ML023610356

License Basis Document Change Request 3-13-02 Elimination of 'N-1' Loop Operation from Technical Specifications

ML023610356
Person / Time
Site:	Millstone
Issue date:	12/11/2002
From:	Price J Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Security and Incident Response
References
B18814
Download: ML023610356 (212)

Text

Dominion Nuclear Connecticut, Inc. 1~'Domine~mion-u Millstone Power Station Rope Ferry Road Waterford, CT 06385 DEC I1 2002 Docket No. 50-423 B18814 RE: 10 CFR 50.90 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Power Station, Unit No. 3 License Basis Document Change Request 3-13-02 Elimination of 'N-I' Loop Operation from Technical Specifications Introduction On January 12, 1998,(1) Northeast Nuclear Energy Company (NNECO) notified the Nuclear Regulatory Commission (NRC) of their intent to modify the licensing bases of Millstone Unit No. 3 to remove the option of 'N-I' loop operation. This decision was based on Millstone Unit No. 3 never needing or utilizing 'N-I' loop operation, and as a result, not completing all of the necessary procedure modifications to support 'N-i' loop operation. As part of this submittal, NNECO committed to imposing interim administrative controls to remove the option of 'N-i' loop operation, and NNECO committed to remove all references to 'N-I' loop operation from the facility Technical Specifications and Final Safety Analysis Report. As part of the implementation of this commitment, procedure controls were implemented that prohibited 'N-I' loop, or three loop operation.

Therefore, pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC),

hereby proposes to amend Operating License NPF-49 by incorporating the attached proposed changes into the Millstone Unit No. 3 Technical Specifications. The purpose of the proposed changes is to eliminate technical specification requirements associated with 'N-I' loop operation from the Millstone Unit No. 3 Technical Specifications.

(1) John P McElwain to U.S NRC, "Millstone Nuclear Power Station, Unit No. 3, Modification of Licensing Bases for N-1 Loop Operation," dated January 12, 1998 Auo (

U.S. Nuclear Regulatory Commission B18814/Page 2 Specifically, the Technical Specification changes will:

Delete Technical Specification requirements associated with 'N-i' loop, or three loop plant operation. Delete references to four loop operation whose only purpose is to distinguish from three loop operation;

• Add requirements to maintain RCS loop stop valves open at all times during MODES 1 through 4; and Require that each RCS loop be maintained OPERABLE at all times during MODES 1 and 2.

The Bases and Index for these Technical Specifications will also be modified to reflect these changes as applicable. provides a discussion of the proposed changes and the Safety Summary. provides the Significant Hazards Consideration. Attachment 3 provides the marked-up version of the appropriate pages of the current Technical Specifications. provides the retyped pages of the Technical Specifications.

Environmental Considerations DNC has evaluated the proposed changes against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.22. DNC has determined that the proposed changes meet the criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and as such, has determined that no irreversible consequences exist in accordance with 10 CFR 50.92(b). This determination is based on the fact that the changes are being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to use of a facility component located within the restricted area, as defined by 10 CFR 20, or that changes a surveillance requirement, and that the amendment request meets the following specific criteria.

(i) The proposed changes involve no Significant Hazards Consideration.

As demonstrated in Attachment 2, the proposed changes do not involve a Significant Hazards Consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released off site.

The proposed changes provide for the removal of the option for Millstone Unit No. 3 'N-I' loop operation. The proposed changes will not result in an increase in power level, will not increase the production of radioactive waste and byproducts, and will not alter the flowpath or method of disposal of radioactive waste or byproducts. Therefore, the proposed changes will not increase the type and amounts of effluents that may be released off site.

U.S. Nuclear Regulatory Commission B18814/Page 3 (iii) There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed changes will not result in changes in the configuration of the facility. There will be no change in the level of controls or methodology used for processing radioactive effluents or the handling of solid radioactive waste. There will be no change to the normal radiation levels within the plant. Therefore, there will be no increase in individual or cumulative occupational radiation exposure resulting from the proposed change.

Conclusions The proposed changes were evaluated and we have concluded that they are safe. The proposed changes do not involve an adverse impact on public health and safety (see the Safety Summary provided in Attachment 1) and do not involve a Significant Hazards Consideration pursuant to the provisions of 10 CFR 50.92 (see the Significant Hazards Consideration provided in Attachment 2).

Site Operations Review Committee and Management Safety Review Committee The Site Operations Review Committee and Management Safety Review Committee have reviewed and concurred with the determinations.

Schedule We request issuance of this amendment for Millstone Unit No. 3 prior to December 15, 2003, with the amendment to be implemented within 90 days of issuance.

In a January 9, 2002,(2) letter the NRC imposed a condition upon Millstone Unit No. 3 which restricted 'N-I' loop, or three loop operation. As noted in the January 9, 2002, letter, this condition will expire upon NRC issuance of this proposed license amendment.

State Notification In accordance with 10 CFR 50.91(b), a copy of this License Amendment Request is being provided to the State of Connecticut.

(2) Victor Nerses, Sr., U.S. NRC, to J. A. Price, "Millstone Nuclear Power Station, Unit No. 3 - Issuance of Amendment Re: Reactor Coolant System - Isolated Loop Startup," TAC No. MB1785, dated January 9, 2002.

U.S. Nuclear Regulatory Commission B18814/Page 4 There are no regulatory commitments contained within this letter.

If you should have any questions on the above, please contact Mr. Ravi Joshi at (860) 440-2080.

Very truly yours, DOMINION NUCLEAR CONNECTICUT, INC.

Site President - Millstone Sworn to and subscribed before me this *"'day of "),*CerYlb * , 2002 Notary Public My Commission expires 7u Y)C 3Oi2-OO,"

ELENA LLOCKET Notary Public Attachments (4) State of Connecticut County of New London cc: H. J. Miller, Region I Administrator V. Nerses, NRC Project Manager, Millstone Unit No. 3 NRC Senior Resident Inspector, Millstone Unit No. 3 Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Docket No. 50-423 B1 8814 Attachment 1 Millstone Power Station, Unit No. 3 License Basis Document Change Request 3-13-02 Elimination of 'N-I' Loop Operation from Technical Specifications Discussion of Proposed Changes and Safety Summary

U.S. Nuclear Regulatory Commission B18814/Attachment 1/Page 1 License Basis Document Change Request 3-13-02 Elimination of 'N-I' Loop Operation from Technical Specifications Discussion of Proposed Chanqes and Safety Summary Introduction Millstone Unit No. 3 is designed and constructed with loop stop valves such that each of the four (4) reactor coolant loops could be isolated from the remaining three loops. As early as 1984,(3) Northeast Utilities stated its intent to pursue Nuclear Regulatory Commission (NRC) approval for operation of Millstone Unit No. 3 with one loop isolated and out-of-service ('N-I' loop operation). 'N-I' loop operation allows continued plant operation in the event of an equipment failure (i.e. reactor coolant pump failure).

On January 21, 1987,(4) the NRC issued a safety evaluation report for Millstone Unit No.

3 documenting the acceptability of proposed protection system modifications which were needed to support 'N-I' loop operation. On November 16, 1987,(') the NRC issued their final safety evaluation documenting the acceptability of plant operation with an isolated loop.

On January 12, 1998, Northeast Nuclear Energy Company (NNECO) notified the NRC of their intent to modify the licensing bases of Millstone Unit No. 3 to remove the option of 'N-I' loop operation. This decision was based on Millstone Unit No. 3 never needing or utilizing 'N-I' loop operation, and as a result, not completing all of the necessary procedure modifications to support 'N-I' loop operation. As part of this submittal, NNECO committed to imposing interim administrative controls to remove the option of

'N-I' loop operation, and NNECO committed to remove all references to 'N-I' loop operation from the facility Technical Specifications (TSs) and Final Safety Analysis Report (FSAR). As part of the implementation of this commitment, procedure controls were implemented that prohibited 'N-I' loop, or three loop operation.

Therefore, pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC),

hereby proposes to amend Operating License NPF-49 by incorporating the attached proposed changes into the Millstone Unit No. 3 TSs. The purpose of the proposed changes is to eliminate TS requirements associated with 'N-I' loop operation from the Millstone Unit No. 3 TSs. The Bases and Index for these TSs will also be modified to reflect these changes as applicable.

(3) W. G. Counsil to B. J. Youngblood, U.S. NRC, "Millstone Nuclear Power Station, Unit No. 3, N-1 Loop Operation," dated April 9, 1984.

(4) Elizabeth L. Doolittle, U.S. NRC to E. J. Mroczka, "Safety Evaluation Report for Millstone Nuclear Power Station Unit No. 3, N-1 Loop Operation (License Condition 2.C(4))," dated January 21, 1987.

(5) Robert L. Ferguson, U.S. NRC to Edward J. Mroczka, "Millstone Nuclear Power Station, UnitNo. 3 Three (N-i) Loop Operation," TAC No. 60387, dated November 16, 1987.

U.S. Nuclear Regulatory Commission B18814/Attachment 1/Page 2 Technical Specification Changes Elimination of 'N-I' Loop Operation Requirements The proposed changes will provide for the elimination of TS requirements associated with 'N-I' loop operation. Each of the changes identified in the following table provides for the consistent elimination of TS information which addresses 'N-I' loop operation.

Additionally, since the capability to operate the plant in the 'N-I' loop, or three loop operation is being removed, except for the mitigation of an emergency or abnormal event, specific references to four (4) loop operation, which are only used to distinguish from 'N-I' loop operation, are also being eliminated. The table below identifies the TS sections affected by these changes. Changes which are more complex than the simple elimination of 'N-I' Loop requirements (TSs 3.4.1.1 and 3.4.1.5) are not identified in this table. Detailed discussions for the proposed changes to these specifications are addressed separately.

Technical . Affected Item(s) Affected Page(s)

Specifiicatio n Number_________ __

2.1.1

• Safety Limit 2.1.1 2-1,2-2, 2-3, and B 2-1 0 Figure 2.1-1 0 Figure 2.1-2 (deleted)

• Bases 2.2.1 0 Table 2.2 Functional Units 2.a, 7, 8, 18.c, 2-5, 2-6, 2-7, 2-8, 2-9, and 21 B 2-4, B 2-5, and B 2-8 a Table 2.2 Note 1

• Bases 3.1.1.1.1 LCO 3.1.1.1.1 3/4 1-1 3.1.1.1.2 0 LCO 3.1.1.1.2 3/4 1-3, 1-4, and 1-5

"* Figure 3.1-1

"* Figure 3.1-2 (deleted) 3.1.3.1 Actions b.3.d) and b.3.e) 3/4 1-21 3.1.3.2 Actions a.2, a.3, b.2, and b.3 3/4 1-23 3.2.1.1 Title 3/4 2-1 3.2.1.2 Entire Specification (deleted) 3/4 2-3 and 2-4 3.2.2.1 Title 3/4 2-5 3.2.2.2 0 'Entire Specification (deleted) 3/4 2-12, 2-13, 2-14, 2-15, 2-16, 2-17, and 2-18 3.2.3.1

• Title 3/4 2-19, B 3/4 2-3, 2-4, 1 Bases and 2-5

U.S. Nuclear Regulatory Commission B18814/Attachment 1/Page 3 Technical Affected Itemf(s) Affected Pagde(s)ý

,Specification'

-.Number-3.2.3.2 0 Entire Specification (deleted) 3/4 2-22, 2-23, B 3/4 2 0 Bases 3, 2-4, and 2-5 3.2.5

• Table 3.2-1 3/4 2-28, B 3/4 2-5, and

• Bases B 3/4 2-6 3.3.1 0 Table 3.3 Functional Units 7, 8, 12, 13, 14, 3/4 3-2, 3-3, 3-4, 3-5, and 20 and 3-12

• Table 3.3 Action Statement 2.c

• Table 4.3 Functional Unit 20 3.3.2

• Table 3.3 Functional Unit 5.d 3/4 3-20 and 3/4 3-28

• Table 3.3 Functional Unit 5.d 3.3.5 0 LCO 3.3.5.b.1 3/4 3-82 and B 3/4 3-7 "0 Bases 3.4.9 Bases B 3/4 4-10 3.7.1.1

• LCO 3.7.1.1 3/4 7-1, 3/4 7-2, and B

"* Action 3.7.1.1a 3/47-1

"* Action 3.7.1.1.b

"* Table 3.7-1 (Title)

"* Table 3.7-2 (deleted)

"* Bases 3.10.2.1 Title 3/4 10-2 3.10.2.2 Entire Specification (deleted) 3/4 10-3 Technical Specification 3.4.1.1 Current TS 3.4.1.1 requires that during MODES 1 and 2 (except as provided in Special Test Exception Specification 3.10.4) all reactor coolant loops are in operation or three reactor coolant loops are in operation with associated THERMAL POWER restrictions.

Consistent with the deletion of 'N-I' loop, or three loop operation, this specification will be revised to remove the option for Mode 1 and 2 operation with only three reactor coolant loops OPERABLE.

1. The LCO for TS 3.4.1.1 currently states:

"Either:

a) All reactor coolant loops shall be in operation, or b) Three reactor coolant loops shall be in operation with THERMAL POWER restricted to less than or equal to 65% of RATED THERMAL POWER."

The word "Either:" will be deleted. The designator, or "a)" will be deleted for LCO item a). The word "All" of LCO item a) will be replaced with the word

U.S. Nuclear Regulatory Commission B18814/Attachment 1/Page 4 "Four." LCO item b) will be deleted. The phrase "OPERABLE and" will be inserted after the phrase "loops shall be." The revised LCO will state:

"Four Reactor Coolant Loops shall be OPERABLE and in operation."

The proposed changes eliminate the capability to operate the plant in MODES 1 and 2 with one reactor coolant loop not operating and incapable of providing heat removal from the reactor. The proposed changes retain the requirement that all reactor coolant loops be in operation, but specifies the actual number of reactor coolant loops. The proposed changes add a requirement that all four reactor coolant loops be OPERABLE, as opposed to just operating, thereby explicitly stating that all criteria associated with OPERABILITY must be satisfied.

These are more restrictive changes.

2. The Title of TS 3.4.1.1 will be revised by correcting a typographical error. The word "OPERTION" will be correctly spelled as "OPERATION." This is a non technical change.

Technical Specification 3.4.1.5 Each reactor coolant loop is equipped with loop isolation, valves that permit any loop to be isolated from the reactor vessel. One valve is installed on each hot leg and one on each cold leg. The loop isolation valves are used to perform maintenance on an isolated loop. Plant operation in MODES 1, 2, 3, and 4 with a loop isolated is not permitted, except to mitigate the consequences of an emergency or abnormal event.

The following changes will be made:

Current TS 3.4.1.5 requires that the RCS loop stop valves of an isolated loop shall be shut with power removed from each valve operator while operating in MODES 1, 2, 3, and 4. If these requirements are not met, the associated action statement requires that either the applicable loop stop valves be shut and power removed from the operator within one hour, or the plant be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Consistent with the deletion of requirements associated with 'N-I' loop, or three loop operation this specification is proposed to be revised to require each RCS loop stop valve to be open (four loop operation) during operation in MODES 1, 2, 3, and 4.

1. The word "The" will be replaced with the word "Each." The phrase "of an isolated loop" will be deleted. The word "shut" will be replaced with the word "open." The words "valves" and "operators" will be changed to their singular state. The revised LCO will state, "Each RCS loop stop valve shall be open and the power removed from the valve operator."

The proposed changes remove the option for operating the plant with one or more RCS loop stop valves closed, i.e. an 'N-I' loop configuration, except for the mitigation of an emergency or abnormal event. The proposed change requires that if any RCS loop stop valve is closed or power available to the ,associated

I U.S. Nuclear Regulatory Commission B18814/Attachment 1/Page 5 valve operator, the applicable action statement(s) must be entered and complied with. The proposed changes eliminate the capability to operate the plant with a reactor coolant loop incapable of providing heat removal from the reactor core.

This is a more restrictive change.

2. The Action for this specification will be revised consistent with focusing this specification on maintaining each RCS loop stop valve open. The revised Action will state:

"a. With power available to one or more loop stop valve operators, remove power from the loop stop valve operators within 30 minutes or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.(1) With one or more RCS loop stop valves closed, maintain the valve(s) closed and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />."

The proposed changes require that with power available to any valve operator, such power must be removed within 30 minutes or action must be taken to bring the plant to COLD SHUTDOWN. The proposed changes also require' that with one or more RCS loop stop valves closed, the valve must be maintained closed and action must be taken immediately to bring the plant to COLD SHUTDOWN.

The proposed changes are consistent with the current shutdown times of the existing Action, and provide an appropriate timeframe for performing the required actions consistent with the complexity of the actions required for compliance with the applicable action. These are more restrictive changes.

3. A note, Note (1), will be added to Action Statement b. of TS 3.4.1.5. Note (1) shall state:

",,1 All required actions of Action Statement 3.4.1.5.b. shall be completed whenever this action is entered."

The proposed change requires that all of the required actions associated with Action Statement 3.4.1.5.b. must be completed prior to exiting this action statement. This change clarifies that once a loop stop valve is closed, action must be taken to shutdown the reactor regardless of any other actions which may be taken to restore OPERABILITY of the affected loop stop valve(s). This is a more restrictive change.

4. Surveillance Requirement 4.4.1.5 will be revised consistent with focusing this specification on maintaining each RCS loop stop valve open. The revised surveillance requirement will state:

"Verify each RCS loop stop valve is open and the power removed from the valve operator at least once per 31 days."

U.S. Nuclear Regulatory Commission B18814/Attachment 1/Page 6 The proposed change is consistent with the current surveillance frequency for verification of the state of each RCS loop stop valve, and provides an appropriate timeframe for performing the required actions. Surveillance Requirement 4.4.1.5 is performed at least once per 31 days to ensure that the RCS loop stop valves are open, with power removed from the loop stop valve operators. The purpose of this surveillance is to ensure that power is removed from the valve operators, since Surveillance Requirement 4.4.1.1 requires verification every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in MODES 1 and 2 that all loops are operating and circulating reactor coolant, thereby ensuring that the loop stop valves are open.

The frequency of 31 days ensures that the required flow can be made available, is based on engineering judgment, and has proven to be acceptable. Operating experience has shown that the failure rate is so low that the 31 day Frequency is justified. The proposed change eliminates the capability to operate the plant with one loop isolated. This is a more restrictive change.

5. The title of this specification will be revised by replacing the phrase "ISOLATED LOOP" with the phrase "LOOP STOP VALVES," consistent with the proposed changes described herein for this specification. This is a non-technical change.

Precedence The proposed changes to TS 3.4.1.5 are similar to License Amendment No. 231 to Operating License NPF-4 and Amendment No. 212 to Operating License respectively).

NPF-7 (North Anna Units 1 and 2, Technical Specification 3.4.1.6 TS 3.4.1.6 provides the requirements for opening the RCS loop stop valves of an isolated reactor coolant loop while operating in MODES 5 and 6. License Amendment No. 202(6) issued changes to TS 3.4.1.6, including changes which affected restrictions that were included as part of the original Millstone Unit No. 3 licensing basis. As noted in the NRC Safety Evaluation for this amendment, the Millstone Unit No. 3 design has been historically modified to include additional design features such as cold leg and hot leg loop stop valves interlocks to ensure safe operation of the unit while operating in an N-1 or three loop configuration. DNC does not propose any modifications to these design features as part of this license amendment request.

Additionally, as part of the issuance of this amendment, a license condition was added to the facility license prohibiting operation with one or more loops isolated in MODES 1 through 4, except for the mitigation of abnormal or emergency events when operating in MODES 3 or 4. Since this specification has already been modified consistent with four loop operation (three loop operation prohibited) including NRC review and concurrence (issuance of Amendment No. 202), and since Millstone Unit No. 3 has unique design (6) Victor Nerses, U.S. NRC to J. A. Price, "Millstone Nuclear Power Station, -Unit No. 3 - Issuance of Amendment Re: Reactor Coolant System - Isolated Loop Startup (TAC No. MB1785)," dated January 9, 2002.

