be in MODE 5 with the RCS depressurized and vented through at least a 3 square inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when a PORV is not restored within the allowed outage time or both PORVs are inoperable.

This change will allow 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to place the plant in MODE 5, depressurize and vent the RCS under these conditions. This change is acceptable because 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is considered by the industry to be insufficient time to perform an orderly shutdown.

In addition, the requirement to place the unit in MODE 5 is consistent with the intent of the CTS. The additional time will maintain plant safety by allowing the operators to plan the shutdown and prevent challenges to plant systems which may initiate an overpressure event which the shutdown intends to prevent. This change is consistent with the guidance of NUREG-1431 as modified by approved TSTF-352, Revision 1.

Since this change allows more time to fully implement remedial action, this is a less restrictive change.

Prairie Island Units 1 and 2 21 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 M 51 CTS 3.1.A.2.c(2) and 3.1.A.2.c(3). Restrictions on Sl pump operability isolation have been included for clarity. These requirements are specified in CTS LCO 3.3. New Action Statements are included for the possibility one (Specification 3.4.13) or two (Specification 3.4.12) SI pumps are capable of injecting into the RCS. CTS does not contain these action statement requirements; thus in this event CTS LCO 3.0.C (equivalent to PI ITS 3.0.3) would be entered which would allow 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for action to be taken. Under these circumstances, 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months would be an inappropriate response time. The new Action Statements require immediate action which is more restrictive than CTS. This more restrictive change is included to make the P1 ITS complete.

M 52 CTS 3.1.A.2.c(2) and 3.1.A.2.c(3). Restrictions on ECCS accumulator isolation have been included for clarity. These requirements are specified in CTS LCO 3.3. Associated action statements, clarifying notes for exceptions and applicability, and surveillance requirements, SRs 3.4.12.2 and 3.4.13.2, have been added. Adding LCO 3.4.12 and 3.4.13, Required Action C also added a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time that was not previously in the CTS. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time is acceptable to PI since it is based on operating experience that the associated activities of Condition C can be accomplished in this time period and the low probability of an event requiring LTOP during the allowed times. Since these are added requirements they are more restrictive on plant operations.

These requirements have been included to make the PI ITS complete and consistent with the guidance of NUREG-1431.

Prairie Island Units 1 and 2 22 1/2/02

Part D Package 3.4 Part D NSHD Change Category Number Discussion of Change 3.4 LR 53 CTS 3.1.A.2.c(2) and 3.1 .A.2.c(3). The specific status of associated equipment has been relocated to the Bases for consistency with NUREG-1431. These provisions are assumed to be part of the OPERABILITY requirement for the PORVs. Since ITS Bases (under the Bases Control Program in Section 5.5 of the ITS) is licensee controlled, relocation of CTS requirements to the Bases is a less restrictive change.

M 54 New SRs, 3.4.12.1,3.4.12.2, 3.4.12.3, 3.4.13.1, 3.4.13.2, 3.4.13.3, and 3.4.13.4, have been added in conformance with NUREG-1431. These SRs require verification that only one (or no) SI pump capable of injecting, RCS is vented and PORV block valves are open as applicable for the MODES of operation and the method of providing overpressurization protection. Adding these SRs also added new Frequencies that are not in the CTS. Evaluating these new Frequencies, PI believes them to be applicable to our plant. The new Frequencies provide adequate times considering administrative controls available to the operator, using other indications and alarms available to the operator in the control room to verify required status of the equipment, and the unlikelyhood of a low temperature overpressure event occurring during this time.

These SRs are more restrictive on plant operations since CTS do not require these surveillances. This is consistent with the guidance of NUREG-1431. These more restrictive changes are included to make the PI ITS complete.

55 Not used.

Prairie Island Units I and 2 23 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 R 56 CTS 3.1 .A.3 and 4.18. The reactor vessel head vent system specifications, CTS 3.1.A.3 and associated surveillance requirements in TS 4.18, are not included in the PI ITS since this system does not meet the 10CFR50.36 Technical Specification Selection Criteria. These vents are designed to exhaust noncondensable gases and steam from the RCS which could inhibit natural circulation following an accident with an extended loss of offsite power. Credit for this vent system is not assumed in the safety analyses nor in the PI IPE. Therefore, the reactor vessel head vent does not meet the TSSC and these requirements have been relocated to the TRM which is maintained under the regulatory controls of 10CFR50.59. This is consistent with the guidance of NUREG-1431 which does not include specifications for the reactor vessel head vent system.

Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

The reactor coolant head vents are not installed instrumentation, nor are they used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a Design Basis Accident (DBA). Therefore, the reactor coolant head vents do not meet criterion 1.

Criterion 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.

The reactor head vents are not process variables, design features, or operating restrictions that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a Prairie Island Units 1 and 2 24 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 R 56 (continued) challenge to the integrity of a fission product barrier. P1 does not take any credit for the reactor head vents in any accident analysis.

Therefore, the reactor head vents do not meet the criterion 2.

Criterion 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.

The reactor head vents are not structures, systems or components that are part of the primary success path which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, the reactor head vents do not meet criterion 3.

Criterion 4: A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The reactor head vents are not considered in the plant IPE and is not a system which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. Therefore, the reactor head vents do not meet criterion 4.

Therefore, since the screening criteria have not been satisfied, the reactor head vent LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications. The reactor head vents will be relocated to the TRM which is controlled by the provisions of 10CFR50.59.

