Text

Attachment 1 to Supplement Dated 02/15/2002 to License Amendment Request Dated 12/11/2000 Conversion to Improved Technical Specification, NRC RAIs Section 3.4, Reactor Coolant System & NMC Responses
	ML021900498
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
	Download: ML021900498 (29)
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Prairie Island Nuclear Generating Plant 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 4[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 3.10.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." PI 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 540 0F. 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 P1 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.l.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 Ž_ 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-1 14, Rev. 0 was approved by the NRC on Dec 31, 1996. TSTF-1 14 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 if the 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.1.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 1 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 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 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 Ž_ 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 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 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 2 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.

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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.

14

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 A3.4-1 10 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 P1 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 < SI 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

SI pump disable temperature. STS 3.4.12 has the LTOP 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 PI. 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°F. This change is described as A3.4-08. CTS 3.1.A.1.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

ML021900498

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