U.S. Nuclear Regulatory Commission B18814/Attachment 1/Page 7 features as a result of the historic capability for three loop operation in MODES 1, 2, 3, and 4 (necessitating the retention of the existing requirements of Specification 3.4.1.6),

no additional changes are proposed for this specification.

Index Pages Index page viii will be revised to reflect the correct titles for existing Figures 3.4-4a and 3.4-4b. The correct title for Figure 3.4-4a is "High Setpoint PORV Curve For the Cold Overpressure Protection System" and the correct title for Figure 3.4-4b is "Low Setpoint PORV Curve For the Cold Overpressure Protection System." These are non-technical changes.

Index pages iii, iv, v, vii, ix, and xii will be revised consistent with the proposed changes described herein.

Technical Specification Bases TS Bases Sections 2.1.1, 2.2.1, 3.2.2.2, 3.2.3.1, 3.2.3.2, 3.2.5, 3.3.5, 3.4.1.1, 3.4.1.5, 3.4.9, and 3.7.1.1 shall be modified consistent with the proposed changes described herein.

Safety Summary The proposed changes to Millstone Unit No. 3 TSs 2.1.1, 2.2.1, 3.1.1.1.1, 3.1.1.1.2, 3.1.3.1, 3.1.3.2, 3.2.1.1, 3.2.1.2, 3.2.2.1, 3.2.2.2, 3.2.3.1, 3.2.3.2, 3.2.5, 3.3.1, 3.3.2, 3.3.5, 3.4.1.1, 3.4.1.5, 3.7.1.1, 3.10.2.1, and 3.10.2.2 do not pose a condition adverse to safety and do not create any adverse safety consequences. The rationale for this conclusion is provided below.

The proposed changes to TS 3.4.1.1, which add a requirement that all four (4) reactor coolant loops be in operation, ensure that the applicable reactor coolant loop is removing heat from the reactor core, thereby minimizing the probability of inappropriate localized temperatures within the reactor.

TS 3.4.1.5 currently requires that the loop stop valves of an isolated loop be closed with power removed in MODES 1 through 4. The proposed changes would require that all loop stop valves be open with power removed in MODES 1 through 4, preventing operation in a configuration which is currently prohibited through the use administrative controls. This configuration is consistent with existing operating practices and administrative requirements and ensures that all reactor coolant loops are OPERABLE and capable of removing heat from the reactor core during plant operation. The proposed changes require that if any RCS loop stop valves are closed, actions are taken to shut down the plant. The proposed changes also require that if power is available for a loop stop valve closure, timely action is taken to minimize the probability of loop stop valve closure by removal of such electrical power, thereby maintaining the capability of reactor core heat removal. This change, together with the proposed changes to TSs 3.4.1.1 and 3.1.1.1.2, as well as existing specifications 3.4.1.2 and 3.4.1.6, will prohibit N-1 operation and ensure the existing accident analyses remain

U.S. Nuclear Regulatory Commission B18814/Attachment 1/Page 8 bounding.

With the elimination of the N-1 accident analysis, the facility accident analyses (FSAR Chapter 15) will only address power operation with all loop stop valves open and four RCS loops in operation. The proposed changes to TS 3.4.1.1 to require all four reactor coolant loops to be OPERABLE and in operation, ,and the proposed changes to TS 3.4.1.5 to maintain each RCS stop valve open are consistent with the facility accident analyses.

For operation in Mode 3, FSAR Chapter 15 describes analyses for two events, the uncontrolled rod withdrawal from sub critical event and the boron dilution event. FSAR Section 15.4.1 describes the assumptions made for the uncontrolled rod withdrawal from subcritical. For this event, three loops are assumed to be in operation with all of the loop stop valves open. This assumption bounds operation with the loop stop valves closed. The proposed changes to TS 3.4.1.5, in conjunction with the requirements of TS 3.4.1.2, ensure plant operation is maintained consistent with the current FSAR accident analysis.

FSAR Section 15.4.6 describes the assumptions made for the boron dilution event.

Two different cases are analyzed for a Mode 3 boron dilution event: Case 1, with all four loops unisolated, and Case 2 with three loops unisolated and one loop isolated with the loop stop valves closed. TS 3.1.1.1.2 provides two figures (Figures 3.1-1 and 3.1-2) which reflect the different boron concentration limits for the two analyzed cases.

Since the revised TS 3.4.1.5 will maintain all loop stop valves open, Case 2 is no longer necessary and will be deleted from the facility FSAR, and will also result in the deletion of TS Figure 3.1-2. The requirements of the Mode 3, Case 1 analysis will continue to be assured by the revised TS.

As described in FSAR Section 15.4.4, the startup of an inactive reactor coolant pump at an incorrect temperature and boron concentration is not analyzed because it is precluded by existing TS requirements and administrative procedures. TS 3.4.1.6 requires an isolated loop in MODES 5 and 6 to remain isolated until the isolated loop is within 20 degrees F of the operating loops and the boron concentration is greater than that required for MODES 5 or 6. This TS is not being changed. Since the revised TS 3.4.1.5 will require all loop stop valves to be open in MODES 1 through 4, the RCS loops can only be isolated/unisolated in MODES 5 and 6, where TS 3.4.1.6 will continue to assure that the startup of a reactor pump at an incorrect temperature and boron concentration is precluded. Thus, no changes are required to FSAR Section 15.4.4.

Finally, the capability to use the loop stop valves to mitigate transients and accidents will be retained, as reflected in the current license condition and in the proposed changes to TS Bases Section 3.4.1.5.

The remaining proposed changes to eliminate the option of plant operation with a 'N-I' (three loop) configuration will not modify the value of any parameter which relates to current plant operation (i.e. four loop operation). Changing or removing the capability to operate in an 'N-I' configuration does not impact any analytical methods, nor does it have any impact on the calculations performed for current or future reloads.

U.S. Nuclear Regulatory Commission B18814/Attachment 1/Page 9 In summary, the analyses of record for Millstone Unit No. 3 four loop operation, as described within the facility FSAR, are not impacted by the proposed changes. No new analyses have been created, no existing analyses have been modified, and no analyses which apply to four loop operation have been deleted. The ,proposed changes have no effect on how any of the associated systems or components function to mitigate the consequences of any design basis accident. Administrative controls are in in a 'N-I' configuration.(4) place for Millstone Unit No. 3 which prohibit plant operation The proposed changes remove the requirements from the TSs which allow plant operation in a configuration that is administratively prohibited.

Therefore, the proposed changes to TSs 2.1.1, 2.2.1, 3.1.1.1.1, 3.1.1.1.2, 3.1.3.1, 3.1.3.2, 3.2.1.1, 3.2.1.2, 3.2.2.1, 3.2.2.2, 3.2.3.1, 3.2.3.2, 3.2.5, 3.3.1, 3.3.2, 3.3.5, 3.4.1.1, 3.4.1.5, 3.7.1.1, 3.10.2.1, and 3.10.2.2 will not have an adverse impact on public health and safety and the proposed changes are safe.

Docket No. 50-423 B18814 Attachment 2 Millstone Power Station, Unit No. 3 License Basis Document Change Request 3-13-02

'N-I Loop Operation Siqnificant Hazards Consideration

U.S. Nuclear Regulatory Commission B18814/Attachment 2/Page 1 License Basis Document Change Request 3-13-02 Elimination of 'N-I' Loop Operation from Technical Specifications Significant Hazards Consideration Description of License Amendment Request Dominion Nuclear Connecticut, Inc. (DNC), hereby proposes to amend Operating License NPF-49 by incorporating the attached proposed changes into the Millstone Unit No. 3 Technical Specifications.

Specifically, the Technical Specification changes will:

Delete Technical Specification requirements associated with 'N-I' loop, or three loop plant operation. Delete references to four loop operation whose only purpose is to distinguish from three loop operation; Add requirements to maintain RCS loop stop valves open at all times during MODES 1 through 4; and

• Require that each RCS loop be maintained OPERABLE at all times during MODES 1 and 2.

The Bases and Index for these Technical Specifications will also be modified to reflect these changes as applicable. Refer to Attachment 1 of this submittal for a detailed discussion of the proposed changes.

Basis for No Significant Hazards Consideration In accordance with 10 CFR 50.92, DNC has reviewed the proposed changes and has concluded that they do not involve a Significant Hazards Consideration (SHC). The basis for this conclusion is that the three criteria of 10 CFR 50.92(c) are not compromised. The proposed changes do not involve a SHC because the changes do not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes do not alter the way any structure, system, or component functions and would not alter the way which the plant is operated. The proposed changes do not involve any physical plant modifications. The proposed changes incorporate existing plant operational restrictions into the facility Technical Specifications and provide for the removal of information which is not applicable to plant operation.

The proposed allowed outage times (i.e. the required action times for Specification 3.4.1.5) are reasonable and consistent with the existing technical specification outage times and consistent with industry guidelines, thereby ensuring affected components are restored in a timely manner. The proposed changes to surveillance requirements are also consistent with existing

U.S. Nuclear Regulatory Commission B18814/Attachment 2/Page 2 surveillance frequencies and focus the Technical Specifications on verifying normal plant configurations are maintained. The design basis accidents, including the uncontrolled rod withdrawal from subcritical and boron dilution events, will remain the same postulated events described in the Millstone Unit No. 3 Final Safety Analysis Report (FSAR), and the consequences of these events will not be affected.

Therefore, the proposed changes will not increase the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not alter the plant configuration (no new or different type of equipment will be installed) or require any new or unusual operator actions. The proposed changes do not alter the way any structure, system, or component functions and do not alter the manner in which the plant is operated.

The proposed changes do not introduce any new failure modes. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Involve a significant reduction in a margin of safety.

The proposed changes will not reduce the margin of safety since they have no impact on any accident analysis assumption. The proposed changes do not decrease the scope of equipment currently required to be OPERABLE or subject to surveillance testing, nor do the proposed changes affect any instrument setpoints or equipment safety functions. The effectiveness of Technical Specifications will be maintained since the changes will not alter the operation of any component or system, nor will the proposed changes affect any safety limits or safety system settings which are credited in a facility accident analysis.

Therefore, there is no reduction in a margin of safety.

Docket No. 50-423

B18814 Attachment 3 Millstone Power Station, Unit No. 3 License Basis Document Change Request 3-13-02 Elimination of 'N-I' Loop Operation from Technical Specifications Marked Up Pages

U.S. Nuclear Regulatory Commission B18814/Attachment 3/Page 1 License Basis Document Change Request 3-13-02 Elimination of 'N-I' Loop Operation from Technical Specifications List of Affected Pages Technical Title of Section  ; 'Affected Page with

.Specification~ Amendmen Nmer section Number ____________________

Index iii, Original Issue iv, Amendment 197 v, Amendment 207 vii, Amendment 204 viii, Amendment 204 ix, Amendment 126 xii, Amendment 207 2.1.1 Safety Limits - Reactor Core 2-1, Amendment 173 2-2, Amendment 60 2-3, Amendment 60 B 2-1, Amendment 60 2.2.1 Limiting Safety System Settings - Reactor 2-5, Amendment 159 Trip System Instrumentation Setpoints 2-6, Amendment 159 2-7, Amendment 159 2-8, Amendment 159 2-9, Amendment 152 B 2-4, Amendment 159 B 2-5, Amendment 152 B 2-8, Amendment 85 3.1.1.1.1 Boration Control, Shutdown Margin - Modes 3/4 1-1, Amendment 113 1 And 2 3.1.1.1.2 Boration Control, Shutdown Margin - Modes 3/4 1-3, Amendment 164 3, 4 and 5 Loops Filled 3/4 1-4, Amendment 164 3/4 1-5, Amendment 164 3.1.3.1 Reactivity Control Systems, Movable Control 3/4 1-21, Amendment 191 Assemblies - Group Height 3.1.3.2 Reactivity Control Systems, Position 3/4 1-23, Amendment 207 Indication Systems - Operating 3.2.1.1 Power Distribution Limits, Axial Flux 3/4 2-1, Amendment 60 Difference - Four Loops Operating 3.2.1.2 Power Distribution Limits, Axial Flux 3/4 2-3, Amendment 60 Difference - Three Loops Operating 3/4 2-4, Amendment 60 3.2.2.1 Power Distribution Limits, Heat Flux Hot 3/4 2-5, Amendment 170 Channel Factor, Fo(Z) - Four Loops Operating

U.S. Nuclear Regulatory Commission B18814/Attachment 3/Page 2 Technical Title' f Section Affected Page with S cAmndment Number Section Number 3.2.2.2 Power Distribution Limits, Heat Flux Hot 3/4 2-12, Amendment 120.

Channel Factor, Fa(Z) - Three Loops 3/4 2-13, Amendment 120 Operating 3/4 2-14, Amendment 99 3/4 2-15, Amendment 120 3/4 2-16, Amendment 99 3/4 2-17, Amendment 120 3/4 2-18, Amendment 99 B 3/4 2-3, Amendment 60 B 3/4 2-4, Amendment 170 3.2.3.1 Power Distribution Limits, RCS Flow Rate 3/4 2-19, Amendment 114 and Nuclear Enthalpy Rise Hot Channel Factor - Four Loops Operating 3.2.3.2 Power Distribution Limits, RCS Flow Rate 3/4 2-22, Amendment 60 and Nuclear Enthalpy Rise Hot Channel 3/4 2-23, Amendment 100 Factor - Three Loops Operating 3.2.5 Power Distribution Limits - DNB Parameters 3/4 2-28, Amendment 60 B 3/4 2-5, Amendment 60 B 3/4 2-6, Amendment 60 3.3.1 Instrumentation - Reactor Trip System 3/4 3-2, Amendment 116 Instrumentation 3/4 3-3, Amendment 129 3/4 3-4, Amendment 164 3/4 3-5, Amendment 164 3/4 3-12, Amendment 164 3.3.2 Instrumentation - Engineered Safety 3/4 3-20, Amendment 70 Features Actuation System Instrumentation 3/4 3-28, Amendment 159 3.3.5 Instrumentation - Shutdown Margin Monitor 3/4 3-82, Amendment 164 B 3/4 3-7, Amendment 164 3.4.1.1 Reactor Coolant System - Reactor Coolant 3/4 4-1, Original Issue Loops and Coolant Circulation B 3/4 4-1, Amendment 197 3.4.1.5 Reactor Coolant System - Isolated Loop 3/4 4-7, Original Issue B 3/4 4-1a, Amendment 202 3.4.9 Pressure/Temperature Limits - Reactor B 3/4 4-10, Amendment 197 Coolant System (Except the Pressurizer) 3.7.1.1 Plant Systems - Turbine Cycle 3/4 7-1, Amendment 57 3/4 7-2, Amendment 102 B 3/4 7-1, Amendment 102 3.10.2.1 Special Test Exceptions, Group Height, 3/4 10-2, Original Issue Insertion, and Power Distribution Limits Four Loops Operating 3.10.2.2 Special Test Exceptions, Group Height, 3/4 10-3, Original Issue Insertion, and Power Distribution Limits Three Loops Operating

INDEX SAFFTY I TMTTS ANn I TMTTTNI, -AFFTY SYSTFM ;FTTTNf.

SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE .............. .................. ......... .. 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE . . .. 2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - OUR LOOPS IN OPERATI . . . 2-2 2REACTOR CORE SAFETY LIMIT - THREE LOOPS IN OPERATION.. 2-3 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS ..... .......... 2-4 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS . . . . 2-5 AASFS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE .................. ......................... B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE ...... ................ ... B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS .............. B 2-3 MILLSTONE - UNIT 3 iii

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY ........... ... ........................ 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - MODES 1 AND 2 .... ............ .. 3/4 1-1 Shutdown Margin - MODES 3, 4, AND 5 LOOPS FILLW . 3/4 1-3 FIGURE 3.1-1 REQUIRED SHUTDOWN MARGIN FOR MODE 3..

(TTHII FOUR LOUPS IN OPERATION . ./4 1-4 FIGURE 3.1-2 i*-QUIRED SHUTDOWN MARGIN FOR MODE 3 WITH THREE LOOPS IN OPERATION 1-5 FIGURE 3.1-3 REQUIRED SHUTDOWN MARGIN FOR MODE 4 ........... .. 3/4 1-6 FIGURE 3.1-4 REQUIRED SHUTDOWN MARGIN FOR MODE 5 WITH RCS LOOPS FILLED .... ............ S. .. . 3/4 1-7 Shutdown Margin - Cold Shutdown Loops Not Filled . . . . .. . . . . . . . . . . . .3/4 1-8 FIGURE 3.1-5 REQUIRED SHUTDOWN MARGIN FOR MODE 5 WITH RCS LOOPS DRAINED .... ............... 3/4 1-9 Moderator Temperature Coefficient .......... 3/4 1-10 Minimum Temperature for Criticality ......... 3/4 1-12 3/4.1.2 BORATION SYSTEMS DELETED .......... .................... 3/4 1-13 DELETED .......... .................... 3/4 1-14 DELETED .......... ..................... 3/4 1-15 DELETED .......... ..................... 3/4 1-16 DELETED .......... .................... 3/4 1-17 DELETED .......... .................... 3/4 1-18 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height . . . . . . . . . . . . . . . . . . . . 3/4 1-20 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD . ... . . . 3/4 1-22 Position Indication Systems - Operating . ... . . . 3/4 1-23 MILLSTONE - UNIT 3 iv Amendment No. P, **f,*

y,

INDEX JulY-40,2D4 2 -L LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE DELETED Rod Drop Time ............................................ 3/4 1-25 Shutdown Rod Insertion Limit ............................. 3/4 1-26 Control Rod Insertion Limits ............................. 3/4 1-27 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE.......................

(Four Loops Operating .....................................

3/4 .2 .2 ,Three LoopsHOTOperating HEAT FLUX F CO CHANN EL ..........................

c 7 . . . . . . . .".... .

fFOUr Loops Operating .....................................

3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR .................................. ................ 3/4 2-19 )

hree Loops Uperating......................... S............

/Qhree Loops Operating ........................ .. /.....

42-22 3/4.2.4 QUADRANT POWER TILT RATIO .................... S............ 3/4 2-24 3/4.2.5 DNB PARAMETERS ........................................... 3/4 2-27 TABLE 3.2-1 DNB PARAMETERS ........................................ 3/4 2-28 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION ...................... 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION ................... 3/4 3-2 TABLE 3.3-2 DELETED TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ............................................. 3/4 3-10 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION .......................................... 3/4 3-15 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION .......................................... 3/4 3-17 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS ........................... 3/4 3-26 MILLSTONE - UNIT 3 V Amendment No. , , *, *J,

INDEX January 29, 2001 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-5 DELETED TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . 3/4 3-36 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring for Plant Operations .... 3/4 3-42 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS .............. 3/4 3-43 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS ...... 3/4 3-45 TABLE 3.3-7 DELETED TABLE 4.3-4 DELETED TABLE 3.3-8 DELETED TABLE 4.3-5 DELETED Remote Shutdown Instrumentation ... .......... . . . 3/4 3-53 TABLE 3.3-9 REMOTE SHUTDOWN INSTRUMENTATION ... .......... . . . 3/4 3-54 TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ..... ............ . . . 3/4 3-58 Accident Monitoring Instrumentation ........... . . . 3/4 3-59 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION .......... . . . 3/4 3-60 TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . 3/4 3-62 TABLE 3.3-11 DELETED I

TABLE 3.3-12 DELETED TABLE 4.3-8 DELETED MILLSTONE - UNIT 3 vi Amendment No. go, Y7, 1??, 193

I INDEX M~a-y8,..2MQ2-e LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-13 DELETED TABLE 4.3-9 DELETED 3/4.3.4 DELETED 3/4.3.5 SHUTDOWN MARGIN MONITOR ........ .................. .. 3/4 3-82 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation .... ......... 3/4 4-1 Hot Standby ...... .................. 3/4 4-2 Hot Shutdown ................ 3/4 4-3 Cold Shutdown - Loops Filled ........ 3/4 4-5 Cold Shutdown - Loo Nt Filled ... 3/4 4-6 Usolated Looj....- - - -................... 3/4 4-7 Isolated Loop Startup .... ............. 3/4 4-8 3/4.4.2 SAFETY VALVES ........... ....................... ..3/4 4-9 DELETED ............. .......................... .. 3/4 4-10 3/4.4.3 PRESSURIZER Startup and Power Operation ........ ................ 3/4 4-11 FIGURE 3.4-5 PRESSURIZER LEVEL CONTROL ..... ............... .. 3/4 4-11a Hot Standby ........... ........................ .. 3/4 4-11b 3/4.4.4 RELIEF VALVES ........... ....................... .. 3/4 4-12 3/4.4.5 STEAM GENERATORS ........... ..................... .. 3/4 4-14 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION ........ ................ 3/4 4-19 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION .... ............ .. 3/4 4-20 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems .......... ................. 3/4 4-21 Operational Leakage .......... .................... .. 3/4 4-22 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES . . 3/4 4-24 3/4.4.7 DELETED ............. .......................... .. 3/4 4-25 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS ......... 3/4 4-26 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS . . . . . ... . . . . . . . . . . . .