Prairie Island Units 1 and 2 25 1/2/02

PartD Package 3.4 P rt D NSHD Change Number Discussion of Change Category 3.4-M 57 New Specification. In conformance with the guidance of NUREG 1431, a new specification, RCS Minimum Temperature for Criticality, is included in the PI ITS. CTS Figure 3.1-1 does include a line labeled "Criticality Limit"; however there are no action statements or applicability associated with this line on the curve.

The Bases for this new specification provide four considerations which the minimum temperature for criticality should satisfy.

Based on PI calculations and consistency with the Low-Low Tave setpoint, this LCO will specify 540°F as the minimum allowable temperature. Compliance with this specification will assure that plant conditions are conservative with respect to the initial conditions assumed in the safety analyses. Since this is a new specification, it does place additional restrictions on plant operations and is thus categorized as more restrictive. This more restrictive change is included to make the PI ITS complete.

58 Not used.

59 Not used.

60 Not used.

Prairie Island Units 1 and 2 26 12/11/00

Part 0 Package 3.4 NSHD Change Category Number Discussion of Change 3.4 A 61 CTS 3.1.B.1.b. The requirements for maintaining RCS pressure and temperature limits has been broken into two sets of action statements, for conditions above 200°F and for conditions below 200'F. This is a logical split since the CTS action is to reduce the temperature below 200°F which would be meaningless ifthe pressure and temperature limits were violated and the temperature were already below 200 0 F. Since the new action statement is addressed below, this change is considered administrative. This change also includes changing to the NUREG-1431 use of MODES to define plant conditions rather than use of prose descriptions. Since use of MODES does not change any plant operations, this is also an administrative change.

M 62 CTS 3.1 .B.1.b. CTS do not specify a time frame for evaluating the integrity of the RCS following an out-of-limit condition. For consistency with NUREG-1431, a limit of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is included.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable to accomplish the evaluation. The evaluation for a mild violation is possible within this time, but more severe violations may require special, event specific stress analyses or inspections. The evaluation will determine whether the RCS structural integrity remains acceptable for continued operation. Since this is a new limit it is more restrictive on plant operation. This more restrictive change is included to make the PI ITS complete.

Prairie Island Units I and 2 27 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4-M 63 CTS 3.1..B.1.b. For consistency with NUREG-1431, new action statements have been included for out-of-limit conditions when the RCS is below 2000F. PI CTS do not address this condition. Since these are new requirements they are considered more restrictive on plant operations. This more restrictive change is included to make the PI ITS complete.

M 64 A new surveillance requirement, SR 3.4.3.1, has been included for consistency with NUREG-1431 which requires verification that RCS pressure, temperature and heatup and cooldown rates are within the specified limits. Plant operators currently monitor these variables for compliance with Specification 3.1 .B.1 .b although it is not explicitly written as a TS SR. Since these are new TS requirements they are considered more restrictive on plant operations. This more restrictive change is included to make the P1 ITS complete.

65 Not used.

Prairie Island Units 1 and 2 28 12/11/00

Part D Package 3.4 NSHID Change Category Number Discussion of Change 3.4 R 66 CTS 3.1..B.2. In conformance with the guidance of NUREG-1431, the pressurizer heatup and cooldown specifications have been relocated from the TS to the PTLR. This change is acceptable since the Bases for Specification 3.4.3 state that the reactor pressure vessel is the most limiting component for brittle fracture; thus the requirements for the pressurizer have not been included in the ITS. The shutdown requirements associated with pressurizer heatup and cooldown limitations have been relocated to the TRM.

Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

The Pressurizer Pressure/Temperature Limits are not installed instrumentation, nor are they used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a Design Basis Accident (DBA). Therefore, the Pressurizer Pressure/Temperature Limits do not meet criterion 1.

Criterion 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.

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Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 R 66 (continued)

The Pressurizer Pressure/Temperature Limits are not process variables, design features, or operating restrictions that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. P1 does not take any credit for the Pressurizer Pressure/Temperature Limits in any accident analysis. Therefore, the Pressurizer Pressure/Temperature Limits do not meet the criterion 2.

Criterion 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.

The Pressurizer Pressure/Temperature Limits are not structures, systems or components that are part of the primary success path which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Therefore, the Pressurizer Pressure/Temperature Limits do not meet criterion 3.

Criterion 4: A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The Pressurizer Pressure/Temperature Limits are a structure, system or component. Therefore, the Pressurizer Pressure/Temperature Limits do not meet criterion 4.

Prairie Island Units I and 2 30 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 R 66 (continued)

Therefore, since the screening criteria have not been satisfied, the Pressurizer Pressure/Temperature Limit LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications. The Pressurizer Pressure/Temperature Limits are being relocated to the PTLR and the associated shutdown requirements are being relocated to the TRM.

R 67 CTS 3.1..B.3. In conformance with the guidance of NUREG 1431, the Steam Generator pressure/temperature limits have been relocated to the PTLR. This operating restriction does not present a challenge to the integrity of a fission product barrier and this limit is not required for safe operation. These specifications do not meet the Technical Specification Selection Criteria defined in IOCFR50.36. The shutdown requirements associated with this pressure/temperature limit have been relocated to the TRM.

Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

The Steam Generator Pressure/Temperature Limits are not installed instrumentation, nor are they used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a Design Basis Accident (DBA). Therefore, the Steam Generator Pressure/Temperature Limits do not meet criterion 1.

Prairie Island Units 1 and 2 31 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 R 67 (continued)

Criterion 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.

The Steam Generator Pressure/Temperature Limits are not process variables, design features, or operating restrictions that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. PI does not take any credit for the Steam Generator Pressure/Temperature Limits in any accident analysis. Therefore, the Steam Generator Pressure/Temperature Limits do not meet the criterion 2.