. 3/4 4-27 3/4.4.8 SPECIFIC ACTIVITY .................... 3/4 4-28 MILLSTONE - UNIT 3 vii Amendment No. 00 XF11, W , X797J,

May-8--2OO2-§--

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE FIGURE 3.4-1 DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >ILCi/gram DOSE EQUIVALENT 1-131 ......... ................... 3/4 4-30 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM ........... .......................... ... 3/4 4-31 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ...... ................ .. 3/4 4-33 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 10 EFPY ...... ................. ... 3/4 4-34 FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 10 EFPY ...... ................. ... 3/4 4-35 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE ......... .................... .. 3/4 4-36 Pressurizer ........... . ........ . . .... ... 3/4 4-37 Overpres S stems.5 3/4 4-38 FIGURE 3.4-4a NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COLD OVERPRESSURE F

SYSTEM I

(FOUR LOOP OPERATION) ... ....... . .. /4 4-40 U

FIGURE 3.4-4b NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COLD OVERPRESSURE SYSTEM (THREE LOOP OPERATION) ......... .... 3/4 4-41 3/4.4.10 DELETED ............ ......................... ... 3/4.4-42 II 3/4.4.11 DELETED ............ ......................... ... 3/4 4-43 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS ............. ....................... 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350°F . 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 350°F .... .......... .. 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK ..... ............... .. 3/4 5-9 3/4.5.5 pH TRISODIUM PHOSPHATE STORAGE BASKETS ............. .. 3/4 5-10 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containmlent Integrity ........ .................. ... 3/4 6-1 Containment Leakage ........ ................... .. 3/4 6-2 Containment Air Locks ........ .................. ... 3/4 6-5 Containment Pressure ......... ................... 3/4 6-7 MILLSTONE - UNIT 3 viii Amendment No. M 97, PY, 779, Of

Insert A to Page viii Figure 3.4-4a High Setpoint PORV Curve For the Cold Overpressure Protection System Figure 3.4-4b Low Setpoint PORV Curve For the Cold Overpressure Protection System

F-ebfuaiy-5499q6'7 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS LECTION

- PAGE Air Temperature ............. ... . 3/4 6-9 Containment Structural Integrity ... .......... . . . 3/4 6-10 Containment Ventilation System. .......... . . . 3/4 6-11 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Quench Spray System . . . 3/4 6-12 Recirculation Spray System ... ............

. . . 3/4 6-13 3/4.6.3 CONTAINMENT ISOLATION VALVES ... ............. . . . 3/4 6-15 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Monitors ................ . . . 3/4 6-16 Electric Hydrogen Recombiners .......... . . . 3/4 6-17 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM Steam Jet Air Ejector .............. . . . 3/4 6-18 3/4.6.6 SECONDARY CONTAINMENT Supplementary Leak Collection and Release System .. . 3/4 6-19 Secondary Containment ............ S. .. 3/4 6-22 Secondary Containment

.. . .,3/4 6-23 Structural Integrity ..... ...............

6-2 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves . . .- 3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH,-INOPERABLE STEAM LINE SAFF3TY VALVES DU GFOUR LOOP OPERAT I)ff* > .-* . . .... 3/4 7-2

• )(IUM*LLOWABLE POWER RANGE NEUTRON FLUX HIGH TABLE 3.7-2 SEPINT WITH INOPERABLE TEAM LINE SAFETYVLE LOP OPERATION . .. .. .. .. .. . .. 3/4 7-2 MILLSTONE - UNIT 3 ix Amendment No. Y, R F7, F7, 799, 779, IV

5/27/98 INDEX

)

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS I SECTION PAGE TABLE 3.7-3 STEAM LINE SAFETY VALVES PER LOOP S. . . . . . .. . 3/4 7-3 Auxiliary Feedwater System ............. 3/4 7-4 Demineralized Water Storage Tank .......... 3/4 7-6

-Specific-Activity- ;-....; .-.... 3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM ................ 3/4 7-8 Main Steam Line Isolation Valves .......... 3/4 7-9 Steam Generator Atmospheric Relief Bypass Lines . . 3/4 7-9a I 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION . . 3/4 7-10 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM .... 3/4 7-11 3/4.7;4 SERVICE WATER SYSTEM ................ 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK ................. 3/4 7-13 3/4.7.6 FLOOD PROTECTION .................. 3/4 7-14 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM 3/4 7-15 3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM .... 3/4 7-18 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM .......... 3/4 7-20 3/4.7.10 SNUBBERS . . . . ................... 3/4 7-22 TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL ......... 3/4 7-27 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST ..... 3/4 7-29 3/4.7.11 SEALED SOURCE CONTAMINATION ............ 3/4 7-30 3/4.7.12 DELETED Table -3.7-4 DELETED Table 3.7*5 DELETED "3/4-.7.13 .DELETED. "

3/4.7.14 AREA TEMPERATURE MONITORING ............ 3/4 7-32 TABLE 3.7-6 AREA TEMPERATURE MONITORING ............ 3/4 7-33 MILLSTONE 0583 UNIT 3 Amendment No. 1, i60 i67, O,

February 2, 2001 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES Operating . . . . . . . . . . . . 3/4 8-1 DELETED . . . . . . . . . . . . . 3/4 8-9 I Shutdown . . . . . . . . . ... . 3/4 8-10 3/4.8.2 D.C. SOURCES

..... ... .. 3/4 8-11 Operating .

3/4 8-13 TABLE 4.8-2a BATTERY SURVEILLANCE REQUIREMENTS 3/4 8-14 TABLE 4.8-2b BATTERY CHARGER CAPACITY .....

Shutdown . . . . . . . . . . . . . 3/4 8-15 3/4.8.3 ONSITE POWER DISTRIBUTION Operating . . . . . . . . . . . . S....... 3/4 8-16 S. . . . . . 3/4 8-18 Shutdown . . . . . . . . . . . . .

3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEV ICES

. . S....... 3/4 8-19 DELETED . . . . . . . . . . .

S....... 3/4 8-21 DELETED . . . . . . . . . . . . .

DELETED . . . . . . . . . . . . . S....... 3/4 8-22 3/4.9 REFUEL ING OPERATIONS 3/4.9.1 BORON CONCENTRATION ....... 3/4 9-1 3/4 9-2 3/4.9.2 INSTRUMENTATION .........

3/4.9.3 DECAY TIME ............ 3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS 3/4 9-4 3/4.9.5 COMMUNICATIONS .......... 3/4 9-5 MILLSTONE - UNIT 3 xi Amendment No. ff, 7l1, 797, 194

Jui4y-a30-,A2-00 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9.6 REFUELING MACHINE ......... ................... * . 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS ....... * . 3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level . . . . . . . . . . . . . . . . . . . . 3/4 9-8 Low Water Level . . . . . . . . . . . . . .. . . . 3/4 9-9 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM . . . . . 3/4 9-10 3/4.9.10 WATER LEVEL - REACTOR VESSEL ............ * . 3/4 9-11 3/4.9.11 WATER LEVEL - STORAGE POOL ............. . . 3/4 9-12 3/4.9.12 FUEL BUILDING EXHAUST FILTER SYSTEM ... .......... .

• 3/4 9-13 3/4.9.13 SPENT FUEL POOL - REACTIVITY ............ . . 3/4 9-16 3/4.9.14 SPENT FUEL POOL - STORAGE PATTERN ..... .......... . . 3/4 9-17 FIGURE 3.9-1 MINIMUM FUEL ASSEMBLY BURNUP VERSUS NOMINAL INITIAL ENRICHMENT FOR REGION I 4-OUT-OF-4 STORAGE CONFIGURATION ........ .................... . . 3/4 9-18 FIGURE 3.9-2 REGION I 3-OUT-OF-4 STORAGE FUEL ASSEMBLY LOADING SCHEMATIC .................................... . . 3/4 9-19 FIGURE 3.9-3 MINIMUM FUEL ASSEMBLY BURNUP VERSUS NOMINAL INITIAL ENRICHMENT FOR REGION 2 STORAGE CONFIGURATION . . . . . 3/4 9-20 FIGURE 3.9-4 MINIMUM FUEL ASSEMBLY BURNUP AND DECAY TIME VERSUS NOMINAL INITIAL ENRICHMENT FOR REGION 3 STORAGE CONFIGURATION ........ .................... . . 3/4 9-21 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN . . . ... 4 1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS I (Four Loops Operating .................. 3/4 10-2 t QThree Loops O.erating ........... . 3/4 10-3 3/4.10.3 PHYSICS TESTS ............ ...................... . .3/4 10-4 3/4.10.4 REACTOR COOLANT LOOPS ............ .................. 3/4 10-5 3/4.10.5 DELETED 3/4.11 DELETED 3/4.11.1 DELETED 3/4.11.2 DELETED 3/4.11.3 DELETED MILLSTONE - UNIT 3 xii Amendment 7y, g, 7gg, 799,

2.0 SAFETY LIMITS'AND LIMITING SAFETY SYSTEM SETTINGS -August.13 9-'2 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tg,) shall not exceed the limits shown in Figur.. 2.1-1. and 2.1-2 for four and three oo operation, respectivey *--.--I APPLICABILITY: MODES I and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop) average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within I hour.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2750 psia be in HOT STANDBY with the Reactor Coolant System pressure within its limit within I hour.

Q MODE*S 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2750 psiai reduce the Reactor Coolant System pressure to within its limit within 9) 5 minutes.

MILLSTONE - UNIT 3 2-1 Amendment No. **

J9 680 660 Li. 640 C,

LzJ 620 600 580 560 0.4 0.6 0.8 1.0 1.2 0 0.2 FRACTION OF RATED T14ERMAL POWER REACTOR CORE SAFETY I

2-2 Amendment "o" 0 1

UILLSTONE - UNIT 3

6'cd MILLSTONE - UNIT 3 2-3 Amendment No. O

5/26/98 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock SetpQ:I-ts shall be set consistent with the Nominal Trip Setpoint values shown in Table 2.2-1. I APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

a. With a Reactor Trip System Instrumentation Channel or Interlock Channel Nominal Trip Setpoint inconsistent with the value shown in the Nominal Trip Setpoint column of Table 2.2-1, adjust the Setpoint consistent with the Nominal Trip Setpoint value.

b. With a Reactor Trip System Instrumentation Channel or Interlock Channel found to be inoperable, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status.

MILLSTONE - UNIT 3 2-4 Amendment No. 1-59 0550

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.

NOMINAL CD FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUL M

N.A. N.A.

1. Manual Reactor Trip

2. Power Range, Neutron Flux

a. High Setpoint 109% ofRTP**

Iq' Four Loops Op erat Ing 109.6% of RTP**

1)Four Loops Operating 80% of RTP** 80.6%of RTP**

L2) Three Loops Operating 25% of RTP** 25.6% of RTP**

b. Low Setpoint 5% of RTP** with 5.6% of RTP** with

3. Power Range, Neutron Flux, a time constant High Positive Rate a time constant k 2 seconds _k 2 seconds N

(7'

4. Deleted Ct, Ct, r1 z

0 5.

6.

7.

Intermediate Range, Neutron Flux Source Range, Neutron Flux Overtemperature AT

a. Four Loops Operating 25% of RTP**

1 X 10* 5cps cz' .

< 27.4% of RTP**

<_1.06 x 10"5 cps V

1) Channels I, II See Note 1 See Note 2!

2) Channels Ill, IV See Note 1

• • RTP - RATED THERMAL POWER

TABLE 2.2-1 (Continued)

DrArTnD TOTO VTFM TJTPItMFNTATTfl1

• j s..s p TPTP *FTPO TNT*

ncArTnD TDTD CVCTPM ILfl"

,n., I'JI'p *,p I lISP ti TNCTDHMPNTATTnN TPTP IZFTPOTNT-q NOMINAL FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE 8

-4

8. Overpower AT (Four Loops Operating)) See Note 3 See Note 4

9. Pressurizer Pressure-Low 1900 psia k 1897.6 psia

10. Pressurizer Pressure-High 2385 psia _2387.4 psia 89% of instrument < 89.3% of instrument

11. Pressurizer Water Level-High span span 90% of loop > 89.8% of loop

12. Reactor Coolant Flow-Low design flow* design flow*

18.1% of narrow k 17.8% of narrow

13. Steam Generator Water range instrument I range instrument Level Low-Low span 0

span N.A. N.A.

14. General Warning Alarm 92.4% of rated > 92.2% of rated

15. Low Shaft Speed - Reactor speed speed Coolant Pumps U1 I

• Minimum Measured Flow Per Loop I/4 of the RCS Flow Rate Limit as listed in Section 3.2.3.1. (Four 1

L Opsperaing); / o e RCS Flow Rate Limit as listed in Sect 2.a r/ erating

REACTOR TRIP SYSTEM INiRUMENTATION TRIP SETPOINTS NOMINAL FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE 0

16. Turbine Trip zn a. Low Fluid Oil Pressiure 500 psig k 450 psig 3C

b. Turbine Stop Valve 1% open k 1%open Closure

17. Safety Injection Input N.A. N.A.

from ESF

18. Reactor Trip System Interlocks

a. Intermediate Range 1 x 10`0 amp k 9.0 x 10"11 amp Neutron Flux, P-6

b. Low Power Reactor Trips

-44 Block, P-7

1) P-10 input (Note 5) 11% of RTP** < 11.6% of RTP**

2) P-13 input 10% RTP** Turbine < 10.6% RTP** Turbine Impulse Pressure Impulse Pressure Equivalent Equivalent

c. Power Range Neutron 2)7, aS- p - *3i. 6-'- P

-ir Flux, P-8 37.5% of RTP** 38.1% of RTP**

0D C 111) 7Four Loops Ooperating M

2)Three Looops Opelrating 37.5% of RTP** <_38.1% of R

• • RTP = RATED THERMAL POWER

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS z NOMINAL ALLOWABLE VALUE FUNCTIONAL UNIT 51% of RTP** 051.6% of RTP**

d. Power Range Neutron Flux, P-9 9% of RTP** > 8.4% of RTP**

e. Power Range Neutron Flux, P-10 (Note 6)

U N.A. N.A.

19. Reactor Trip Breakers N.A. N.A.

20. Automatic Trip and Interlock Logic T

N.A. N.A

1. ree Loop Operation Bypass Circuitry rJ' DLQ V

CD zD C+

03

-p Lfo

• • RTP = RATED THERMAL POWER

I, I j *.

TABLE 2.2-1 (Continued)

TABLE NOTATIONS SNOTE 1: OVERTEMPERATURE AT AT (1+ S) K, -K2(+s) 1% 0 AT / 2 2S K

(1 +,5sS)

(T - T7) +K3 (P -p) -fl(,&I)

C, Where: AT is measured Reactor Coolant System AT, *F; ATo is loop specific indicated AT at RATED THERMAL POWER, *F; (1÷ 1 s)

(1.? 2s) is the function generated by the lead-lag compensator on measured AT; r, and r2 are the time constants utilized in the lead-lag compensator for AT, r, 8 sec, T2-<

3 sec;

, K, 1.20((Four Loops Operating); _K1.20 (Three Loops Operating;X .

K2 0.02456/°F; 0

(1 l+s) 5 is the function generated by the lead-lag compensator for T',;

. r 4 and r. are the time constants utilized in the lead-lag compensator for T.. T4 20 sec, rs 4 sec; T Is measured Reactor Coolant System average temperature, 'F; 0 T' Is loop specific Indicated Tavg at RATED THERMAL POWER, < 587.16F; K>

K3 O.OOl311/psi

• P is measured pressurizer pressure, psia; P' is nominal pressurizer pressure, k 2250 psia; s is the Laplace transform operator, sec';

TABLE 2.2-1 (Continued) 0 TABLE NOTATIONS (Continued) r1 TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued)

NOTE 1: (Continued) and fl(AI) is a function of the indicated difference between top and bottom detectors of the power range neutron ion chambers; with nominal gains to be selected based on measured instrument response during plant startup tests calibrations such that:

(1) For q, - qb between -26% and +3%, f 1,(AI) k 0, where q, and qb are percent RATED THERMAL POWER in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RATED THERMAL POWER; (2) For each percent that the magnitude of q,- qb exceeds -26%, the AT Trip Setpoint shall be N automatically reduced by > 3.55% of its value at RATED THERMAL POWER.

(3) For each percent that the magnitude of q%- qb exceeds +3%, the AT Trip Setpolnt shall be automatically reduced by > 1.98% of its value at RATED THERMAL POWER.

NOTE 2: The maximum channel as left trip setpolnt shall not exceed its computed trip setpoint by more than the following:

(1) 0.4% AT span for the AT channel (2) 0.4% AT span for the Tavo channel (3) 0.4% AT span. for the prehsurizer pressure channel (4) 0.8% AT span for the f(AI) channel 0 ~Lf

• VN

- 1.0 I-I,-

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB. This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux and is indicative of the margin to DNB.

The DNB design basis is as follows: uncertainties in the WRB-1 or WRB-2 correlations, plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, and computer codes are considered statistically such that there is at least a 95 percent probability with 95 percent confidence level that DNBR will not occur on the most limiting fuel rod during Condition I and II events. This establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties. In addition, margin has been maintained in the design by meeting safety analysis DNBR limi s 'n perform* gsfety analyses.

The curves of Figur 2.1-1_-d 2.1-2 how the loci of points of THERMAL POWER, Reactor Coolant System pressure, and average temperature below which the calculated DNBR is no less than the design DNBR value or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

These curves are based on an enthalpy hot channel factor, F H, of 1.70 (includes measurement uncertainty) and a reference cosine with a I&ak ofNI.55 for axial power shape. An allowance is included for an increase in FAH at reduced power based on the expression:

FNH = 1.70 [1 + 0.3 (I-P)]

A where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming axial imbalance is within the limits of F (delta I) function of the Overtemperature trip. When the axial power imbalince is not within the tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits.

MILLSTONE - UNIT 3 B 2-1 Amendment No. fo 0004

JAN 31 1986 SAFETY LIMITS BASES 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor vessel, pressurizer, and the RCS piping, valves and fittings are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2750 psia) of design pressure.

The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated Code requirements.

The entire RCS is hydrotested at 125% (3125 psia) of design pressure, to demonstrate integrity prior to initial operation.

MILLSTONE - UNIT 3 B 2-2

S~5/26/98 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Nominal Trip Setpoints specified in Table 2.2-1 are the nom~inal values at which the reactor trips are set for each functional unit. The Allowable Values (Nominal Trip Setpoints +/- the calibration tolerance) are considered the Limiting Safety System Settings as identified in 10CFR50.36 and have been selected to ensure that the core and Reactor Coolant System are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.

The Setpoint for a Reactor Trip System or interlock function is considered to belbonsistent with the nominal value when the measured "as left" Setpoint is within the administratively controlled (+/-) calibration tolerance identified in plant procedures (which specifies the difference between the Allowable Value and Nominal Trip Setpoint). Additionally, the Nominal Trip Setpoints may be adjusted in the conservative direction provided the calibration tolerance remains unchanged.

Measurement and Test Equipment accuracy is administratively controlled by plant procedures and is included in the plant uncertainty calculations as defined in WCAP-10991. Operability determinations are based on the use of Measurement and Test Equipment that conforms with the accuracy used in the plant uncertainty calculation.

The Allowable Value specified in Table 2:2-1 defines the limit beyond which a channel is inoperable. If the process rack bistable setting is measured within the "as left" calibration tolerance, which specifies the difference between the Allowable Value and Nominal Trip Setpoint, then the channel is considered to be operable.

The methodology, as defined in WCAP-10991 to derive the Nominal Trip Setpoints, is based upon combining all of the uncertainties in the channels.

Inherent in the determination of the Nominal Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels should be capable of operating within the allowances of these uncertainty magnitudes. Occasional drift in excess of the allowance may be determined to be acceptable based on the other device performance characteristics. Device drift in excess of the allowance that is more than occasional, may be indicative of more serious problems and would warrant further investigation.

The various reactor trip circuits automatically open the reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level. In addition to the redundant channels and trains, the design approach provides Reactor Trip System functional diversity. The MILLSTONE - UNIT 3 B 2 -3 Amendment No. 159 0551

A6/9B-2.2 LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) functional capability at the specified trip setting is required for those anticipatory or diverse reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates a turbine trip signal whenever reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.

• 1r Manual Reactor Trip The Reactor Trip System includes manual Reactor trip capability.

Power Range, Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip setting. The Low Setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during powe operations to mitigate the consequences of a reactivity excursion from Il powelevs.J The-High Setpoint rIpIs r.ýe-dufce6d during three loop op~eration_

C~t ale cnsitet a wththe safety analysis./,

The Low Setpoint trip may be manually block6d above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.

Power Range, Neutron Flux, High Positive Rate The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.

Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for all rod ejection accidents.

MILUTONE - UNIT 3 B 2 - 4 Amendment Nlo. 71yJ, #0

LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source, Rance, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon trolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels will initiate a Reactor trip at about 10 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-1O becomes active. No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Trip System.

Overtemperature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors, and pressure is within the range between the Pressurizer High and Low Pressure trips. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1. Although a direction of conservatism is identified for the Overtemperature AT reactor trip function K? and K3 gains, the gains should be set as close as possible to the values contained in Note.1 to ensure that the Overtemperature AT set oint i consistent with the assumptions of the safety analyses. 1 SOperation with a reactor coolant loop out of service requires Reactor Trip System modification. Three loop operation is permissible after resetting the K] ýinput to the Overtemperature AT channels, reducing the Power Range Neutron Flux High setpoint to a value just above the three loop maximum permissible power level, and resetting the P-8 setpoint to its three loop value. These modifications have been chosen so that, in three loop operation, each component of the Reactor Trip System performs its normal four loop function, prevents operation outside the safety limit curves, and prevents the DNBR from going below the design limit during normal operational and antici patdtansients.

Overpower AT The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT 050_6_ orE MILMONE - UNIT 3 B 2-5 Amendment No. 77, ,

LIMITING SAFETY SYSTEM SE7TINGS JAN 31 1986 BASES "

trip, and provides a backup to the High Neutron Flux trip. The Setpoint is automatically varied with: (1) coolant temperature to correct for tempera ture induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, to ensure that the allowable heat genera tion rate .(kW/ft) is not exceeded. The Overpower AT trip provides protection to mitigate the consequences of various size steam-breaks as reported in .

Excessive Secondary Steam Releases."

WCAP-9226, "Reactor Core Response to Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted.

The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.

On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.

The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.

Pressurizer Water Level The Pressurizer Water Level High trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power the Pressurizer High Water Level trip is automatically blocked by P-7 (a power level of approxi mately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, auto matically reinstated by P-7.

Reactor Coolant Flow The Reactor Coolant Flow Low trip provides core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.

On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10%

of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow. Above P-8 (a power level of approximately 38% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. Conversely, on decreasing power between P-8 and the P-7 an automatic Reactor trip will occur on low reactor coolant flow in more than one loop and below P-7 the trip function is automatically blocked.

UNT 3 MILLTONE- 2-MILLSTONE - UNIT 3 8 2-6

LIMITING SAFETY SYSTEM SETTINGS T")r~o c}A 10/21/97 BASES Steam Generator Water Level The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam/feedwater flow mismatch resulting from loss of normal feedwater. The specified Setpoint provides allowances for starting delays of the Auxiliary Feedwater System.

Low Shift Speed -'Reactor Coolant Pumps The Low Shaft Speed - Reactor Coolant Pumps trip provides core protection to prevent DNB in the event of a sudden significant decrease in reactor coolant pump speed (with resulting decrease in flow) on two reactor coolant pumps in any two operating reactor coolant loops. The trip setpoint ensures that a reactor trip will be generated, considering instrument errors and response times, in sufficient time to allow the DNBR to be maintained greater than the design above limit following a four-pump loss of flow event.

Turbine Trip A Turbine trip initiates a Reactor trip. On decreasing power the Reactor trip from the Turbine trip is automatically blocked by P-9 (a power level of

-approximately 50% of RATED THERMAL POWER); and on increasing power, reinstated automatically by P-9.

Safety Iniection InDut from ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF automatic actuation logic channels will initiate Reactor trip upon any signal which initiates a Safety Injection.

The ESF instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-3.

Reactor Trip System Interlocks The Reactor Trip System interlocks perform the following functions:

P-6 On increasing power, P-6 becomes active above the Interlock Allowable Value specified on Table 2.2-1 to allow the manual block of the Source Range trip (i.e., prevents premature block of the 2 Source Range trip during reactor startup) and deenergizes the high voltage to the detectors. On decreasing power during a reactor shutdown, Source Range Level trips are automatically reactivated and high voltage restored when P-6 deactivates. The P-6 deactivation will occur at a value below its activation value and may be calibrated to occur below the P-6 Interlock Allowable Value specified on Table 2.2-1 to prevent overlap and chatter based upon the expected bistable drift.

P-7 On increasing power P-i automatically enables Reactor trips on low flow in more than one reactor coolant loop, reactor coolant pump low shaft speed, pressurizer low pressure and pressurizer high level. On decreasing power, the above listed trips are automatically blocked.

MILLSTONE - UNIT 3 B 2-7 Revised by NRC Letter 0548 dated September 29. 1997 Enclosure

LTITITG SAFETy SYSTEK METINGS BASES Reactor Trip System Interlocks (Continued) --

P-8 On increasing power, P-8 automatically enables Reactoetfrfps on low flow in one or more reactor coolant loops. On decreasing power, the P-8 automatically blocks the above listed trips.

P-9 On increasing power, P-9 automatically enables Reactor tri p on Turbine trip. On decreasing power, P-9 automatically blocks Reactor trip on Turbine trip.

P-1O On increasing power, P-]0 provides input to P-7 to ensure that Reactor Trips on low flow in more than one reactor-coolant loop, reactor coolant pump low shaft speed, pressurizer low pressure and pressurizer high level are active when power reaches 11%. It also allows the manual block of the Intermediate Range trip and the Low Setpoint Power Range trip; and automatically blocks the Source Range trip and deenergizes the Source Range high voltage power.

On decreasing power, P-10 resets to automatically reactivate the Intermediate Range trip and the Low Setpoint Power Range trip before power drops below 9%. It also provides input to reset P-7.

P-13 Provides input to P-7.

Three Loop Operation Bypass Circuitry The Three Loop Operation Bypass Circuitry reactor trip ensures that a suf ficient number and the correct combination of trip circuits remain available to provide necessary protection and mitigation capability during three loop operation. Should more than two channels in one train or two dissimilar channels in two trains be bypassed, a reactor trip will occur. In this manner, it is ensured that sufficient protective features remain to mitigate the consequences of analyzed transients.

ILLTONE - UNIT 3 B 2-8 Amendment No. 0

3/4. 1 REACTIVITY CONTROL SYSTEMS 3/4.1.1.1 BORATION CONTROL SHUTDOWN MARGIN - MODES I AND 2 LIMITING CONDITION FOR OPERATION 3.1.1.1.1

ýot_-ot_our Iloop The and SHUTDOWN three MARGINshall loop oerattihe great r tha or equal to 1.3% Ak/k.ft

.APPLICABILITY: MODES 1 and 2*.

ACTION:

With the SHUTDOWN MARGIN less than 1.3% Ak/k, immediately initiate and con tinue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.3% Ak/k:

a. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod(s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above "required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s);

b. When in MODE I or MODE 2 with K," greater than or equal to 1 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that control bank withdrawal is withIn the limits of Specification 3.1.3.6;

c. When in MODE 2 with KI, less than 1, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;

d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specification 4.1.1.1.2, with the control banks at the maximum insertion limit of Specification 3.1.3.6; and

• See Special Test Exceptions Specification 3.10.1.

MILONE - UNIT 3 3/4 1-1 Amendment No. 19, flj

March 11, 1991 3)pýb Gvýý REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within +/- 1% Ak/k at least once per 31 Effec tive Full Power Days (EFPD). This comparison shall consider at least the fol l owing factors:

1) Reactor Coolant System boron concentration,

2) Control rod position,

3) Reactor Coolant System average temperature,

4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration.

The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading.

MILLSTONE - UNIT 3 3/4 1-2 Amendment No. 60 0007

3/4.1 REACTIVITY CONTROL SYSTEMS 3L4.1.1 BORATION CONTROL SHUTDOWN MARGIN - MODES 3. 4 AND 5 LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.1.1.1.2 The SHUTDOWN MARGIN shall be greater tt shown in Figures 3.1-1, 3.1-3 and 3.1-4 fýr four 1.1-2 for t ree loop opera ion APPLICABILITY: MODES 3, 4 and 5

-ACTION:

With the SHUTDOWN MARGIN less than the required value, immediately initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.2.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the required value:

a. Within I hour after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod(s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s); and

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following factors:

1) Reactor Coolant System boron concentration,

--'--2) Control rod position,

3) Reactor Coolant System average temperature,

4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration.

4.1.1.1.2.2 Valve 3CHS-V305 shall be verified closed and locked at least once per 31 days.

• Additional SHUTDOWN MARGIN requirements, if required, are given in Specification 3.3.5.

MILWONE - UNIT 3 3/4 1-3 Amendment No. YR, yfl, Jý$

I-4 I I 4

I-(2050,3.495) 00 2' 3.5 (2500,3.495)

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Ca, Z2.5 S2 I ____________________

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FIGURE 3.1-1 oOP- g6 REQUIRED SHUTDOWN FOR MODE 3(WITH FOUR LOOPS IN OPERATIOt

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"FIGURE 33 .1

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oxz rn CD 6.5 z

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RCS CRITICAL BORON CONCENTRATION (ppm) CD O!

FIGURE 3.1-3 REQUIRED SHUTnnWN MARGIN FOR MODE 4

a 0=

--1 7

6.5 6

5.5 5

z 4.5 D

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-I 3.5 Lr 3

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0 500 1,000 1,500 2,000 2,500 RCS CRITICAL BORON CONCENTRATION (ppm) CD

!q cl, FIGURE 3.1-4 REQUIRED SHUTDOW1N MARGIN FOR MODE 5 14ITII RCS LOOPS FILLED

REACTIVITY CONTROL SYSTEMS .).2IC.rK .

3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT March 11, 1991 ITMITTING CONITION FOR OPFRATION 3.1.3.1 All full-length shutdown and control rods shall be OPERABLE and positioned within +/-12 steps (indicated position) of their group step counter demand position.

APPLICABILITY: MODES I and 2 ACTION:

a. With one or more full-length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With one full-length rod trippable but inoperable due to causes other than addressed by ACTION a., above, or misaligned from its group step counter demand height by more than +/-12 steps (indicated position), POWER OPERATION may continue provided that within I hour:

1. The rod is restored to OPERABLE status within the above alignment requirements, or

2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within

+12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Specification 3.1.3.6. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or

3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:

a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions; b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

MILLSTONE - UNIT 3 3/4 1-20 Amendment No. go, 60

REACTIVITY CONTROL SYSTEMS Dreem.er-2-20f LIMITING CONDITION FOR OPERATION ACTION (Continued) c) A power distribution map is obtained from the movable incore detectors within their limits FQ(Z) and and within FN. are verified to be 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s-and d) Wth four loops operating, the HER2A&LOER level is re 6 ess than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER.orZ-e) With three loops operating, the THERMAL POWER level is reduced to less than or equal to 50% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the Neutron Flux High Trip Setpoint is reduce to less than or equal to 60% of RATED THERMAL POWER.

c. With more than one rod trippable but inoperable due to causes other than addressed by ACTION a. above, POWER OPERATION may continue provided that:

1. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the remainder of the rods in the bank(s) with the inoperable rods are aligned to within +12 steps of the inoperable rods while maintaining the rod sequence and insertion limits of Specification 3.1.3.6. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, and

2. The inoperable rods are restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

d. With more than one rod misaligned from its group step counter demand height by more than +12 steps (indicated'position), be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the rod position deviation monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 92 days.

MILLSTONE - UNIT 3 3/4 1-21 Amendment No. X, *, J

I CMar ch II,1991 TABLE 3 -1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Loss of Reactor Coolant from Small Ruptured Pipes or from Cracks in Large Pipes Which Actuates the Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal at Full Power

_-Major Reactor Coolant System Pipe Ruptures (Loss-of-Coolant Accident)

Major Secondary Coolant System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)

MILLSTONE - UNIT 3 3/4 1-22 Amendment No. P, 60 1

REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS - OPERATING I1MITING CONDITTON FOR OPFRATION 3.1.3.2 The Digital Rod Position Indication System and the Demand Position Indication System shall be OPERABLE and capable of determining the control rod positions within +/-12 steps.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. With a maximum of one digital rod position indicator per bank inoperable:

1. Determine the position of the nonindicating rod(s) indirectly by the movable incore detectors at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and immediately after any motion of the nonindicating rod which exceeds 24 steps in one direction since the last determ~ntti of the rod's position, or

2. With four loops operating, reduce THERMAL POWER to less than 32%of RATED THERMAL POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. t

b. With a maximum of one demand position indicator per bank inoperable:

1. Verify that all digital rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of ea ter at least once per 8 hourý,.o

2. QiWth four loops operatingp, educe THERMAL POER to less Man 50% ot RATED THERMAL PUwER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. o...) Lky..A

3. With three loops operating, reduce THERMAL POWER to less thaný 32% of RATED THERMAL POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

SIIRVFTI IANCF RFQIITRFMFNT-S 4.1.3.2.1 Each digital rod position indicator shall be determined to be OPERABLE by verifying that the Demand Position Indication System and the Digital Rod Position Indication System agree within 12 steps at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the rod position deviation monitor is inoperable, then compare the Demand Position Indication System and the Digital Rod Position Indication System at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.1.3.2.2 Each of the above required digital rod position indicator(s) shall be determined to be OPERABLE by verifying that the digital rod position indicators agree with the demand position indicators within 12 steps when exercised over the full-range of rod travel at least once per 24 months.

MILLSTONE - UNIT 3 3/4 1-23 Amendment No. , ,, 7,

Ma-Hr19P 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE A LIMITING CONDITION FOR OPERATION 3.2.1.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:

-a. The limits specified in the CORE OPERATING LIMITS REPORT (COLR) for Relaxed Axial Offset Control (RAOC) operation, or

b. Within the target band about the target flux difference during base load operation, specified in the COLR.

APPLICABILITY: MODE I above 50% RATED THERMAL POWER*.

ACTION:

a. For RAOC operation with the indicated AFD outside of the applicable limits specified in the COLR, specified

1. Either restore the indicated AFD to within the COLR limits within 15 minutes, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux--High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b. For base load operation above APLND with the indicted AFD outside of the applicable target band about the target flux differences:

1. Either restore the indicated AFD to within the COLR specified target band within 15 minutes, or

2. Reduce THERMAL POWER to less than APLND of RATED THERMAL POWER and discontinue base load operation within 30 minutes.

c. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR.

• See Special Test Exception 3.10.2 I

MILTONE - UNIT 3 3/4 2-:1 Amendment No. ,

WOJ 0;;*-z1~.

%A.,-,.k 11 1001 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS be within its limits 4.2.1.1.1 The indicated AFD shall be determined to by:

during POWER OPERATION above 50% of RATED THERMAL POWER channel at

a. Monitoring the indicated AFD for each OPERABLE excore OPERABLE:

least once per 7 days when the AFD Monitor Alarm is excore

-b. Monitoring and logging the indicated AFD for each OPERABLEat least and channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Alarm is once per 30 minutes thereafter, when the AFD Monitor be assumed inoperable. The logged values of the indicated AFD shall to exist during the interval preceding each logging.

of its limits when 4.2.1.1.2 The indicated AFD shall be considered outside AFD to be outside the two or more OPERABLE excore channels are indicating the limits.

difference of each 4.2.1.1.3 When in base load operation, the target flux at least once per OPERABLE excore channel shall be determined by measurement 4.0.4 are not 92 Effective Full Power Days. The provisions of Specification applicable.

shall be 4.2.1.1.4 When in base load operation, the target flux difference determining updated at least-once per 31 Effective Full Power Days by either requirements the target flux difference in conjunction with the surveillance the most of Specification 4.2.1.1.3 or by linear interpolation between cycle life.

recently measured value and the calculated value at the end of The provisions of Specification 4.0.4 are not applicable.

MILLSTONE - UNIT 3 3/4 2-2 Amendment No. .0,60 0011

March II, 499-f' POWER DISTRIBUTION LIMITS AXIAL FLUX DIFFERENCE THREE LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.2.1.2 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:

-a. The limits specified in the CORE OPERATING LIMITS REPORT (COLR) for Relaxed Axial Offset Control (RAOC) operation, or

b. Within the target band specified in the COLR about the target flux difference during base load operation.

APPLICABILITY: MODE I above 37.5% of RATED THERMAL POWER.*

ACTION:

a. For RAOC operation with the indicated AFD outside of the applicable limits specified in the COLR,

1. Either restore the indicated AFD to within the COLR specified limits within 15 minutes, or

2. Reduce THERMAL POWER to less than 37.5% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux--High Trip setpoints to less than or equal to 41% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b. For base load operation above APLND with the indicated AFD outside of the applicable target band about the target flux differences:

1. Either restore the indicated AFD to within the COLR specified

- .target band within 15 minutes, or

2. 'Reduce THERMAL POWER to less than APLND of RATED THERMAL POWER and discontinue base load operation within 30 minutes.

c. THERMAL POWER shall not be increased above 37.5% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR.

• See Special Test Exception 3.10.2.

MILLSTONE - UNIT 3 3/4 2-3 Amendment No.

POWER DISTRIBUTION LIMITS P

SURVEILLANCE REQUIREMENTS 4.2.1.2.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 37.5% of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE excore channel at least once per 7 days when the AFD Monitor Alarm is OPERABLE:

-b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

4.2.1.2.2 The indicated AFD shall be considered outside of its limits when two or more OPERABLE excore channels are indicating the AFD to be outside the limits.

4.2.1.2.3 When in base load operation, the target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days. The provisions of Specification 4.0.4 are not applicable.

4.2.1.2.4 When in base load operation, the target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference in conjunction with the surveillance requirements of Specification 4.2.1.2.3 or by linear interpolation between the most recently measured value and the calculated value at the end of cycle life.

The provisions of Specification 4.0.4 are not applicable.

MILLSTONE - UNIT 3 3/4 2-4 Amendment No. W#

POWER DISTRIBUTION LIMITS 314.2.2 HEAT FLUX HOT CHANNEL FACTOR - F,(Z)

OUR LOOPS OPERATING 90 LI 2VONýDýý ITION FOR OPERATION 3.2.2.1 F,(Z) shall be limited by the following relationships:

F RTP F0 (Z) _< '0 K(Z) for P > 0.5 P

F RTP F0(Z) < -- K(Z) for P < 0.5 0.5 FRTP = the F. limit at RATED THERMAL POWER (RTP) provided in the core operating limits report (COLR).