Criterion 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.

The Steam Generator Pressure/Temperature Limits are not structures, systems or components that are part of the primary success path which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, the Steam Generator Pressure/Temperature Limits do not meet criterion 3.

Prairie Island Units 1 and 2 32 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 R 67 (continued)

Criterion 4: A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The Steam Generator Pressure/Temperature Limits are not a structure, system or component. Therefore, the Steam Generator Pressure/Temperature Limits do not meet criterion 4.

Therefore, since the screening criteria have not been satisfied, the Steam Generator Pressure/Temperature Limit LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications. The Steam Generator Pressure/Temperature Limits are being relocated to the PTLR and the associated shutdown requirements are being relocated to the TRM.

Prairie Island Units 1 and 2 33 1/2/02

Part D Package 3.4 NSHID Change Category Number Discussion of Change 3.4-L 68 CTS 3.1.C.1. CTS requires two means of leakage detection. One of those means is specified as.the radionuclide detection instrumentation, the other means is not specified. For consistency with the guidance of NUREG-1431, a second means of detecting RCS leakage, containment sump A pump run time, has been specified. This is a more restrictive change which is acceptable since it is consistent with CTS Bases and USAR.

Also in conformance with the guidance of NUREG-1431, action statements and remedial actions, in the event one or both means of leakage detection are inoperable, have been provided. The CTS are unnecessarily restrictive in that inoperability of one means of leakage detection requires shutdown of the plant. There are remedial measures which can be implemented to compensate for an inoperable leakage detection channel as allowed by NUREG 1431. The PI ITS includes provisions for compensatory sampling and surveillance testing if leakage detection channels are inoperable. Implementation of this change will provide increased plant operational flexibility and thus this is a less restrictive change.

69 Not used.

70 Not used.

Prairie Island Units 1 and 2 34 12/11/00

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4-A 71 CTS 3.1.0.1. The CTS requirements described here are the same as ITS LCO 3.0.3 (CTS LCO 3.0.c) which is the appropriate action if no leakage detection instrumentation is operable. Thus this is rewritten to require immediate entry into LCO 3.0.3. The change of applicability of this requirement is addressed in L3.4-68. The time and the implications of the use of the time are the same; therefore this is an administrative change.

M 72 Four new SRs, 3.4.16.1, 3.4.16.2, 3.4.16.3 and 3.4.16.4, have been included to perform Channel Checks, COTs and Calibrations of containment radiation monitors and Calibration of sump pump run time instrumentation. Since these requirements are new to the PI TS they are considered more restrictive to plant operation.

These SRs have been included for consistency with NUREG-1431.

These more restrictive changes are included to make the PI ITS complete.

A 73 CTS 3.1.C.2.a. The CTS specifics of where leakage is not originating from have not been included in ITS LCO 3.4.14. The ITS definition of LEAKAGE includes the required details of what is considered leakage; thus these details are not necessary in the specification. Since this change does not change any plant operations, this is an administrative change.

Prairie Island Units 1 and 2 35 12/11/00

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4-LR 74 CTS 3.1 .C.2.b. The requirement to evaluate the leakage for continued safe operation is a level of detail beyond that contained within NUREG-1431. Therefore this requirement has been relocated to the Bases. Since ITS Bases (under the Bases Control Program in Section 5.5 of the ITS) is licensee controlled, this change is less restrictive.

75 Not used.

76 Not used.

A 77 CTS 3.1.C.2.d. The CTS requirement to remain in COLD SHUTDOWN until the condition is corrected is not included in ITS specification 3.4.14. The rules of ITS use in LCO 3.0.4 require ITS LCOs to be met prior to resumption of power operation. Thus, it is unnecessary to include these requirements in ITS LCO 3.4.14.

Since the change does not involve any changes to plant operations, this is an administrative change.

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Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 A 78 CTS 3.1.C.2.e. CTS requires inservice steam generator tube inspection in accordance with TS 4.12. Since this TS has been relocated to the Steam Generator Program in ITS Section 5.5, no technical changes are associated with this change; therefore, this is an administrative change.

L 79 CTS 3.1 .C.3. A new action statement is included in the ITS which allows 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to isolate the flow path of a PIV with leakage outside the allowed limits and 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore the PIV to within limits. CTS allows one hour to initiate shutdown. These changes are consistent with the guidance of NUREG-1431. These changes are acceptable since the safety function of the PIV is met once the PIV flow path is isolated. Furthermore, the isolation valve is also required to meet the same leakage limits as the PIV, in accordance with the additional new requirement for isolation valve leakage limits. An additional three hours to take action is acceptable since these leakage limits are very conservative and it is very unlikely that these valves will suffer complete failure during this limited time. When the PIV flow path is isolated, this may cause some other system, such as RHR, to be inoperable and require entry into an Action Statement for another Specification which would require completion of required actions within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Thus this change is consistent with other allowed outage times for significant safety features. Since this change allows additional plant operational flexibility, this is a less restrictive change.

Prairie Island Units l and 2 37 12/11/00

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4-80 Not used.

M 81 CTS 3.1.D.2.a. In accordance with the guidance of NUREG-1431, verification that the limits of Figure TS.3.1-3 (ITS Figure 3.4.17-1) are met shall be performed every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Since CTS do not specify a time frequency, this is a more restrictive change. This change is acceptable since it will assure that the limits are met on a timely basis and it does not introduce any unsafe plant operating conditions.

L 82 CTS 3.1.C.3 and Table 4.1-2B, Item 4a. CTS RCS specific activity limits when the RCS temperature is below 500'F and above cold shutdown have not been retained. The purpose of Specification 3.4.17 is to limit SGTR releases to a small fraction of 10CFR100 limits. This change is acceptable because below 500°F the release of radioactivity in the event of a SGTR is unlikely since the saturation pressure of the reactor coolant is below the lift pressure settings of the main stream safety valves. Since this change would not retain CTS requirements it is less restrictive on plant operations. This change is consistent with the guidance of NUREG-1431.

Prairie Island Units I and 2 38 12/11/00

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 A 83 CTS 3.3.A.3 and 3.3.A.4. For clarity and to be consistent with the guidance of NUREG-1431, "MODE 4, MODE 5 when the SG primary system manways and pressurizer manway are closed and secured in position, and MODE 6 when the reactor vessel head is on and the SG primary system manways and pressurizer manway are closed and secured in position." have been included in the description of when this specification is applicable. This is an administrative change since the plant is by definition in MODE 4 when the RCS temperature meets the criteria for the OPPS enable temperature.

M 84 CTS Table 4.1-1C, Item 25. For consistency with NUREG-1431, the functional test of the low temperature overpressure protection system will be performed monthly when the RCS temperature is below the LTOP enable temperature. Since CTS require this test each refueling outage, this is a more restrictive requirement. This change will provide additional assurance that the LTOP system will perform as required and is consistent with the surveillance interval for instrumentation providing similar plant protection.

M 85 Table 4.1-1C, Note 38. CTS do not specify a time frame within which this SR must be in compliance. Since ITS includes a time limit of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , this is a more restrictive change. This change conforms to the guidance of NUREG-1431 and is acceptable because it assures that the plant is maintained in a safe condition.

Prairie Island Units I and 2 39 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 L 86 CTS 4.6.C and Table 4.1-2A, Item 7. In accordance with GL 91 04, the surveillance interval for PORV functional testing and emergency pressurizer heater power supply are increased to 24 months to accommodate planned future extended reactor fuel cycles. Since this testing will occur less frequently, this is a less restrictive change. PORV functional tests were reviewed for a five year period and no problems were identified. Therefore it was concluded that an increased surveillance interval would have a minimal effect on plant safety. The emergency pressurizer heater power supply is currently tested prior to each refueling outage so that if a problem is identified it could be corrected during the ensuing outage. Review of testing experience on the emergency power supply did not identify any problems. Therefore it was concluded that an increased surveillance interval would have a minimal effect on plant safety.

L 87 CTS 4.3. This proposed change would revise the allowable PIV leakage of 1.0 gpm to 0.5 gpm per inch of nominal valve size up to 5 gpm maximum. This change is acceptable since the CTS 1.0 gpm limit imposes an unjustified limitation on larger valves. The restrictive limit, when applied to the larger valve, would require a repair effort when the relative degradation of the valve does not warrant the cost or exposure. A leakage limit based on valve size is more apt to provide meaningful information with respect to the mechanical condition of the valve, and is considered superior.

This change is consistent with the guidance of NUREG-1431.

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P~rt D Package 3.4 P rt D NSHD Change Number Discussion of Change Category 3.4 L 88 CTS Table 4.1-2B, Item 1. The surveillance interval for RCS gross activity determination would be increased to once per week by this change in conformance with the guidance of NUREG-1431. This change is acceptable because fuel failures are most likely to occur during startup and fast power changes and not during steady state power operation during which the majority of sampling is performed. Gross fuel failures will also result in letdown radiation alarms and possibly containment radiation alarms providing additional operator indication.

89 Not used.

90 Not used.

Prairie Island Units I and 2 41 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4-L 91 CTS Table 4.1-2B, Item 4a. The CTS requirement to sample once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when specific activity exceeds 100 1 pICi/gram has not been included. In accordance with ITS 3.4.17 Action B, whenever this limit is not met, the plant is required to be in MODE 3 with Tave

< 500°F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Thus the requirement for sampling in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months serves no useful purpose and is unnecessary. Since this change will require less sampling this change is less restrictive.

This change is consistent with the guidance of NUREG-1431.

92 Not used.

93 Not used.

LR 94 CTS Table 4.1-2B, Items 5 and 6, and Note 2. The purpose of Specification 3.4.17, RCS Specific Activity is to limit the offsite radioactivity dose consequences from a SGTR to a small fraction of 10CFR100. This change will relocate Items 5 and 6, RCS Radiochemistry and RCS Tritium activity, from CTS Table TS.4.1 2B to the TRM since these items are not significant in limiting SGTR offsite dose and therefore should not be in TS. This is less restrictive since the TRM is under licensee control. However this change is acceptable since the TRM is under the controls of 10CFR50.59. This change conforms the PI ITS to the guidance of NUREG-1431.

Prairie Island Units 1 and 2 42 12/11/00

PartD Package 3.4 NSHD Change Category Number Discussion of Change 3.4 95 Not used.

LR 96 CTS Table 4.1-2B, Item 8 and Note 4. RCS boron concentration measurement at power was not included in the ITS since RCS Chemistry does not meet the NRC Policy Statement for TS Screening Criteria and is not required to be addressed within the TS. This requirement is relocated to the TRM. While this is a less restrictive change since the TRM is under licensee control, this change is acceptable because the TRM is under the controls of 10CFR50.59. This change is consistent with the guidance of NUREG-1431.

LR 97 CTS 4.3. For consistency with NUREG-1431, the list of valves and the test methodology have been relocated to the Bases. This detailed information is not required in the TS to run the plant in a safe manner. Since ITS Bases (under the Bases Control Program in Section 5.5 of the ITS) is licensee controlled, relocation of CTS requirements to the Bases is a less restrictive change.