Where: P - THERMAL POWER and RATED THERMAL POWER K(Z) = the normalized F0 (Z) as a function of core height specified in the COLR.

APPLICABILITY: MODE 1.

ACTION:

With FQ(Z) exceeding its limit:

a. For RAOC operation with Specification 4.2.2.1.2.b not being satisfied or for base load operation with Specification 4.2.2.1.4.b not being satisfied:

(1) Reduce THERMAL POWER at least 1% for each 1% F0 (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower AT Trip setpoints have been reduced at least 1% for each 1% Fa(Z) exceeds the limit, and MILLSTONE - UNIT 3 3/4 2-5 Amendment No. 3,yo, F, 779,

POWER DISTRIBUTION LIMITS October 18, 1995 HEAT FLUX HOT CHANNEL FACTOR - F THREE LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.2.2.2 FQ(Z) shall be limited by the following relationships:

"F 0 (Z) * [K(Z)] for P > 0.375 P

F0 (Z) F-F-;P

( [K(Z)] for P : 0.375 FoTP = The F. limit at RATED THERMAL POWER (RTP) specified in the CORE OPERATING LIMITS REPORT (COLR).

THERMAL POWER , and Where: P=

RATED THERMAL POWER K(Z) = the normalized F,(Z) as a function of core height speci fied in the COLR.

APPLICABILITY: MODE 1.

ACTION:

With FO(Z) exceeding its limit:

a. For RAOC operation with F,(Z) outside the applicabl e limit specifie d in the COIR:

(1) Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the applicable AFD limits by 1% AFD for each percent Fo(Z) exceeds its limits. Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reset the AFD alarm setpoints to these modified limits, or (2) Reduce THERMAL POWER at least 1% for each 1% F,(Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent MILLSTONE - UNIT 3 3/4 2-12 Amendment No. Pp, f, Y7, N

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)

POWER OPERATION may proceed provided the Overpower AT Trip Setpolnts have been reduced at least 1% for each 1% Fo(Z)

The Overpower AT Trip Setpoint reduction exceeds the limit.

shall be performed with the reactor in at least HOT STANDBY, or I (3) Verify that the requirements of Specification 4.2.2.1.3 for base load operation are satisfied and enter base load operation.

Where it is necessary to calculate the percent that F,(Z) exceeds the limits for Items (1) and (2) above, it shall be calculated as the maximum percent over the core height (Z) that F0 (Z) exceeds its limit by the following expression:

[F.Fo(Z)

F0o~

x W(Z) osdtIpxt00 for x K(Z) 0.375 P

F03 ~

(Z) x W (Z) -1x100 for P< 0.375 0.375xKZ

.=

b. For base load operation outside the applicable limit specified in the COLR, perform either of.the following actions:

(1) Place the core in an equilibrium condition where the limit in 4.2.2.2.2.C is satisfied, and remeasure FA(Z), or (2) Reduce THERMAL POWER at least 1% for each 1% F,(Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F0 (Z)

The Overpower AT Trip Setpoint reduction K

exceeds the limit.

shall be performed with the reactor in at least HOT STANDBY.

The percent that Fa exceeds the limit shall be calculated as the maximum percent over the core height (Z) that F,(Z) exceeds the limit using the following expression:

Amendment No. M1, A KILLSTONE - UNIT 3 3/4 2-13

December 29, 1994 POWER DISTRIBUTION LIMIITS LIMITING CONDITION FOR OPERATION (Continued)

Ig (Z)

~ x W(Z)B j -10xl fox P IAPL'

c. Identify and correct the cause of the out-of-limit condition prior S*...* UDMai onujr phabve the rpduced limit required by ACTION a or b, above; THERMAL POWER may then be increased provided F,(Z) is demonstrated through incore mapping to be within its limit.

flVF1I I ANCE REOVIREMENTS SURVETUANCE REnuIREMENTS 4.2.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2.2 For RAOC operation, FO(Z) shall be evaluated to determine if Fa(Z) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b. Increasing the measured F0 (Z) component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties. Verify the requirements of Specification 3.2.2.2 are satisfied.

c. Satisfy the following relationship:

FIR'r x K(Z)

FfTjx(Z) s a for P > 0.375

-Z P x W(Z)

Z:FT x K(Z) for P s 0.375 F'(Z) HW(Z) x 0.375 where FA(Z) is the measured F0 (Z) increased by the allowances for manufacturing tolerances and measurement uncertainty, F1 "' is the F0 limit, K(Z) is the normalized F0 (Z) as a function of core height, P is the relative THERMAL POWER, and W(Z) is the cycle-dependent function that accounts for power distribution transients encountered during normal operation. F0P, K(Z), and W(Z) are specified in the COLR as per Specification 6.9.1.6.

MILSTONE - UNIT 3 3/4 2-14 Amendment No. 0, ,.

October 18, 1995 P DSTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

d. Measuring FA(Z) according to the following schedule:

(1) Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWER at which F,(Z) was last determined,* or (2) At least once per 31 Effective Full Power Days, whichever occurs first.

e. With the maximum value of F"(Z)

K(Z) over the core height (Z) increasing since the previous determination of F'(Z), either of the following actions shall be taken:

(1) FA(Z) shall be increased over that specified in Specification 4.2.2.2.2c by an appropriate factor specified in the COLR, or (2) F4(Z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that the maximum value of F"(Z)

K(Z) over the-core height (Z) is not increasing.

f. The limits specified in Specifications 4.2.2.2.2c and 4.2.2.2.2e are not applicable in the following core plane regions:

(1) Lower core region from 0% to 15%, inclusive.

(2) Upper core region from 85% to 100%, inclusive.

• During power escalation at the beginning of each cycle, the power level may be (

increased until a power level for extended operation has been achieved and Spower distribution map obtained.

- UNIT 3 aus-%<

3/4 2-15 Amendment.No. 17, f, F

December 29, 1994-* -'

POWER DlI"MRIBUION LIMITS (Continued)

SURVEILLANCE REQUIREMENTS if the following 4.2.2.2.3 Base load operation is permitted at powers above APLND conditions are satisfied:

APLD

a. Prior to entering base load operation, maintain THERMAL POWER abovefor at and less than or equal to that allowed by Specification 4.2.2.2.2 least the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Maintain base load operation surveillance of (AFD within the target band limit about the target flux difference Specification 3.2.1.2) during this time period.

is Base load operation is then permitted providing THERMAL POWER is maintained between APLND and APLu or between APLND and 100% (whichever most limiting) and Fo surveillance is maintained pursuant to of:

Specification 4.2.2.2.4. APOL is defined as the minimum value Apt'= Fo" xK(Z)

F' (Z) x W(Z) over the core height (Z) where: F,(Z) is the measured F0 (Z) increased by the allowances for manufacturing tolerances and measurement uncertainty. The Fo limit is FOTP- W(Z)B. is the cycle-dependent function that accounts for limited power distribution transient encountered during base load operation. FoFn*, K(Z), and W(Z)BL are specified in the COLR as per Specification 6.9.1.6.

During base load operation, if the THERMAL POWER is decreased before below b.

APLND then the conditions of 4.2.2.2.3.a shall be satisfied reentering base load operation.

to determine if 4.2.2.2.4 During base load operation F,(Z) shall be evaluated Fo(Z) is with~ji, its limit by:

a. Using'the movable incore detectors to obtain a power distribution map above APL"D.

at any THERMAL POWER map

b. Increasing the measured FO(Z) component of the power distribution by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties. Verify the requirements of Specification 3.2.2.2 are satisfied.

/

3/4 2-26 Amendment No. 77, F9, F9, # I MILTONE - UNIT 3

October 18, 1995 POWE. DStRIBUyION tMTS SURVEILLANCE REQUIREMENTS(I ontinued)

c. Satisfying the fol Ilowing relationship:

Z) S Fr x K(Z) for P > APLND r x. W1 JB'I'L r x wkL)AL where: FA(Z) is the measured Fo(Z). The F'W is the F0 limit, the normalized Fa(Z) as a function of core height. P is the relative I

THERMAL POWER. W(Z)O Is the cycle-dependent function that accounts for limited power distribution transients encountered during base load operation. FF*, K(Z), and W(Z)k are specified in the COLR as per Specification 6.9.1.6.

d. Measuring FA(Z) in conjunction with target flux difference determina tion according to the following schedule:

(1) Prior to entering .base load operation after satisfying Sec tion 4.2.2.2.3, unless a full core flux map has been taken in the previous 31 Effective Full Power Days with the relative THERMAL POWER having been maintained above APLNO for the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to mapping, and (2) At least once per 31 Effective Full Power Days.

e. With the maximum value of F'"(z)

K(z) over the core height (Z) increasing since the previous determination of Fo(Z), either of the following actions shall be taken:

(1) Fm,(Z) shall be increased over that specified in 4.2.2.2.4.c by an appropriate factor specified in the COLR, or NILLSTONE - UNIT 3 3/4 2-17 Amendment No. 77, 9P, MP, PF, 0330-U-

frSPOWER December 29, 1994 DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

(2) Fm(Z) shall be measured at least once per 7 Effective Full Power Days until 2 successive maps indicate that the maximum value of aF(Z)

K(Z) f.

over the core height (Z) is not increasing.

The limits specified in 4.2.2.2.4.c and 4.2.2.2.4.e are not applicable in the following core plane regions:

(1) Lower core region D% to 15%, inclusive.

4 (2) Upper core region 85% to 100%, inclusive.

4.2.2.2.5 When Fo(Z) is measured for reasons other than meeting the require ments of Specifications 4.2.2.2.2 or 4.2.2.2.4, an overall measured Fo(Z) shall be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty. J MILLSTONE - UNIT 3 3/4 2-18 Amendment No. 77, 1p9, gp, ft

May "281995(*

POWER DISTRIBUTION LINITS

• 3/4.2.3 RCS FLOW RATE rAvlND NNUC~LE E1111,LPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.1.3.1' The indicated Reactor Coolant System (RCS) total flow rate and FAm shall be maintained as follows:

a. RCS total flow rate k 371,920 gpm, and

b. S Ff [. + (1.0 - P)]

Where:

1) P THERAL POWER RATED THERMAL POWER

2) FQ - Measured values of F,* obtained by using the movable incore detectors to obtain a power distribution map. The measured value of F, should be used since Specification 3.2.3.1b.

takes into consideration a measurement uncertainty of 4% for incore measurement,

3) FNT - The FA imit at RATED THERMAL POWER in the CORE OPERATING LIMITS REPORT (COLR),

4) PF&4 - The power factor multiplier for FA provided in the COLR, and

5) The measured value of RCS total flow rate shall be used since uncertainties of 2.4% for flow measurement have been included in Specification 3.2.3.1a.

APPLICABILITY: MODE 1.

ACTION:

With the RCS total flow rate or FQ outside the region of acceptable operation:

a. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:

1. Restore the RCS total flow rate and FQ to within the above limits, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

IIUONE - UNIT 3 3/4 2-19 Amendment No. 77, J PP /4/

January 3, 1995 POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued)

b. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate that EL and RCS total flow rate are restored to within the above limits, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

C. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2. and/or b., above; subsequent POWER OPERATION may proceed provided that FH and indicated RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the region of acceptable operation prior to exceeding the following THERMAL POWER levels:

1. A nominal 50% of RATED THERMAL POWER,

2. A nominal 75% of RATED THERMAL POWER, and

3. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of attaining greater than or equal to 95% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.2.3.1.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.1.2 RCS total flow rate and QH shall be determined to be within the acceptable range:

a. -Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and

b. At least once per 31 Effective Full Power Days.

4.2.3.1.3 The indicated RCS total flow rate shall be verified to be within the acceptable range at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the most recently obtained value of F, obtained per Specification 4.2.3.1.2, is assumed to exist.

4.2.3.1.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months. The measurement instrumentation shall be calibrated within 7 days prior to the performance of the calorimetric flow measurement.

MILLSTONE cm) - UNIT 3 3/4 2-20 Amendment No. FP, 7l 100

March 11, 1991 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.3.1.5 The RCS total flow rate shall be determined by precision heat balance the measurement at least once per 18 months. Within 7 days prior to performing of steam precision heat balance, the instrumentation used for determination feedwater venturi AP in pressure, feedwater pressure, feedwater temperature, and the calorimetric calculations shall be calibrated.

4.2.3.1.6 If the feedwater venturis are not inspected at least once per 18 months, an additional 0.1% will be added to the total RCS flow measurement uncertainty.

3/4 2-21 Amendment No. 77,60 MILLSTONE - UNIT 3 0011 I

POWER DISTRIBUTION LIMITS RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR THREE LOOPS OPERATING I TMTTTIJ(

I Lull I Nr rANnITTAN rnp nnTn * *...

FAR APFRATTAN n"P'F 3.2.3.2 The indicated Reactor Coolant System (RCS) total flow rate and FNH shall be maintained as follows:

a. RCS total flow rate > 303,200 gpm, and

b. FNFH <RTP

- AH [1.0 + PFAH (1.0 - P)]

Where:

1) THERMAL POWER ,

= RATED THERMAL POWER

2) FN = Measured values of FN obtained by using the movable iA ore detectors to obtainU power distribution map.

The measured value of FN should be used since Speci -

I fication 3.2.3.2b. takes intV consideration a measure ment uncertainty of 4% for incore measurement,

3) FRIP = The F limit at RATED THERMAL POWER in the CORE OPRATING LIMI+Y REPORT (COLR),

4) PFAH = The power factor multiplier for FAH in the COLR, and

5) The measured value of RCS total flow rate shall be used since uncertainties of 2.8% for flow measurement have been included in Specification 3.2.3.2a.

A*PPLICABILITY: MODE 1.

ACTION:

With the RCS total flow rate or FAH outside the region of acceptable operation:

a. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:

1. Restore the RCS total flow rate and FN to within the above limits, or

2. Reduce THERMAL POWER to less than 32% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 37% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. IJ MLS E - UNIT 3 3/4 2-22 Amendment No. ;,

January 3, 1995 POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued)

b. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of initially being outside the above limits, verify throuah incore flux mapping and RCS total flow rate that F,' and RCS total flow rate are restored to within the above limitsor reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

C. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2. anch/or b., above; subsequent POWER OPERATION may proceed provided that F*H and indicated RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the region of acceptable operation prior to exceeding the following THERMAL POWER levels:

1. A nominal 32% of RATED THERMAL POWER, and

2. A nominal 50% of RATED THERMAL POWER.

SURVEILLANCE. REQUIREMENTS 4.2.3.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2.2 RCS total flow rate and F. shall be determined to be within the acceptable range at least once per 3VEffective Full Power Days.

4.2.3.2.3 The indicated RCS total flow rate shall be verified to be within the acceptable range at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the most recently obtained value of F*H, obtained per Specification 4.2.3.2.2, is assumed to exist.

4.2.3.2.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months. The measurement instrumentation shall be calibratedwithin 7 days prior to the performance of the calorimetric flow measurement.

4.2.3.2.5 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months. Within 7 days prior to performing the precision heat balance, the instrumentation used for determination of steam pressure, feedwater pressure, feedwater temperature, and feedwater venturi AP in the calorimetric calculations shall be calibrated.

4.2.3.2.6 If the feedwater venturis are not inspected at least once per 18 months, an additional 0.1% will be added to the total RCS flow measurement uncertainty.

NILL9N - UNIT 3 . 3/4 22 Amendment No. 779 FP. 719

March 11, 1991 jj47)

POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the limits shown on Table 3.2-1:

a. Reactor Coolant System Tavg, and

b. Pressurizer Pressure.

APPLICABILITY: 'MODE 1.

ACTION:

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to Tess than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.5 Each of the parameters of Table 3.2-1 shall be verified to be within its limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

MILLSTONE 0011

- UNIT 3 3/4 2-27 Amendment No. A7, 60 I

v91 11.

-arzh TABLE 3.2-1 DNB PARAMETERS T/*hree Loops in Opera-*

1" Four Loops in C Operati on .

U-) PARAMETER

< 591.1"F < 583.30F /

Indicated Reactor Coolant System Tavg S > 2218 psi a*S

> 2218 psia*

Indicated Pressurizer Pressure w

r,*

r*3 co (D

0.

o

• Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.

November 3,2000 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.

4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be verified to be within its limit at least once per 18 months.

Neutron detectors and speed sensors are exempt from response time verification.

Each verification shall include at least one train such that both trains are verified at least once per 36 months and one channel (to include input relays to both trains) per function such that all channels are verified at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip functioii as shown in the "Total No. of Channels" column of Table 3.3-1.

-- ~-e MILLSTONE - UNIT 3 3/4 3-1. Amendment No. *F, 7P, P7, 09, 187

TABLE 3.3-1 z

REACTOR TRIP SYSTEM INSTRUMENTATION zm TOTAL NO. CHANNELS MINIMUM CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION S1. Manual Reactor Trip 2 1 2 I, 2 1 2 1 2 3*, 4*, 5* 11

2. Power Range, Neutron Flux

a. High Setpoint 4 2 3 I, 2 2

b. Low Setpoint 4 2 3 1###, 2 2

3. Power Range, Neutron Flux 4 2 3 I, 2 2 High Positive Rate

4. Deleted

5. Intermediate Range, Neutron Flux 2 1 2 1###, 2 3 Source Range, Neutron Flux S6.

a. Startup 2 1 2 2## 4

b. Shutdown 2 1 2 3*, 4*, 5* 11

7. Overte"m a*. our Loop Operation 4 3 It 2

b. Three Loop Operation 2 2

8. a. Four AT Overpower Loop Operation ,.i*. 4 2 3 1, 2 6

b. Three Loop Operation _____ 22 ,-

q4

= 9. Pressurizer Pressure--Low 4 2 3 1** 6 ( 1)"2 6 (

1)'

0.

1? Pressurizer Pressure--High 4 2 3 I, 2

11. Pressurizer Water Level--High 3 2 2 " 6

TABLE 3.3-I (Continued)

REACTOR TRIP SYSTEAINSTRIIENTATION rn MINII*M I FUNCTIONAL UN/IT TOTAL NO. CHANNELS CHANNELS APPLICABLE OFLCHANNELS TO TRIP AflI I

-4 12. Reactor Coolant Flow--Low f (a' a. Single Loop (Above'P-8) 6

b. Two Loops (Above P-7 and below P-8) 6

13. Steam Generator Water Level--Low-Low 6 (1)

Iii (a) 14. Low Shaft Speed--Reactor Coolant Pumps Four loop Lb.a. Three loon operation onpertionn3ipm 4-]/pump 2 3 1** 6 2

15. Turbine Trip 2 -~~

a. Low Fluid Oil Pressure 3 2 4 1WWw 1z

b. Turbine Stop Valve Closure 4 4 1*** 6 I 16.

17.

Deleted Reactor Trip System Interlocks e

3. a. Intermediate Range Neutron Flux, P-6 2 t4 1 2 210 8 C C

z 0

b. Low Power Reactor Trips Block, P-7 I P-IO Input 4 2 3 1 8 or Ci Cl P-13 Input 2 I 2 I 8 C

"-4 TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION m

MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS a-. TO TRIP OPERABLE MODES ACTION S17. Reactor Trip System Interl6cks (Continued)

c. Power Range Neutron Flux, P-8 4 8 2 3

d. Power Range Neutron Flux, P-9 4 2 3 I 8

e. Power Range Neutron Flux, P-1O 4 2 1,2 8 3

18. Reactor Trip Breakers( 2 ) 1 2 2 1, 2 10, 13 2 3*, 4*, 5* 11

19. Automatic Trip and Interlock 2 1 2 Logic . P",j*Q"- 2 1, 2 13A 1 2 3*, 4*, 5* 11

20. rThree Loop Operation 8 2 8

/ Bypass Circuitry (I switch ner (Frnm fiff--

-61

21. DELETED 141 I

TABLE 3.3-1 (Continued)

TABLE NOTATIONS

• When the Reactor Trip System breakers are in the closed position and the Control Rod Drive System is capable of rod withdrawal.

• • Above the P-7 (At Power) Setpoint.

• • • Above the P-9 (Reactor Trip/Turbine Trip Interlock) Setpoint.

1. 1. Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

1. 1. 1. Below the P-1O (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

(1) The applicable MODES and ACTION statements for these channels noted in Table 3.3-3 are more restrictive and, therefore, applicable.

(2) Including any reactor trip bypass breakers that are racked in and closed for bypassing a reactor trip breaker.

ACTION STATEMENTS ACTION I - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1, and do wles

c. Either, THERMAL POWER is restrictedro less than or e ual to 75% of RATED TERMAL POWER or four loop operation or

ý50% of RATED THERMA POW o three loop operation an the Power Range Neutron Flux Trip Setpoint is re duced to less than or equal to 85 % of RATED THERMAL POWER Ifor tour floop operation or 60% of RATED THERMAL POWER -or three loop operation wit in ours; or, e NT POWER is monitored at leas once r 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Specification 4.2.4.2.

MILLSTONE - UNIT 3 3/4 3-5 Amendment No. Y7, y, j, 0492QP

October 21, 1998 TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a. Below the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint, and

b. Above the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.

ACTION 4 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.

ACTION 5 (Not used) I ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1.

ACTION 7 - (Not used)

ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> determine by observation of the associated permissive annunciator window(s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

0ILLSTONE 0492

- UNIT 3 3/4 3-6 Amendment No. P7, 9, 7X7, 164

03/24/94 TABLE 3.3- (CvtInuedi94 ACTION STATEMENTS (Continuedl ACTION 9 (Not used) I ACTION 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.

ACTION 11 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the Reactor Trip System breakers within the next hour.

ACTION 12 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

b. When the Minimum Channels OPERABLE requirement is met, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of the Turbine Control Valves.

ACTION 13 - With one of the diverse trip features (undervoltage or shunt trip attachments) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or declare the breaker inoperable and apply ACTION 10. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time "required for performing maintenance to restore the breaker to OPERABLE status.

ACTION 13A - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable Channel to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT STANDBY within the next 6. hours; however, one channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.1.1, provided the other channel is operable.

MILLSTONE - UNIT 3 3/4 3-7 Amendment No. 70, 89, Cosa

.J?&V I

6/28/94 This page is intentionally left blank HILLSTONE - UNIT 3 Gals 3/4 3-8 Amendment No. 7Z, 0/t 91 I

> o 6/28/94 This page is intentionally left blank MILLS'TONE - UNIT 3 :3/4 3-9 Amendment NPn J g,

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION -TEST IEST LOGIC-TEST IS REQUIRED

i. Manual Reactor Trip N.A. N.A. N.A. R(14) N.A. 1 2, 4*,

2. Power Range, Neutron Flux N.A.

a. High Setpolnt S o2: 4) Q N.A. 1, 2 M (3, 41 (4:, 5' 1***, 2

b. Low Setpolnt S R 4, S/U( 1) N.A. N.A.

3. Power Range, Neutron Flux, N.A. R(4, 5) Q N.A. N.A. 'I, 2 High Positive Rate

4. Deleted

5. Intermediate Range S R(49 5) S/U (1) N.A. N.A. 1***, 2 S R(4, 5) S/U1) N.A. N.A. 2**t, 3, 4 ,

6. Source Range, Neutron Flux 5 Q N.A. N.A. I, 2

7. Overtemperature AT S R R

Q N.A. N.A. 1, 2

8. Overpower AT S c

9. Pressurizer Pressure--Low S R Q(18) N.A. N.A. 1

10. Pressurizer Pressure--High S R Q(18) N.A. N.A. 1, 2

11. N.A. N.A. 1 U, Pressurizer Water Level--High S R Q .

12. N.A. N.A. 1 Reactor Coolant Flow--Low S R Q A

TABLE 4.3-I (Continued) nravvrnn YBW VIEU ?UCTD!EnITATYflN imVrTt I ANCP flFtNIIAFHFI4TS flWVIVI I lI . C~au 1.511nfA~n .U., MBull I..I,..,,,'..

TRIP ACTUATING

-4 ANALOG MODES WHICH FOR I CHANNEL DEVICE OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE w f CHANNEL CHANNEL 15-REQUIRED CHECK CAL1IBAIONf IEST TEST LOGICTEST FUNCTIONAL UJII R Q(18) N.A. N.A. I, 2

13. Steam Generator Water Level-- S Low-Low (AD

14. Low Shaft Speed - Reactor N.A. R(13) Q N.A. N.A. 1 Coolant Pumps

15. Turbine Trip N.A. R H.A. S/U(It1, I)***N.A. 1

a. Low Fluid Oil Pressure S/U(I, 1O)****N.A. I R N.A.

-2

b. Turbine Stop Valve Closure N.A. P I-

16. Deleted

17. Reactor Trip System Interlocks

a. Intermediate Range N.A. N.A.

Neutron Flux, P-6 N.A. R(4) R

b. Low Power Reactor N.A. R(4) R N.A. N.A.

Trips Block, P-7

c. Power Range Neutron N.A. R(4) R N.A. N.A. 1 Flux, P-8 IIt

d. Power Range Neutron R(4) R N.A. N.A.

Flux, P-9 N.A.

e. Power Range N.A. N.A. 1, 2 N.A. R(4) R Neutron F? ux, P-1O

f. Turbine Impulse Chamber R N.A. N.A. 1 PressureP-13 N.A. R C

C:

(D ro

-4 Ft -.1 z..t (0

r TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS z

m TRIP z

ANALOG ACTUATING MODES FOR

-4 CHANNEL DEVICE WHICH C ANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT .AIjECK CALIBRATION TE.ST TEST . LOGIC TEST IS-REQUIRED

18. Reactor Trip Breaker N.A. N.A. N.A. M(7, 11) N.A. 1 2 3*,

19. Automatic Trip and N.A. N.A. N.A... N.A. M(7)

Interlock Logic 4~,~

20. [Three Loop Operation 21.

Bypass Circuitry Reactor Trip Bypass N.A.

N.A.

N.A. N.A. R N.A.

III~z2 N.A. N.A. M(7, Breaker R(16) 15) N.A.

CAi 22. DELETED w

I-'

P

A

QI)ý April 26, -1995 TABLE 4.3-1 (Continued)

TABLE NOTATIONS

• When the Reactor Trip System breakers are closed and the Control Rod Drive System is capable of rod withdrawal.

Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

Below P-1O (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

S** Above the P-9 (Reactor Trip/Turbine Interlock) Setpoint.

(1) If not performed in previous 31 days.

(2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistentiwith calorimetric power if absolute difference is greater than 2%. The provisions of Specification 4.0.4 are not applicable to entry into MODE 2 or 1.

(3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5) Detector plateau curves shall be obtained, and evaluated and compared to manufacturer's data. For the Source Range, Intermediate Range and Power Range Neutron Flux cha.nnels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(8) (Not used)

(9) Quarterly surveillance in MODES 3*, 4*, and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window.

MILSTONE - UNIT 3" 3/4 3-13 Amendment No. FP, 79,109 02.94

L OA2 OCT 2 1 '-S.,-,

3-','-)

TABLE 4.3-1 (Continued)

TABLE NOTATIONS (Continued)

(10) Setpoint verification is not applicable.

(11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

(12) (not used)

(13) Reactor Coolant Pump Shaft Speed Sensor may be excluded from CHANNEL CALIBRATION.

(14) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit(s).

(15) Local manual shunt trip prior to placing breaker in service.

(16) Automatic undervoltage trip.

(17) (not used).

(18) The surveillance frequency and/or MODES specified for these channels in Table 4.3-2 should be reviewed for applicability.

I MILLSTONE - UNIT 3 .3/4 3-14 Amendment No. 7?, F, 79, 7Y, 0494 7l, 164

Li- 6 5126198 INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Nominal Trip Setpoint column of Table 3.3-4.

APPLICABILITY: As shown in Table 3.3-3.

ACTION:

a. With an ESFAS Instrumentation Channel or Interlock Channel Nominal Trip Setpoint inconsistent with the value shown in the Nominal Trip Setpoint column of Table 3.3-4, adjust the Setpoint consistent with the Nominal Trip Setpoint value.

b. With an ESFAS Instrumentation Channel or Interlock Channel found to be inoperable, declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status.

MILLSTONE - UNIT 3 3/4 3-15 Amendment No. Y1, 159

L4ý 6ýj November 3,2000 INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.

4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME* of each ESFAS function shall be verified to be within the limit at least once per 18 months.

Each verification shall include at least one train such that both trains are verified at least once per 36 months and one channel (to include input relays to both trains) per function such that all channels are verified at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the "Total No. of Channels" column of Table 3.3-3.

• The provisions of Specification 4.0.4 are not applicable for response time verification of steam line isolation for entry into MODE 4 and MODE 3 and turbine driven auxiliary feedwater pump for entry into MODE 3.

MILLSTONE - UNIT 3 3/4 3-16 Amendment No. PP, 7Y, P9, 7$,

187

TABLE 3.3-3 o"

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION o-4 m MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE I

FUNCTIONAL UNIT -- MODES OF-CIIHANNLS TO TRIP, OPERABLE ACTION

-4

1. Safety Injection (Reactor Trip, Feedwater Isolation, Control Building Isolation (Manual Initiation Only),

Start Diesel Generators, and Service Water).

a. Manual Initiation 2 1 2 1, 2, 3, 4 19

b. Automatic Actuation 2 2 1, 2, 3, 4 14 (Al Logic and Actuation Relays I?

(A'

.6 b-J

c. Containment 3 2 2 1,2, 3 20 Pressure--High-i

d. Pressurizer 4 2 3 1, 2, 31 20 Pressure--Low

e. Steam Line Pressure- 3/steam line 2/steam line 2/steam line 1, 2, 30 20 Low in each in any in each operating operating operating Z o loop loop loop Zi 0

2. Containment Spray (CDA) 0 3

a. Manual Initiation 2 1 with 2 1, 2, 3, 4 19 2 coincident switches (0

February 21, 1990 TABLE-3.3-3 (Continued) eu.qlurrnrn CRrrrV rATItU

• )tirL, I I Li1UVILF AITIIATIflP FVIW' i'... -YVTFM IN'TnTIlrNTATIt1N LflIJII1LLflLU 5 AE %.LJL LNGIMLLR&D cc - Ary"atnm

'A MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE

!OF- CHANNELS 1Q0IRIE_ OPERABLE ACTIM I

2. Containment Spray (CDA) (Continued) 2 1, 2, 3, 4 14 w b. Automatic Actuation 2 1 Logic and Actuation Relays

c. Containment Pressure-High-3 4 2 3 1, 2, 3, 4 17 I

3. Containment Isolation I,0 oS

a. Phase oA' Isolation

1) Manual Initiation 2 I 2 i, 2, 3, 4 19 0 I 2 i, 2, 3, 4 14

2) Automatic Actuation 2 Logic and Actuation Relays

3) Safety Injection See Item I. above for all Safety Injection Initiating functions and requirements.

CL b. Phase "B"Isolation rt

2 1) manual Initiation "2 I with 2 I, 2, 3, 4 19 0

2 coincident switches Automatic Actuation I, 2, 3, 4 14 S2) 2 I 2 Logic and Actuation Relays

TABLE 3.3-3_(Continued)

EMIMI ED SAFETY rEATURESJATIONSY* TIEMTATI~

HINIMM TOTAL NO. CHANNELS CHANNELS APPLICABLE FU-TIOAL IlIT TO-RIP- mPEML

3. Containment Isolation (Continued) ri'

3) Containment 4 2 3 t,2, 3, 4 17 Pressure--HIgh-3

c. Purge Isolation 2 I 2 s, 6a 26 I

4. Steam Line Isolation Id

a. Kanual Initiation I/stem line I/steam line I/operating 1, 2, 3, 4 24 Q

1) Individual Id steam ine B

2) System 2 I 2 1, 2, 3, 4 23 2 I 2 t, 2, 3, 4 22

b. Astomatic Actuation Logic and Actuation Relays i c. Containment Pressure-High-2 3 2 2 1, 2, 3, 4 20

d. Steam Line Pressure- 3/steam line 2/steam line 2/steam line t, 2, 39 20 Low In each In any In each operating operating operating C loop loop loop CD 3.=

rt

e. Steam Line Pressure 3/steam line 2/steam line 2/steam line 20 -4 Negative Rate--High in each Inany In each operating operating operating cD loop loop loop

• 0

TABLE 3.3-3 (Continued) 0 s-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 0

z Ill I MINIMUM CHANNELS CHANNELS APPLICABLE TOTAL NO.

FUNCTIONAL UNIT "OF CHANNELS 1QTTLP OPRAL ACTIONf W 5. Turbine Trip and Feedwater Isolation Automatic Actuation 2 I 2 1, 2 25 a.

Logic and Actuation Relays 4/stm. gen. 2/stm. gen. 3/stm. gen. I, 2, 3 20, 21

b. Steam Generator in any oper in each Water Level- in each operating ating loop operating High-High (P-14) loop loop 2 I 2 1, 2 22

c. Safety Injection Actuation Logic

d. Tave Low Coincident with P-4 1)Four Loops operating )

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT SOF CHANNELS TO TRIP OPERABLE MODES ACTION

6. Auxiliary Feedwater

a. Manual Initiation 2 I 2 1, 2, 3 23

b. Automatic Actuation Logic 2 1 2 1, 2, 3 22 and Actuation Relays

c. Stm. Gen. Water Level-Low-Low

1) Start Motor Driven Pumps 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2, 3 20 Ut in any oper in each ating stm. operating gen. stm. gen.

2) Start Turbine Driven Pump 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2, 3 20 in any in each 2 operating operating stm. gen. stm. gen.

d. Safety Injection See Item 1. above for all Safety Injection initiating functions and Start Motor-Driven requirements.

Pumps

e. Loss-of-Offsite Power 2 1 2' 1, 2, 3 19 Start Motor-Driven Pumps

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 03

-J N

.. rI C,,

-I MINIMUM 0 TOTAL NO. CHANNELS CHANNELS APPLICABLE m FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION z I 3-4

-I 6. Auxiliary Feedwater (Continued)

f. Containment Depres See Item 2. above for all CDA functions and requirements.

surization Actuation (CDA) Start Motor-Driven Pumps

7. Control Building Isolation

a. Manual Actuation 2 1 2 19 1 1, 2, 3, 4

b. Manual Safety 2 2 19 Injection Actuation Ie 1, 2, 3, 4

c. Automatic Actuation 2 2 14 Logic and Actuation Relays

d. Containment Pressure- 3 2 2 1, 2, 3 16 High-1 CL e. Control Building Inlet 2/i ntake 1 2/i ntake

• 18 1 Ventilation Radiation 04.

8. Loss of Power

a. 4 kV Bus Under 4/bus 2/bus 3/bus 1, 2, 3, 4 20 voltage-Loss of Voltage

b. 4 kV Bus Undervoltage 0 (A)' Grid Degraded Voltage 4/bus 2/bus 3/bus 1, 2, 3, 4 20 0 0

TABLE 3.3-3 (Continued) 1 U, ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 0

z m

MINIMUM z TOTAL NO. CHANNELS CHANNELS APPLICABLE

-I FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION (A)

9. Engineering Safety Features Actuation System Interlocks

a. Pressurizer Pressure, 3 2 2 1, 2, 3 21 P-11 7

b. Low-Low Tav.g, P-12 4 2 3 1, 2, 3 21 (A)

c. Reactor Trip, P-4 2 2 2 1, 2, 3 23 0 (A) 10. Emergency Generator Load Sequencer 2 I 2 1, 2, 3, 4 14 Li-s CL z

'-4 0

0.

February 20, 2002 TABLE 3.3-3 (ContniQi1d).

TABLE NOTATIONS

1. The Steamline Isolation Logic and Safety Injection Logic for this trip function may be blocked in this MODE below the P-Il (Pressurizer Pressure Interlock) Setpoint.

• MODES 1, 2, 3, 4, 5, and 6.

During fuel movement within containment or the spent fuel pool. I

• • Trip function automatically blocked above P-li and may be blocked below P-I when Safety Injection on low steam line pressure is not blocked.

t During core alterations or movement of irradiated fuel within the containment. The provisions of Specification 3.0.3 are not applicable.

ACTION STATEMENTS ACTION 14 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; however, one channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.

ACTION 15 - (not used).

ACTION 16 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 17 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met.

• One additional channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing

-ptr-Specification 4.3.2.1.

ACTION 18 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 7 days. After 7 days, or if no channels are OPERABLE, immediately suspend CORE ALTERATIONS and fuel movement, if applicable, and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

ACTION 19 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

MILLSTONE - UNIT 3 3/4 3-24 R7, 79, F, 77, Amendment No. 203 0725

¶A-' June 27, 1996 TABLE 3.3-3 (Continued)

ACTION STATEMENTS (Continued)

ACTION 20 -With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

b. the Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.2.1.

ACTION 21 -With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> determine by observation of the associated permissive annunciator window(s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

ACTION 22- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.

ACTION 23 -With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable. channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least .HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 24 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or declare the associated valve inoperable a-fd-t-ake the ACTION required by Specification 3.7.1.5.

ACTION 25 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.

ACTION 26 - With less than the Minimum Channels OPERABLE requirement, the containment purge and exhaust valves shall be maintained closed.

Fuel movement and CORE ALTERATIONS may continue. The containment radiation monitoring channels required for containment area purge and exhaust isolation are not required to be OPERABLE during the performance of Type A containment leakage rate tests.

MILLSTONE - UNIT 3 3/4 3-25 Amendment No. 79, 129

o,-r TABLE 3.3-4 r

,-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 0

NOMINAL TRIP SETPOINT ALLOWABLE VALUE C FUNCTIONAL UNIT I

-4

1. Safety Injection (Reattor Trip, Feedwater Isolation, Control (Ai Building Isolation (Manual Initiation Only), Start Diesel Generators, and Service Water)

N.A. N.A.

a. Manual Initiation N.A. N.A.

b. Automatic Actuation Logic 17.7 psia < 17.9 psia N

(A

c. Containment Pressure--High 1 bý 0

d. Pressurizer Pressure--Low CA) 1892 psia > 1889.6 psla I8 Channels I and II 1892 psia 5 1889.6 psla Channel III and IV 658.6 psig* > 654.7 psig*

e. Steam Line Pressure--Low

2. Containment Spray (CDA)

N.A. N.A.

a. Manual Initiation N.A. N.A.

b. Automatic Actuation Logic and Actuation Relays . co 22.7 psia < 22.9 psia LO

c. Containment Pressure--High-3 0.

3. Containment Isolation

a. Phase "A" Isolation z N.A.

0 N.A.

1) Manual Initiation

-Q H

1.0

TABLE 3,3-4 o X ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS Ir r- NOMINAL 0

FUNCTIONAL UNIT TRIP SETPOINT ALLOWA BLE2VALUE

3. Containment Isolation (Continued)

2) Automatic Actuation Logic N.A. N.A

--4 and Actuation Relays

3) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

b. Phase "B" Isolation 1). Manual Initiation N.A. N.A.

2) Automatic Actuation N.A. N.A.

Logic and Actuation Re lays

%4

3) Containment Pressure- 22.7 psia S 22.9 psia High-3

c. Purge Isolation < 1 R/h < 1 R/h

4. Steam Line Isolation Ln U,

N.A. N.A.

a. Manual Initiation '_1 N.A. N.A. 0)

b. Automatic Actuation Logic co and Actuation Relays CL c. Containment Pressure--Hlgh-2 17.7 psia *_17.9 psia

d. Steam Line Pressure--Low 658.6 psig* > 654.7 psig*

Steam Line Pressure 100 psi/s** <_103.9 psi/s**

e.

04 Negative Rate--High

TABLE 3,3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETEOINTS TRIPNOMINAL SETPOINT ALLOWABLEVALUE FUNCTIONAL UNIT g-4 5. Turbine Trip and Feedwiter Isolation C..

a. Automatic ActuatiOn Logic N.A.

Actuation Relays 80.5% of narrow < 80.8% of narrow instrument

b. Steam Generator Water range instrument range Level--High-High (P-14) span. span.