LR 98 CTS 4.6.C. The methodology for performing this surveillance has been relocated to the Bases. This detailed information is not required in the TS to run the plant in a safe manner. Since ITS Bases (under the Bases Control Program in Section 5.5 of the ITS) is licensee controlled, relocation of CTS requirements to the Bases is a less restrictive change.

Prairie Island Units 1 and 2 43 12/11/00

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 A 99 CTS 4.3. CTS require this SR to be performed prior to resuming power after each refueling. Since the NUREG-1 431 format requires SR Frequency statement in months, this is revised to every 24 months. Since this change does not materially change the testing of these valves, this is an administrative change.

A 100 CTS 3.10.J. CTS 3.10.J states, "Compliance with c. is demonstrated by verifying that the parameter is within its limit after each refueling cycle." The parameter being referenced is the Reactor Coolant Flow which is referred to in the ITS as the RCS total flow rate. In addition, "c" states, "Reactor Coolant Flow >the value specified in the CORE OPERATING LIMITS REPORT." These statements have been editorially changed to be more consistent with the wording of the STS SR 3.4.1.4 as follows, "Verify RCS total flow rate is within the limit specified in COLR." The CTS requires that the Frequency for this SR as "after each refueling". This Frequency has been changed to "24 months" in the ITS which is still consistent with the CTS.

This change is considered to be an Administrative change since only editorial changes were made and no parameter, technical or operational changes made.

Prairie Island Units I and 2 44 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 LR 101 CTS 3.10.J. The CTS requires that an engineering evaluation be performed to determine the effects of the out-of-limit condition on the structural integrity of the RCS. This information is being relocated to the ITS Bases 3.4.3, Action A.1 and A.2 section. Therefore, besides restoring the RCS pressure and temperature to within limits, a determination will be made ifthe RCS is acceptable for continued operation. This is accomplished through an evaluation. The evaluation must verify the RCPB integrity remains acceptable and must be completed before continued operation. Several methods can be used, including an engineering evaluation, comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components. Relocating this CTS information to the Bases retains the CTS and provides additional guidance for performing this determination as required in Required Action A.2.

A 102 CTS 3.1.A.1.c.(1). The CTS states "...whenever the reactor coolant system average temperature is below 350 °F, except during REFUELING .. ." This has been changed to state in Mode 4 which is consistent with the Mode definition changes made in ITS 1.0. In MODE 4 the reactor temperature is 350 OF

> Tavg > 200 OF. This is considered to be an Administrative change since the temperature limits stated in the CTS are the same as Mode 4. This change is consistent with NUREG-1431.

Prairie Island Units 1 and 2 45 1/2/02

Part D Package 3.4 Prt D..

NSHD Change Category Number Discussion of Change 3.4 A 103 CTS 3.1.A.1.c.(1). The CTS states "... whenever the reactor coolant system average temperature is below 350 OF, except during REFUELING .. ." This has been changed to state in Mode 5 with RCS loops filled which is consistent with the Mode definition changes made in ITS 1.0. In MODE 5 the reactor temperature is < 200 OF. This is considered to be an Administrative change since the temperature limits stated in the CTS are the same as Mode 5. This change is consistent with NUREG-1431.

A 104 CTS 3.1 .A.l.c (2). Both the CTS and ITS require two loops (methods) of decay heat removal be OPERABLE when the reactor is in MODE 4. The CTS states that with only one OPERABLE method of removing decay heat, initiate prompt action to restore two OPERABLE methods of removing decay heat. This requirement has been editorially changed to be consistent with NUREG 1431 by stating, one required loop inoperable, immediately initiate action to restore a second loop OPERABLE. The actions are the same in both the CTS and ITS in that when one loop (method) of decay heat removal is inoperable, then the second loop (method) must immediately (promptly) be restored to OPERABLE. Since the Required Actions and associated Completion Times are the same and no technical changes or operating practices were changed, this is considered to be an Administrative change.

Prairie Island Units 1 and 2 46 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4-A 105 CTS 3.1.A.1 .c (2). The CTS states that if one method of decay heat removal is inoperable and if the only remaining operable method is an RHR loop, be in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The ITS editorially changes this requirement such that when one required loop is inoperable, Required Action A.2 requires that the unit be placed in MODE 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months only if the required RHR loop is OPERABLE. The Actions and Completion Times in both the CTS and ITS are the same, no technical or operational changes have been made to this requirement, therefore this is an Administrative change.

Prairie Island Units 1 and 2 47 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 A 106 CTS 3.1.A.1.c (3). The CTS requires that with no methods of removing decay heat, suspend all operations involving a reduction in boron concentration of the RCS and initiate prompt action to restore one loop OPERABLE method of removing decay heat. This requirement has been editorially changed to be consistent with NUREG 1431 by stating that two required loops inoperable or required loop not in operation, suspend operations that would cause introduction into RCS coolant with boron concentration less than required to meet SDM of LCO 3.1.1 and immediately initiate action to restore one loop OPERABLE and in operation. Both the CTS and the ITS require the same actions and associated Completion Times.

Both the CTS and ITS require two loops (methods) of decay heat removal. This specific action is for when both loops (methods) of decay heat removal are inoperable. In both cases, the primary action is to suspend actions that might reduce RCS coolant boron concentration. The ITS is more specific by referring to SDM as stated in LCO 3.1.1 whereas the CTS does not specifically refer to the TS LCO but the required actions and intent are the same as the ITS. The Completion Times in both the CTS and ITS are the same. In addition, both the CTS and ITS require that a loop (method) of decay heat removal be restored to OPERABLE status. The CTS requires prompt action whereas the ITS uses the phrase immediate action. Both terms mean the same at Pl. Based on the above, this is considered to be an Administrative change since no technical or operational changes are being made to the CTS.