Trip Safety Injection Actuation See Item 1. above for all Safety Injection

c. Setpoints and Allowable Values.

Logic

d. Tave Low Coincident with Reactor Trip (P-4) A (- _4*

.o.

1) Four Loops Operating 564-F > -3.6F ci,

2) Three Loops Operating 564"F > 563.6"F

6. Auxiliary Feedwater

a. Manual Initiation N.A. N.A.

N.A.

b. Automatic Actuation Logic ir and Actuation Relays V

c. Steam Generator Water Level--Low-Low _>17.8% of narrow

1) Start Motor-Driven 18.1% of narrow range range instrument span.

=D 0

Pumps instrument span.

TABLL .3-4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 0I r-C1 FUNCTIONAL UNIT

6. Auxiliary Feedwater (Cofitinued)

NOMINAL TRIP SETPOINT ALLOWABLE VALUE I

2) Start Turbine I 18.1% of > 17.8% of narrow Driven Pumps narrow range range instrument instrument span.

span.

d. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

e. Loss-of-Offsite Power 2800V > 2720V Start Motor-Driven Pumps

f. Containment Depressurization See Item 2. above for all CDA Trip Setpoints and Allowable Values.

U, Actuation (CDA) Start Motor-Driven Pumps U,

7. Control Building Isolation

a. Manual Actuation N.A N.A.

b. Manual Safety Injection N.A N.A.

Actuation

c. Automatic Actuation N.A. N.A.

Logic and Actuation Re ays

d. Containment 17.7 psia < 17.9 psia z Pressure--High 1 0

e. Control Building <1.5 x 10"1ci/cc <1.5 x 10"Xui/cc, Inlet Ventilation Radiation co I .

'N.

TABLE 3.3-4 (Continued) o ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS C1

-H 01 FUNCTIONAL UNIT i NOMINAL TRIP SETPOINT ALLOWABLE VALUE I

8. Loss of Power

--4 a. 4 kV Bus Undervoltage 2800 > 2720 volts volts with with a <2 w4 (Loss of Voltage) second time a < 2 second time delay. delay.

b. 4 kV. Bus Undervoltage 3730 volts > 3706 volts with a < 8 with a < 8 (Grid Degraded Voltage) second time second time delay with ESF delay with ESF actuation or actuation or

< 300 second < 300 second time delay time delay (A) without ESF without ESF actuation. actuation.

(A, C.,

0 9. Engineered Safety Features Actuation System Interlocks

a. Pressurizer Pressure, P-11 1999.7 psia

• 2002.1 psia

b. Low-Low Tavg, P-12 553*F > 552.6"F N.A. N.A.

c. Reactor Trip, P-4 N.A. N.A.

E3 10. Emergency Generator Load 01 o

Sequencer cn i

0.

p-1

INSTRUMENTATION 3/4.3.5 SHUTDOWN MARGIN MONITOR LIMITING CONDITION FOR OPERATION 3.3.5 Two channels of Shutdown Margin Monitors shall be OPERABLE

a. With a minimum count rate as designated in the CORE OPERATING LIMITS REPORT (COLR), or

b. If the minimum count rate in Specification 3.3.5.a cannot be met, then the Shutdown Margin Monitors may be made operable with a lower minimum count rate, as specified in the COLR, by borating the Reactor Coolant System above the requirements of Specification 3.1.1.1.2 or 3.1.1.2. The additional boration shall be:

1. A minimum of 150 ppm above the7SHUTDOWN MARGIN require-_*dt ,Z ments of Fi ure 3.1-I Mode 3 - 4 loops o eration an_

Figure 3.1-2 (Mode 3 - 3 loops i-n operation)), or*

2. A minimum of 350 ppm above the SHUTDOWN MARGIN require ments of Figure 3.1-3 (Mode 4), Figure 3.1-4 (Mode 5 - RCS loops filled) and Figure 3.1-5 (Mode 5 - RCS loops drained).

APPLICABILITY: MODES 3*, 4, and 5.

ACTION:

a. With one Shutdown Margin Monitor inoperable, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

b. With both Shutdown Margin Monitors inoperable or one Shutdown Margin Monitor inoperable for greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, immediately suspend all operations involving positive reactivity changes via dilution and rod withdrawal. Verify the valves listed in Specifica tion 4.1.1.2.2 are closed and secured in position within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least once per 14 days thereafter.** Verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1.2 or 3.1.1.2, as applicable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

• The shutdown margin monitors may be blocked during reactor startup in accordance with approved plant procedures.

• • The valves may be opened on an intermittent basis under administrative controls as noted in Surveillance 4.1.1.2.2.

MILLSTONE - UNIT 3 3/4 3-82 Amendment No. 10

INSTRUMENTATION OCT 2 1 1-.-

3/4.3.5 SHUTDOWN MARGIN MONITOR (continued)

SURVEILLANCE REQUI REMENTS SURVEILLANCE REQUIREMENTS 4.3.5 a. Each of the above required shutdown margin monitoring instruments shall be demonstrated OPERABLE by an ANALOG CHANNEL OPERATIONAL TEST at least once per 92 days that shall include verification that the Shutdown Margin Monitor is set per the Core Operating Limits Report (COLR).

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> VERIFY the minimum count rate (counts/sec) as defined within the COLR.

"_IIITT I *IA - Amendment No. 164 MTI I CTflMI "1 * " v,,

0495

3/4.4

• REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER JLQ.,,.TIQN jey-ei-+soeQ I1MITTNG CflNflITOflN FAR OPFRATTON 3.4.1.1 Either:

a) 11 reactor coolant loops shall be 1Ain operation.

b) Three reactor coolant loops shall be in operation with THERMAL POWER restricted to less than or equal to 65% of RATED THERMAL POWER.

APPLICABILITY: MODES I and 2.*

ACTION:

With less than the above required reactor coolant loops in operation, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SIIRVFTI ANCTF RFQITRFMFNTS 4.4.1.1 The above required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• See Special Test Exceptions Specification 3.10.4.

MILLSTONE - UNIT 3 3/4 4-1

REACTOR COOLANT SYSTEM LOOP

-OAE L- PSW IA-V "

J ry 81, i **

LIMITING CONDITION FOR OPERATION 3.4.1.5 RCS loop stop valveo an,isolated loop shall be shu and the pover removed. from the valve operato s APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION: vlf R Wihth equirements of the above specification not satisfied: either shut theJ Sloop stop valves and remove power from the valve operators within one hour, or be/

Sin at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the*

foll1owing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />,.j SURVEILLANCE REQUIREMENTS 4.4.1.5 The RCS loop stop valves of an isolated loop shall be verified shut and ower removed from the valve operators at least once per 31 days.

xl 4cs oypA@ VCV %% op cri QAC-C iimcC -n~ANL\jAA ýN-_k n MILLSTONE - UNIT 3 3/4 4-7

Insert B to Page 314 4-7

a. With power available to one or more loop stop valve operators, remove power from the loop stop valve operators within 30 minutes or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.ý') With one or more RCS loop stop valves closed, maintain the valve(s) closed and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Insert C to Page 314 4-7 (1) All required actions of Action Statement 3.4.1.5.b. shall be completed whenever this action is entered.

REACTOR COOLANT SYSTEM ISOLATED LOOP STARTUP January 9, 2002 LIMITING CONDITION FOR OPERATION 3.4.1.6 A reactorcoolant loop shall remain isolated with power removed from the associated RCS loop stop valve operators until:

a. The temperature at the cold leg of the isolated loop is within 20°F of the highest cold leg temperature of the operating loops, and I

b. The boron concentration of the isolated loop is greater than or equal to the boron concentration required by Specifications 3.1.1.1.2 or-.

3.1.1.2 for MODE 5 or Specification 3.9.1.1 for MODE 6.

APPLICABILITY: MODES 5 and 6.

ACTION:

a. With the requirements of the above specification not satisfied, do not open the isolated loop stop valves.

SURVEILLANCE REQUIREMENTS 4.4.1.6.1 The isolated loop cold leg temperature shall be determined to be within 20°F of the highest cold leg temperature of the operating loops within 30 minutes prior to opening the cold leg stop valve.

4.4.1.6.2 The isolated loop boron concentration shall be determined to be greater than or equal to the boron concentration required by Specifications 3.1.1.1.2 or 3.1.1.2 for MODE 5 or Specification 3.9.1.1 for MODE 6 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to opening the hot or cold leg stop valve.

I MILLSTONE - UNIT 3 3/4 4-8 Amendment No. 77, ý7, PY, 7 7, 0776 202,

August 27, 2001 High Setpoiin PORV Curve For the Cold Overpressure Protection System to (3L

'a E

0 z

0 50 100 150 200 250 Auctioneered Low Measure RCS Temperature (F)

FIGURE 3.4-4a MILLSTONE - UNIT 3 3/4 4-40 Amendment No. fp, 7P7, 197

Z~c~bC)r August 27, 2001 Low Setpoint PORV Curve For the Cold Overpressure Protection System 800 700 600 500

)400 0

IL.

-E 0

z 300 100 0

0 5o 100 150 200 250 Auctioneered Low Measure ROCSTemperature (F)

FIGURE 3.4-4b MILLSTONE - UNIT 3 3/4 4-41 Amendment No. fq, 0Jl, 197

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION IF-3.7.1.1 All main steam line Code safety valvesi ssoi~at w-ith ph- eta,--.

ated reac*or coo°ant. *l°00 shall be OPERABLE with lift egeneraf)r ' a APPLICABILITY: MODES 1, 2, and 3.

ACTIC

a. With four reactor coolant loops and associated steam generatorsj"L)

ýr'erration and withfofne or- more main steam line Code safety valves inoperable, operation in MODES 1, 2, and 3 may 'proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY fi hours. the next within hourp, and in HOT SHUTDOWN within the following o 6 AQ(Ž'

b. With three reactor coolant loops and associated steam generators in operation and with one or more main steam line Code safety valves associated with an operating loop inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the Range inoperable valve is restored to OPERABLE status or the Power Neutron Flux High TripfollowingSetpoint is reduced per Table 3.7-2; 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

6HOT otherwise, in at the SHUDOWNbe within least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5.

AMENDMENT NO. ft MILLSTONE - UNIT 3 3/4 7-1

Januar-31* 99k, MAXIMUM ALLOWABLE POWER RANGE NP N SAFETY V _ _ _ _ _ _ _ _T OP MAXIMUM ALLOWABLE POWER RANGE MAXIMUM NUMBER OF INOPERABLE NEUTRON FLUX HIGH SETPOINT SAFETY VALVES ON ANY (PERCENT OF RATED THERMAL POWER)

OPERATING STEAM GENERATOR 1 65 46 2

28 3

• • I~MU AXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH.,SEPOINT WITH-'

INOPERABLE STEAM LINE SAFETY VALVES DURING THREE LOOP OPjERW&TIOLN MAXIMUM HNDMER OF INOPERABLE MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SEIPOINT OPERATING STEAM GENERATOR* (PERCENT OF RATED THERMAL POWER) 1 47 2 33 3 19 non-operating steam

• At least two safety valves shall be OPERABLE on the generator.

I 3/4 7-2 Amendment No.

MILLSTONE - UNIT 3

March 17, 1995 TABLE 3.7-STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER LIFT SETTING* (+/-3%-** ORIFIE SI~-ZE I 1185 psig 16.0 square square inches inches RV22A psig 16.0 1195 16.0 square inches RV23A 1205 psig RV24A 16.0 square inches RV25A 1215 psig 16.0 square inches 1225 psig RV26A iQLOO inches 1185 psig 16.0 square square inches RV22B psig 16.0 1195 16.0 square inches RV23B 1205 psIg RV24B 16.0 square inches 1215 psig square inches RV25B psig 16.0 1225 RV26B LOOP 3 inches RV22C 1185 1195 psig psig 16.0 16.0 square square inches RV23C 16.0 square inches 1205 psig 16.0 square inches RV24C 1215 psig RV25C 16.0 square inches 1225 psig RV26C LOOP 4 inches psig 16.0 RV22D 1185 psig 16.0 square inches 1195 16.0 square inches RV23D 1205 psig RV24D psig 16.0 square inches 1215 psig 16.0 square inches RV25D 1225 RV26D conditions of the valve

• The lift setting pressure shall correspond to ambient at nominal operating temperature and pressure.

main steam line Code safety

• • The lift setting shall be within +/- 1% following valve testing.

Amendment No. 106 MILLSTONE - UINIT 3 3/4 7-3 0276

SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION. AND POWER DISTRIBUTION LIMITS jjUR LOOPS OPERATING 1ý LIMITING CONDITION FOR OPERATION 3.10.2.1 The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

- a. The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and

b. The limits of Specifications 3.2.2.1 and 3.2.3.1 are maintained and determined at the frequencies specified in Specification 4.10.2.1.2 below.

APPLICABILITY: MODE 1.

ACTION:

With any of the limits of Specification 3.2.2.1 or 3.2.3.1 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1.1, and 3.2.4 are suspended, either:

a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2.1 and 3.2.3.1, or

b. Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.10.2.1.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.2.1.2 The Surveillance Requirements of the below listed specifications shall be performed at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during PHYSICS TESTS:

a. Specifications 4.2.2.1.2 and 4.2.2.1.3, and

b. Specification 4.2.3.1.2.

MILLSTONE - UNIT 3 3/4 10-2

SPECIAL TEST EXCEPTIONS JAN~ 3 1 S GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS THREE LOOPS OPERATING LIMITING CONDITION FOR OPERATION limits of 3.10.2.2 The group height, insertion, and power distribution may be suspended 3.2.4 Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1.2, and during the performance of PHYSICS TESTS provided:

55% of RATED

a. The THERMAL POWER is maintained less than or equal to THERMAL POWER, and maintained

b. The limits of Specifications 3.2.2.2 and 3.2.3.2 are and determined at the frequencies specified in Specification 4.10.2.2.2 below.

APPLICABILITY: MODE 1.

ACTION:

3.2.3.2 being exceeded while With any of the limits of Specification 3.2.2.2 or 3.1.3.5, 3.1.3.6, 3.2.1.2, and 3.2.4 the requirements of Specifications 3.1.3.1, are suspended, either:

..n,,n...... ...4... ++ tstisfv n the ACTION requirements L 32UW.tSL.

2 and 3.22, o ions 3.2.2.2 and 3.2.3.2, or of Specificat Sa. b. Reduce IHLKMM, Be in HOT STAINDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS to be less than or equal to 4.10.2.2.1 The THERMAL POWER shall be determined during PHYSICS TESTS.

hour 55% of RATED THERMAL POWER at least once per of the below listed specifications 4.10.2.2.2 The Surveillance Requirements 12 hours during PHYSICS TESTS:

shall be performed at least once per and

a. Specifications 4.2.2.2.2 and 4.2.2.2.3,

b. Specification 4.2.3.2.2.

3/4 10-3 MILLSTONE - UNIT 3

POWER DISTRIBUTION LIMITS AXIAL FLUX DIFFERENCE (Continued)

(2) APLND (for base load operation). Penalty deviation minutes for base load operation are not accumulated based on the short period of time during which operation outside of the target band is allowed.

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR

" " The limits on heat flux hot channel factor, RCS flow rate, and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200"F ECCS acceptance criteria limit.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to ensure that the limits-are maintained provided:

a. Control rods in a single group move together with no individual rod insertion differing by more than +12 steps, indicated, from the group demand position;

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6;

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and

d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

FN&H will be maintained within its limits provided Conditions a. through

d. above are maintained. The relaxation of FN. as a function of THERMAL POWER allows changes in the radial power shape for all permssibl,uo nsertion limits.

The FN&H as calculated in Specificatior 3.2.3.1 a .. 2 are sed in the various accident analyses where FN&H influences parameters o er than DNBR, e.g., peak clad temperature, and thus is the maximum "as measured" value allowed.

MILLSTONE - UNIT 3 B 3/4 2-3 Amendment No. 60, 0

POWER DISTRIBUTION LIMITS RASFS 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

Margin is maintained between the safety analysis limit DNBR and design limit DNBR. the This margin is more than sufficient to offset any rod bow penalty and transition core penalty. The remaining margin is available for plant design flexibility.

When an FQ measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3%.-allowance is appropriate for manufacturing tolerance.

The heat flux hot channel factor, F,(Z), is measured periodically using the incore detector system. These measurements are generally taken with the core at or near steady state conditions. Using the measured three dimensional power distributions, it is possible to derive FQM(Z), a computed value of FQ(Z).

However, because this value represents a steady state condition, ITt does not include the variations in the value of Fo(Z) that are present during nonequilibrium situations.

To account for these possible variations, the steady state limit of F0 (Z) is adjusted by an elevation dependent factor appropriate to either load operation, W(Z) or W(Z)BL, that accounts for the calculated RAOC or base worst case transient conditions. The W(Z) and W(Z)B, factors described above for normal operation are specified in the COLR per Specification 6.9.1.6.

Core monitoring and control under nonsteady state conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and control ro.d insertion. Evaluation of the steady state F,(Z) limit is performed in Specification 4.2.2.1.2.b and 4.2.2.1.4.b while evaluation nonequilibrium limits are performed in Specification 4.2.2.1.2.c and 4.2.2.1.4.c.

When RCS flow rate and FN m are measured, no additional allowances necessary prior to comparison with the limits of the Limiting are Cgndition for O aion. Measurement errors of 2.4% four loop w and 2.8% for three7 o - _

flowpfor RCS total flow rate and 4% for FNw have been allowed for in deterjint fthe design DNBR value.

The measurement error for RCS total flow rate is based upon performing precision heat balance and using the result to calibrate a the RCS flow rate indicators. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non conservative manner. Therefore, a penalty of 0.1% for undetected the feedwater venturi will be added if venturis are not inspected fouling of and cleaned at least once for 18 months. Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by.monitoring and trending various plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

. MILLSTONE - UNIT 3 B 3/4 2-4 Amendment No. 17, yo, J~fi

)Iaveh 11-, 199 POWER DISTRIBUTION LIMITS BASES RISE HOT HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY CHANNEL FACTOR (Continued) to The 12-hour periodic surveillance of indicated RCS flow is sufficient detect only flow degradation which could lead to operation outside the accept able-region of operation defined in Specification3.2.3-I~nd 3-3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during power operation.

The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in F is depleted. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.

The 2-hour time allowance for- operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc tion of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normali-ed .symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.

3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain each analyzed a minimum transient. DNBR greater than the design limit Nhroug.put The indicated Tavg value of 591.I'F (f*_urloo4 MILýONE - UNIT 3 B 3/4 2-5 Amendment No. 77, 50,0

Ma 3 POWER DISTRIBUTION LIMITS BASES DNB PARAMETERS (Continued) operations or 583.3"F th loops operating) and the i ted ressurizer or three loop operation) The pressure value is 2218 psia (four loo calculated values of the DNB related parameters will be an average of the indicated values for the operable channels.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

Measurement uncertainties have been accounted for in determining the parameter limits.

- .~-.

MIL$ONE - UNIT 3 B 3/4 2-6 Amendment No. r;z,60 I

INSTRUMENTATION BASES 3/4 3.5 SHUTDOWN MARGIN MONITOR The Shutdown Margin Monitors provide an alarm tha a Boron Dilution Event may be in progress. The minimum count rate of Specif cation 3/4.3.5 and the SHUTDOWN MARGIN requirements of Figures 3.1-1, 1- 3.1-3, 3.1-4, and 3.1-5 ensure that at least 15 minutes are available for operator action from the time of the Shutdown Margin Monitor alarm to total loss of shutdown margin.

By bor additional 150 ppm above the SHUTDOWN MARGIN required by Figure 3.1-1 r 3.1-2) or 350 ppm above the SHUTDOWN MARGIN required by Figure 3.1-3, 3.1-4, or . - lower alues of minimum count rate are accepted.

Shutdown Marcin Monitors

Background:

The purpose of the Shutdown Margin Monitors (SMM) is to annunciate an increase in core subcritical multiplication allowing the operator at least 15 minutes of response time to mitigate the consequences of the inadvertent addition unborated primary grade water (boron dilution event) into the Reactor Coolant System (RCS) when the reactor is shut down (Modes 3, 4, and 5).