Prairie Island Units 1 and 2 48 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4-A 107 CTS 3.1 .A.1 .c.(1). The CTS requires that two methods for decay heat removal be OPERABLE with one in operation. The CTS further states that acceptable methods for removing decay heat are at least one reactor coolant pump (RCP) and its associated steam generator (SG), or residual heat removal (RHR) loop including a pump and its associated heat exchanger. As stated in DOC LR3.4-24, the RCS loop consists of the RCP and associated SG. In addition, the RHR loop consists of the RHR pump and associated heat exchanger.

Therefore, the ITS uses the RHR loop and RCP loop instead of all the associated components. Even though the components are not specifically mentioned in the LCO, they are included in the definitions of loop. The CTS also states that two acceptable methods of decay heat removal are needed and provide those acceptable methods. The intent is that any of the two methods identified are acceptable. This has been changed in the ITS to state that two loops consisting of any combination of the RHR loops or RCS loops are the acceptable methods. This is only an editorial change in that the both the CTS and ITS allow the use of the same decay heat removal system or a combination of the two to meet the LCO requirements. Therefore, this is considered to be an Administrative change.

Prairie Island Units 1 and 2 49 S1/2/02

Part D Package 3.4 PartD Package 3.4 NSHD Change Category Number Discussion of Change 3.4-108 Not used.

Prairie Island Units I and 2 50 1/2/02

m

A J Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 L 109 CTS 3.1 .A.2.b(1). CTS 3.1 .A.2.b(1) requires two RCS pressurizer safety valves (PSVs) to be operable when the RCS temperature is > 350 0F. This change will require two RCS PSVs to be operable whenever both RCS cold leg temperatures are greater than the OPPS enable temperature specified in the PTLR. This change is consistent with the guidance of NUREG-1431 and will provide additional overpressure protection when the RCS is between 350OF and the OPPS enable temperature. Per the ITS, in the event that an inoperable PSV cannot be restored to OPERABLE status within 15 minutes or if both PSVs are inoperable, the unit must be placed in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 4 with any RCS cold leg temperatures < the OPPS enable temperature specified in the PTLR in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The CTS requires the plant to be below 350 degrees F within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Increasing the Completion Time from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is considered to be a less restrictive change. The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. With any RCS cold leg temperatures at or below the OPPS enable temperature specified in the PTLR, overpressure protection is provided by the LTOP function. This Completion Time change is consistent with the guidance of NUREG-1431 as modified by approved traveler TSTF-352, Revision 1. These changes are acceptable because they will not cause any unsafe plant operating or testing conditions.

Prairie Island Units 1 and 2 51 1/2/02

Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 A 110 CTS 3.1.A.2.c (1) (b) 4. CTS and ITS LCO 3.4.11, Condition C both require that with one block valve inoperable, place its associated PORV in manual control within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and restore the block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The ITS provides additional clarification by adding a Note stating that the Required Actions C.1 and C.2 do not apply if sole reason for the block valves being declared inoperable is as a result of power being removed to comply with other Required Actions. In this event, the Required Actions for inoperable PORV(s) are adequate to address the conditions. Even though not specifically stated, the intent and required operational practices stated in the ITS are the same as the CTS. Since there are no operational or technical changes associated, this is considered to be an Administrative change. Making an editorial change to the CTS by specifically adding the Note for clarification also supports this change as Administrative.

A 111 CTS 3.1.A.2.c (1) (b) 5. CTS and ITS LCO 3.4.11, Condition F both require that with both block valves inoperable, restore one block valve to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The ITS provides additional clarification by adding a Note stating that the Required Action F.1 does not apply if sole reason for the block valves being declared inoperable is as a result of power being removed to comply with other Required Actions. In this event, the Required Actions for inoperable PORV(s) are adequate to address the conditions. Even though not specifically stated, the intent and required operational practices stated in the ITS are the same as the CTS. Since there are no operational or technical changes associated, this is considered to be an Administrative change. Making an editorial change to the CTS by specifically adding the Note for clarification also supports this change as Administrative.

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Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 A 112 CTS Table 4.1-2A Item 9. CTS Item 9 requires that the Primary System Leakage be evaluated daily. A Note has been added stating that this evaluation is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation. The RCS water inventory balance must be met with the reactor at steady state operating condition (stable temperature, power level, equilibrium xenon, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows).

Therefore, this SR Note does not require the performance of this SR until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established. Performance of this SR at steady state conditions is standard PI operating practice and consistent with the intent of the CTS. Since this SR Note only provides clarification for CTS requirements, this change is considered to be an Administrative change.

A 113 CTS Table 4.1-2A, Item 6. CTS requires that the Pressurizer PORV Block Valves be functionally tested every 92 days. The ITS adds a note stating that this test only be required to be performed in Modes I and 2. This Note allows entry into Mode and operation in Mode 3 prior to performing the SR. This allows the test to be performed under operating temperatures and pressures. This is considered to be an Administrative change since this is consistent with current operating practices and the intent of the CTS. In addition, no technical nor operational changes are made as a result of this Note.

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Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4-A 114 CTS Table 4.1-2A, Item 7. CTS requires that the PORVs be functionally tested every 18 months. The ITS adds a note stating that this test only be required to be performed in Modes 1 and 2.

This Note allows entry into Mode and operation in Mode 3 prior to performing the SR. This allows the test to be performed under operating temperatures and pressures. This is considered to be an Administrative change since this is consistent with current operating practices and the intent of the CTS. In addition, no technical nor operational changes are made as a result of this Note.

115 Not used.

116 Not used.

M 117 CTS Table 4.1-2.B, Note 1. Note 1 states that a sample is to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer. The ITS Note states that the SR is not required to be performed until 31 days after a minimum of 2 EFPD and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The CTS does not have specific time limit of 31 days to perform the SR, therefore making this a more restrictive change. The 31 days is acceptable since it further ensures that the radioactive materials are at equilibrium so the analysis for E is representative and not skewed by a crud burst or other similar abnormal event.

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Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 L 118 CTS 3.1.A.1.b(3)(a). The CTS requires that with both reactor coolant pumps inoperable or not in operation, immediately "de energize all control rod drive mechanisms, and ... ." ISTS LCO 3.4.5, Required Action D.1 requires that iftwo RCS loops are inoperable or required RCS loops are not in operation, immediately place the control rod drive system in a condition incapable of rod withdrawal. This is considered to be a Less Restrictive change since the CTS only allows for de energization of the control rod drive system, whereas, the ISTS allows for various methods of making the control rod drive system incapable of rod withdrawal. The overall intent, assuring that control rods can not be withdrawn and thereby increase any potential heat input to the reactor coolant is maintained. Since the revised Actions still assure rod withdrawal is precluded, details of specifically stating de energization of the control rod drive system is not necessary nor required to provide adequate protection of the public health and safety. The requirement that the control rods are inserted and are not capable of being withdrawn is also maintained.

This change allows alternate options to preclude rod withdrawal. These options are necessary to allow testing. This Less Restrictive change provides several options to assure that the control rods are not capable to withdraw. This change is consistent with the guidance provided by NUREG-1431 and TSTF-87, Rev. 2.

119 Not used.

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Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 A 120 CTS 3.1.A.1.c (2). Both the CTS and ITS require two loops (methods) of decay heat removal be OPERABLE when the reactor is in MODE 5. The CTS states that with only one OPERABLE method of removing decay heat, initiate prompt action to restore two OPERABLE methods of removing decay heat. Although not specifically stated, the methods for decay heat removal can be a combination of two RHR loops or a RHR loop and a SG. This is also true in the ITS. The CTS requirement has been editorially changed to be consistent with NUREG 1431 by stating, one required loop inoperable and one RHR loop OPERABLE, immediately initiate action to restore a second loop to OPERABLE status. The actions are the same in both the CTS and ITS in that when one loop (method) of decay heat removal is inoperable, then the second loop (method) must immediately (promptly) be restored to OPERABLE. Another option of restoring adequate decay heat removal is ensuring that the SGs secondary side water level is adequate so they are capable of decay heat removal. The only difference between the CTS and ITS is that the ITS specifically identifies the methods for ensuring decay heat removal, whereas the CTS is not as specific even though it has the same intent.

Since the Required Actions and associated Completion Times are the same and no technical changes or operating practices were changed, this is considered to be an Administrative change.

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Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 A 121 CTS 3.1.A.l.c (2). Both the CTS and ITS require two loops (methods) of decay heat removal be OPERABLE when the reactor is in MODE 5. The CTS states that with only one OPERABLE method of removing decay heat, initiate prompt action to restore two OPERABLE methods of removing decay heat. Although not specifically stated, the methods for decay heat removal can be a combination of two RHR loops or a RHR loop and a SG. This is also true in the ITS. The CTS requirement has been editorially changed to be consistent with NUREG 1431 by stating, one or more SGs not capable of decay heat removal and one RHR loop OPERABLE, immediately initiate action to restore a second loop to OPERABLE status or initiate action to restore the required SG capable to remove decay heat. The actions are the same in both the CTS and ITS in that when one loop (method) of decay heat removal is inoperable, than the second loop (method) must immediately (promptly) be restored to OPERABLE. Another option of restoring adequate decay heat removal is ensuring that the SGs secondary side water level is adequate so it is capable of decay heat removal. The only difference between the CTS and ITS is that the ITS specifically identifies the methods for ensuring decay heat removal, whereas the CTS is not as specific even though it has the same intent. Since the Required Actions and associated Completion Times are the same and no technical changes or operating practices were changed, this is considered to be an Administrative change.

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Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 A 122 CTS 3.1.A.1.c (3). The CTS requires that with no methods of removing decay heat, suspend all operations involving a reduction in boron concentration of the RCS and initiate prompt action to restore one loop OPERABLE method of removing decay heat. This requirement has been editorially changed to be consistent with NUREG 1431 by stating that two required loops inoperable or required loop not in operation, suspend operations that would cause introduction into RCS coolant with boron concentration less than required to meet SDM of LCO 3.1.1 and immediately initiate action to restore one loop OPERABLE and in operation. Both the CTS and the ITS require the same actions and associated Completion Times.

Both the CTS and ITS require two loops (methods) of decay heat removal. This specific action is for when both loops (methods) of decay heat removal are inoperable. In both cases, the primary action is to suspend actions that might reduce RCS coolant boron concentration. The ITS is more specific by referring to SDM as stated in LCO 3.1.1 whereas the CTS does not specifically refer to the TS LCO but the required actions and intent are the same as the ITS. The Completion Times in both the CTS and ITS are the same. In addition, both the CTS and ITS require that a loop (method) of decay heat removal be restored to OPERABLE status. The CTS requires prompt action whereas the ITS uses the phrase immediate action. Both terms mean the same at Pl. Based on the above, this is considered to be an Administrative change since no technical or operational changes are being made to the CTS.

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Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 M 123 CTS 3.1 .A.2.c (1) (b) 5. The CTS requires that with both block valves inoperable, within one hour either restore the block valve to OPERABLE or place the PORVs in manual control.

The ITS requires that the block valves be restored to OPERABLE and does not provide an alternate option of placing the PORVs in manual. This option has been deleted from the CTS. Although in the CTS, due to the short Completion Time of placing the PORVs in manual in addition to restoring the block valve to OPERABLE status within the next hour, placing the PORVs in manual is not a commonly utilized action. Since the ITS does not include this option, deleting additional flexibility is a More Restrictive change.

Deleting the action of placing the PORVs in manual control when both block valves are inoperable, does not reduce the safety or operation of the plant since the ultimate action is to either restore the block valve to OPERABLE status within the next hour (2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months total) or initiate a reactor shutdown. In addition, if a block valve cannot be restored to OPERABLE within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months the plant will be required to be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 4 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . When the plant is shutting down, the PORVs will be needed for Temperature Over Pressure Protection. Therefore, the PORVs should not be placed in manual. The CTS Completion Time of a total of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to restore an inoperable block valve, when both are inoperable, is consistent with NUREG-1431.

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Part D Package 3.4 NSHD Change Category Number Discussion of Change 3.4 A 124 CTS Table 4.1-2.B Item 3. The CTS requires that the RCS Radiochemistry and E determination test be preformed once per 6 months when the reactor is at power. This is also annotated with a footnote that states, "Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer." The ITS has editorially revised the CTS SR and associated footnote to state, "Determine E from a sample taken in MODE 1 after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ." The only changes made are that the ITS combines the CTS SR and footnote, and the CTS refers to POWER OPERATIONS vs. MODE 1 in the ITS.

The CTS definition of POWER OPERATIONS is the same as MODE 1. The CTS definition of POWER OPERATIONS was revised to be consistent with MODES as documented in ITS markup of Section 1.0. Since this change is purely editorial and no technical or operational changes were made, this change is considered to be an Administrative change.

A 125 CTS Table 4.1-2.B, Item 2. The CTS requires the RCS Isotopic Analysis for DOSE EQUIVALENT 1-131 Concentration test be performed once every 14 days when the reactor is at power. The ITS requires the same test to be performed at the same Frequency. However, the ITS in conformance with the ISTS contains a note that states that the SR is only required to be performed in MODE 1. This change is considered to be an Administrative change since the only change made is editorial and in presentation. The CTS has a requirement of when the reactor is at power whereas the ITS contains the same requirement but it is in a note. The CTS definition of power operation is essentially equivalent to MODE 1 in the ITS. In addition, this change does not make any technical or operational changes and is therefore considered an Administrative change.

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PairtD Package 3.4 Part D NSHD Change Category Number Discussion of Change 3.4 L 126 CTS 3.1.A.1.c.(1). The CTS requires two methods for removing decay heat with one of the methods in operation.

The CTS further states that acceptable methods for removing decay heat are at least one reactor coolant pump (RCP) and its associated steam generator (SG) or a residual heat removal loop including its associated heat exchanger. DOCs A3.4-107 and LR3.4-24 document what constitutes a RCS loop or RHR loop. No other discussion on this will be discussed in this DOC. The CTS states that a RCP and its associated SG are an acceptable method whereas the ITS only requries the SG to be OPERABLE if the RHR is not used as the second method of decay heat removal. Ifthe SG is the second method of decay heat removal, its associated RCP is not needed to be OPERABLE in order for the SG to perform its decay heat removal function. Therefore, the ITS does not require the RCP to be OPERABLE for this LCO making this a less restrictive change. This change is acceptable since the one RHR loop that is OPERABLE and in operation provides forced circulation to perform the safety functions of the reactor coolant under Mode 5, loops filled condition. An additional RHR loop is required to be OPERABLE to provide redundancy. However, if the standby RHR loop is not OPERABLE, an acceptable alternate method is a SG. The SG could be used to remove decay heat via natural circulation. This change is consistent with NUREG-1431.

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Part 0 Package 3.4 P rt D NSHD Change Category Number Discussion of Change 3.4 A 127 CTS3.3.A.3 (b). The CTS requires that only one SI pump may be capable of injecting into the RCS whenever the RCS temperature is less than the OPPS enable temperature specified in the PTLR except that both Sl pumps may be run for up to one hour while conducting the integrated S1 test**

when either of the following conditions is met: (b) the reactor vessel head is removed. This statement is not being incorporated into the ITS since it is no longer applicable. In the conversion process from the CTS to the ITS, the MODES and Conditions of Applicability have been changed. The ITS provides two LCOs for the LTOP function. LC 3.4.12 is with the MODE of Applicability in MODE 4 when any RCS cold leg temperture is greater than the SI pump disable temperature specified in the PTLR. The second LCO is 3.4.13 with the MODE of Applicability in MODE 4 when any RCS cold leg temperture is less than or equat to the SI pump disable temperature specified in the PTLR, MODE 5 when the SG primary system manways and pressurizer manway are closed and secured in position, and MODE 6 when the reactor vessel head is on and the SG primary system manways and pressurizer manway are closed and secured in position.

Based on the redefinition of the MODES and conditions of Applicability in the ITS, the CTS requirement is not applicable.

The CTS requirement of requiring the reactor vessel head to be removed would take the reactor out of the MODE of Applicability and therefore, ITS 3.4.12 and 3.4.13 would not be applicable nor would their subject LCO notes. Based on the above, the CTS requirement for reactor head removal is not incorporated into the ITS.

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ML020570030

Text

3.4.7 APPLICABILITY

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