The SMMs utilizes two channels of source range instrumentation (GM detectors).

Each channel provides a signal to its applicable train of SMM. The SMM channel uses the last 600 or more counts to calculate the count rate and updates the measurement after 30 new counts or I second, whichever is longer.

that hold the counts (20 registers X 30 count Each channel has 20 registers 600 counts) for averaging the rate. As the count rate decreases, the longer it takes to fill the registers (fill the 30 count minimum). As the instrument's measured count rate decreases, the delay time in the instrument's response increases. This delay time leads to the requirement of a minimum count rate for OPERABILITY.

the During the dilution event, count rate will increase to a level above normal steady state count rate. When this new count rate level increases operator above the instrument's setpoint, the channel will alarm alerting the of the event.

Applicable Safety Analysis and The SMM senses abnormal increases in the source range count per second This alarm will occur alarms the operator of an inadvertent dilution event.

at least 15 minutes prior to the reactor achieving criticality. This the 15 minute window allows adequate operator response time to terminate dilution, FSAR Section 15.4.6.

LCO of LCO 3.3.5 provides the requirements for OPERABILITY of the instrumentation dilution event. Two trains are the SMMs that are used to mitigate the boron required to be OPERABLE to provide protection against single failure.

MILLSTONE - UNIT 3 B 3/4 3-7 Amendment Ho.

0498ý>

3/4.4 REACTOR COOLANT SYSTEM A-., *_,s t 2 7. 20 01!-

BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Tela F i'sdesignedd to operate in MODES I and 2 with three or fourl eator coolant loopsop inn operation plan and maintain DNBR greater than the desi n/

lmt during all normal operations and anticipated transients~jWith less than the required reactor coolant loops in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

In MODE 3, three reactor coolant loops, and in Mode 4, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, in MODE 3 a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., the Control Rod Drive System(I\

is not capable of rod withdrawal.

In MODE 4, if a bank withdrawal accident can be prevented, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (any combination of RHR or RCS) be OPERABLE.

In MODE 5, with reactor coolant loops filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two RHR loops or at least one RHR loop and two steam generators be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE.

In MODE 5, during a planned heatup to MODE 4 with all RHR loops removed from operation, an RCS loop, OPERABLE and in operation, meets the requirements of an OPERABLE and operating RHR loop to circulate reactor coolant. During the heatup there is no requirement for heat removal capability so the OPERABLE and operating RCS loop meets all of the required functions for the heatup condition. Since failure of the RCS loop, which is OPERABLE and operating, could also cause the associated steam generator to be inoperable, the associated steam generator cannot be used as one of the steam generators used to meet the requirement of LCO 3.4.1.4.1.b.

The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The restrictions on starting the first RCP in MODE 4 below the cold overpressure protection enable temperature (226°F), and in MODE 5 are provided to prevent RCS pressure transients. These transients, energy additions due to the differential temperature between the steam generator secondary side and the RCS, can result in pressure excursions which could challenge the P/T limits.

The RCS will be protected against overpressure transients and will not exceed the reactor vessel isothermal beltline P/T limit by restricting RCP starts based on the MILLSTONE - UNIT 3 B 3/4 4-1 Amendment No- 7, f, 97, XX7,

Insert D to Page B 314 4-1 The purpose of Specification 3.4.1.1 is to require adequate forced flow rate for core heat removal in MODES 1 and 2 during all normal operations and anticipated transients. Flow is represented by the number of reactor coolant pumps in operation for removal of heat by the steam generators. To meet safety analysis acceptance criteria for DNB, four reactor coolant pumps are required at rated power. An OPERABLE reactor coolant loop consists of an OPERABLE reactor coolant pump in operation providing forced flow for heat transport and an OPERABLE steam generator in accordance with Specification 3.4.5.

3/4.4 REACTOR COOLANT SYSTEM BASES (Continued) differential water temperature between the secondary side of each steam generator and the RCS cold legs. The restrictions on starting the first RCP only apply to RCPs in RCS loops that are not isolated. The restoration'of isolated RCS loops is normally accomplished with all RCPs secured. If an isolated RCS loop is to be restored when an RCP is operating, the appropriate temperature differential limit between the secondary side of the isolated loop' steam generator and the in'service RCS cold legs is'applicable, and shall'be met prior to opening the loop isolation valves.

The temperature differential limit between the secondary side of the steam generators and the RCS cold legs is based on the equipment providing cold overpressure protection as required by Technical Specification 3.4.9.3. If the pressurizer PORVs are providing cold overpressure protection, the steam generator secondary to RCS cold leg water temperature differential is limited to a maximum of 50F. If any RHR, relief valve is providing cold overpressure protection and RCS cold leg temperature, is* above 150"F, the' steam generator secondary water temperature must be at or below RCS cold leg water temperature.

If any RHR relief valve is providing cold overpressure protection and RCS cold leg temperature is at or below 150"F, the steam generator secondary to RCS cold le water temperature differential is limited to a maximum of 50"F.

e requirement to maintain the isolated loop stop valves shut ith power removed ensures that no reactivity addition to the core could occur~pe to the startup of an isolated loop. Verification of the boron concentration in an isolated loop prior to opening the first stop valve provides a reassurance of the adequacy of the boron concentration in the isolated loop.

MILLSTONE - UNIT 3 B 3/4 4-1a Amendment No. 7

Insert E to Page B 314 4-1a (Page 1 of 2)

Specification 3.4.1.5 The reactor coolant loops are equipped with loop stop valves that permit any loop to be isolated from the reactor vessel. One valve is installed on each hot leg and one on each cold leg. The loop stop valves are used to perform maintenance on an isolated loop. Operation in MODES 1-4 with a RCS loop stop valve closed is not permitted except for the mitigation of emergency or abnormal events. If a loop stop valve is closed for any reason, the required actions of this specification must be completed. To ensure that inadvertent closure of a loop stop valve does not occur, the valves must be open with power to the valve operators removed in MODES 1, 2, 3 and 4.

The safety analyses performed for the reactor at power assume that all reactor coolant loops are initially in operation and the loop stop valves are open. This LCO places controls on the loop stop valves to ensure that the valves are not inadvertently closed in MODES 1, 2, 3 and 4. The inadvertent closure of a loop stop valve when the Reactor Coolant Pumps (RCPs) are operating will result in a partial loss of forced reactor coolant flow. If the reactor is at rated power at the time of the event, the effect of the partial loss of forced coolant flow is a rapid increase in the coolant temperature which could result in DNB with subsequent fuel damage if the reactor is not tripped by the Low Flow reactor trip. If the reactor is shutdown and a RCS loop is in operation removing decay heat, closure of the loop stop valve associated with the operating loop could also result in increasing coolant temperature and the possibility of fuel damage.

The loop stop valves have motor operators. If power is inadvertently restored to one or more loop stop valve operators, the potential exists for accidental closure of the affected loop stop valve(s) and the partial loss of forced reactor coolant flow. With power applied to a valve operator, only the interlocks prevent the valve from being operated. Although operating procedures and interlocks make the occurrence of this event unlikely, the prudent action is to remove power from the loop stop valve operators. The time period of 30 minutes to remove power from the loop stop valve operators is sufficient considering the complexity of the task.

Should a loop stop valve be closed in MODES 1 through 4, the affected valve must be maintained closed and the plant placed in MODE 5. Once in MODE 5, the isolated loop may be started in a controlled manner in accordance with LCO 3.4.1.6, "Reactor Coolant System Isolated Loop Startup." Opening the closed loop stop valve in MODES 1 through 4 could result in colder water or water at a lower boron concentration being mixed with the operating RCS loops resulting in positive reactivity insertion. The time period provided in Action 3.4.1.5.b. allows time for borating the operating loops to a shutdown boration level such that the plant can be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The allowed action times are reasonable, based on

Insert E to Page B 314 4-1a (Page 2 of 2) operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

Surveillance Requirement 4.4.1.5 is performed at least once per 31 days to ensure that the RCS loop stop valves are open, with power removed from the loop stop valve operators. The primary function of this Surveillance is to ensure that power is removed from the valve operators, since Surveillance Requirement 4.4.1.1 requires verification every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that all loops are operating and circulating reactor coolant, thereby ensuring that the loop stop valves are open. The frequency of 31 days ensures that the required flow is available, is based on engineering judgment, and has proven to be acceptable. Operating experience has shown that the failure rate is so low that the 31 day frequency is justified.

Specification 3.4.1.6

REACTOR COOLANT SYSTEM August 27, 2001 BASES SPECIFIC ACTIVITY (Continued)

Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> between sample taking and completing the initial analysis is based upon a typical time necessary to perform the sampling, transport the sample, and perform the analysis of about 90 minutes. After 90 minutes, the gross count should be made in a reproducible geometry of sample and counter having reproducible beta or gamma self-shielding.

properties. The counter should be reset to a reproducible efficiency versus energy. It is not necessary to identify specific nuclides. The radiochemical determination of nuclides should be based on multiple counting of the sample within typical counting basis following ýampling of less than I hour, about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, about I day, about 1 week, and about I month.

Reducing Tavg to less than 500°F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves.

The Surveillance Requirements provide adequate assurance'that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM (EXCEPT THE PRESSURIZER)

BACKGROUND All c6mponents of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads-are introduced by' startup (heatup) and shutdown (cooldown) operations, power transients, aid reactor trips. This LCO limits the pressure and temperature changes during RCSý'

heatup and 6b6Tdown, within the design assumptions and the stress limits for cyclic operation".

Figures 3.4-2 and 3.4-3 contain P/T limit curves for heatup; cooldown, I inservice leak and hydrostatic (ISLH) testing, and data for the maximum rate'of change of reactor coolant temperature. 1" Each P/T limit curve defines an acceptable region for normal operation. The usual Use of the curves is operational requirements during heatup or.cooldown maneuvering, when pressure and temperature indications are monitored and compared' to the applicable curve to determine that opdration is within the allowable region. A heatup or cooldown is defined as a temperature increase 'or decrease of greater than or equal to IO0F in any one hour period. This definition of heatup and cooldown is based upon the ASME definition of isothermal conditions described in ASME,Section XI, Appendix E.

MILLSTONE - UNIT 3 B 3/4 4-7 Amendment No. 797, 197

REACTOR BAE COOLANT SYSTEM cl--- , OA 2 2001 August 27, BASES PRESSURE/TEMPERATURE LIMITS (continued)

Steady state thermal conditions exist when temperature increases or decreases are <10F in any one hour period and when the plant is not performing a planned heatup or cooldown in accordance with a procedure.

The LCO establishes operating limits that provide a margin to brittle failure, applicable to the ferritic material of the reactor coolant pressure I boundary (RCPB). The vessel is the component most subject to brittle failure, and the LCO limits apply mainly to the vessel. The limits do not apply to the Pressurizer, which has different design characteristics and operating functions which are addressed by LCO 3.4.9.2, "Pressurizer".

The P/T limits have been established for the ferritic materials of the RCS considering ASME Boiler and Pressure Vessel Code- Section XI, -Appendix G (Reference 1) as modified by ASME Code Case N-640 (Reference 2), and the additional requirements of 10 CFR 50 Appendix G (Reference 3). Implementation of the specific requirements provide adequate margin to brittle fracture of ferritic materials during normal operation, anticipated operational occurrences, and system leak and hydrostatic tests.

The neutron embrittlemer't effect on the material toughness is reflected by increasing the nil ductility reference temperature (RTNDT) as exposure to neutron fluence increases.

The actual shift in the RTNDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 4) and Appendix H of 10 CFR 50 I (Ref. 5). The operating P/T limit curves will be adjusted, as necessary, based I on the evaluation findings and the recommendations of Regulatory Guide 1.99 (Ref. 6).

The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the.feactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations mi"yH more restrictive, and thus, the curves are composites of the most restrictive regions.

• The heatup curve represents a different set of restrictions than the cooldown curve because the directions bf the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.

The P/T limits include uncertainty margins to ensure that the calculated limits are not inadvertently exceeded. These margins include gauge and system loop uncertainties, elevation differences, containment pressure conditions and system pressure drops between the beltline region of the vessel and the pressure gauge or relief valve location.

MILLSTONE - UNIT 3 B 3/4 4-8 Amendment No. 09, W7, 197

REACTOR COOLANT SYSTEM Augus 2001 BASES PRESSURE/TEMPERATURE LIMITS (continued)

The criticality limit curve includes the Reference I requirement that it be

> 40*F above the heatup curve or the cooldown curve, and not less than 160°F above the minimum permissible temperature for ISLH testing. This limit provides the required margin relative to brittle fracture. However, the criticality curve is not operationally limiting; a more restrictive limit exists in LCO 3.1.1.4, "Minimum Temperature for Criticality."

The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the ferritic RCPB materials, possibly leading to a nonisolable leak or loss of coolant- accident. I In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components. The ASME Code,Section XI, Appendix E (Ref. 7) provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.

APPLICABLE SAFETY ANALYSIS The P/T limits are not derived from Design Basis Accident (DBA) analyses.

They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, an unanalyzed condition. Reference 1, as modified by Reference 2, combined with the additional requirements of Reference 3 provide the methodology for determining I the P/T limits. Although the P/T limits are not derived from any DBA, the P/T limits are acceptance limits since they preclude operation in an unanalyzed condition.

RCS P/T limits satisfy Criterion 2 of 10CFR50.36(c)(2)(ii).

LCO The LCO limits apply to the ferritic components of the RCS, except the Pressurizer.,.Vjiese limits define allowable operating regions while providing margin against nonductile failure for the controlling ferritic component.

The limitations imposed on the rate of change of temperature have been established to ensure consistency with the resultant heatup, cooldown, and ISLH testing P/T limit curves. These limits control the thermal gradients (stresses) within the reactor vessel beltline .(the limiting component). Note that white these limits are to provide protection to ferritic components within the reactor coolant pressure boundary, a limit of 100°F/hr applies to the reactor coolant pressure boundary (except the pressurizer) to ensure that operation is maintained within the ASME Section III design loadings, stresses, and fatigue analyses for heatup and cooldown.

MILLSTONE - UNIT 3 B 3/4 4-9 Amendment No. 7X7, 197

REACTOR COOLANT SYSTEM fto 2r2OP BASES PRESSURE/TEMPERATURE LIMITS (continued)

Violating the LCO limits places the reactor vessel outside of the bounds of the analyses and can increase stresses in other RCPB components. The consequences depend on several factors, as follows:

a. The severity of the departure from the allowable operating P/iT regime or the severity of the rate of change of. temperature;

b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and

c. The existences, sizes, and orientations of flaws in the vessel material.

APPLICABILITY The RCS P/T limits LCO provides a definition of acceptable operation for prevention of nonductile failure of ferritic RCS components using ASME Section XI, II Appendix G, as modified by Code Case N-640 and the additional requirements of 10CFR50, Appendix G (Ref. 1). The P/T limits were developed to provide requirements for operation during heatup or cooldown (MODES 3, 4, and 5) or ISLH testing, in keeping with the concern for nonductile failure. The limits do not apply to the Pressurizer. D* c Z3 During MODES I and 2, other Technical 'pecifications provid limits for operation that can be more restrictive than or/can supplement thes /T limits. LCO 3.2.5, "DNB Parameters"; LCO 3.2.3.15and 3.2.3.2) "RCS Flo ate and Nuclear Enthalpy Rise Hot Channel Factor - 0 erating/Three 00opsOperatingi';

LC03.1.1.4, "Minimum Temperature'for Criticality"; and Safety Limit 2.1, "Safety Limits," also provide operational restrictions for pressure and temperature and maximum pressure. Furthermore, MODES I and 2 are above the temperature range of concern for nonductile failure, and stress analyses have been performed for normal maneuvering profiles, such as power ascension or descent.

ACTIONS Operation outside the P/T limits must be corrected so that the RCPB is returhed to a condition that has been verified by stress analyses. The Allowed Outage Times (AOTs) reflects the urgency of restoring the parameters to within ,the I 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't'odetermine 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 comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.

MILLSTONE - UNIT 3 B 3/4 4-10 Amendment No. 0.7, 10

DFATAD CflflIAJT ZYTFM BASES August 27, 2001 OVERPRESSURE PROTECTION SYSTEMS (continued)

The use of a PORV for Cold Overpressure Protection is limited to those conditions when no more than one RCS loop is isolated from the reactor vessel. When two or I more loops are isolated, Cold Overpressure Protection must be provided by either the two RHR suction relief valves or a depressurized and vented RCS.

The reactor vessel material is less tough at low temperatures than at normal operating temperature. As the vessel neutron exposure accumulates, the material toughness decreases and becomes less resistant to stress at low temperatures I (Ref. 3). RCS pressure, therefore, is maintained low at low temperatures and is increased only as temperature is increased.

The potential for vessel overpressurization is most acute when the RCS is water solid, occurring while shutdown; a pressure fluctuation can occur more quickly than an operator can react to relieve the condition. Exceeding the RCS P/T limits by a significant amount could cause nonductile cracking of the reactor vessel. LCO 3.4.9.1, "Pressure/Temperature Limits - Reactor Coolant System,"

requires administrative control of RCS pressure and temperature during heatup and cooldown to prevent exceeding the limits provided in Figures 3.4-2 and 3.4-3.

This LCO provides RCS overpressure protection by limiting mass input capability and requiring adequate pressure relief capacity. Limiting mass input capability requires all Safety Injection (SIH) pumps and all but one centrifugal charging pump to be incapable of injection into the RCS. The pressure relief capacity requires either two redundant relief valves or a depressurized RCS and an RCS vent of sufficient size. One relief valve or the open RCS vent is the overpressure protection device that acts to terminate an increasing pressure event.

With minimum mass input capability, the ability to provide core coolant addition is restricted. 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 Cold Overpressure Protection MODES and the expected core decay heat levels, the makeup system can provide adequate flow via the makeup control valve.

If a loss of RCS inventory or reduction in shutdown margin event occurs, the appropriate response will be to correct the situation by starting RCS makeup pumps. If the loss of inventory or shutdown margin is significant, this may necessitate the use of additional RCS makeup pumps that are being maintained not capable of injecting into the RCS in accordance with Technical Specification 3.4.9.3.- The use of these additional pumps to restore RCS inventory or shutdown margin will require entry into the associated action statement. The action statement requires immediate action to comply with the specification. The restoration of RCS inventory or shutdown margin can be considered to be part of the immediate action to restore the additional RCS makeup pumps to a not capable of injecting status. While recovering RCS inventory or shutdown margin, RCS pressure will be maintained below the P/T limits. After RCS inventory or shutdown margin has been restored, the additional pumps should be immediately made not capable of-injecting and the action statement exited.

MILLSTONE - UNIT 3 B 3/4 4-16 Amendment No. f?, F9, 779, 707, 197

REACTOR COOLANT SYSTEM - August 27, 2001 BASES OVERPRESSURE PROTECTION SYSTEMS (continued)

PORV Requirements As designed, the PORV Cold Overpressure Protection (COPPS) is signaled to open if the RCS pressure approaches a limit determined by the COPPS actuation logic.

The COPPS actuation logic monitors both RCS temperature and RCS pressure and determines when the nominal setpoint of Figure 3.4-4a or Figure 3.4-4b is approached. The wide range RCS temperature indications are auctioneered to select the lowest temperature signal.

The lowest temperature signal is processed through a function generator that calculates a pressure setpoint for that temperature. The calculated pressure setpoint is then compared with RCS pressure measured by a wide range pressure I channel. If the measured pressure meets or exceeds the calculated value, a PORV I is signaled to open.

The use of the PORVs is restricted to three and four RCS loops unisolated: for a loop to be considered isolated, both RCS loop stop valves must be closed. If more than one loop is isolated, then the PORVs must have their block valves closed or COPPS must be blocked. For these cases, Cold Overpressure Protection must be provided by either the two RHR suction relief valves or a depressurized RCS and an RCS vent. This is necessary because-the PORV mass and heat injection transients have only been analyzed for a maximum of one loop isolated, the use of the PORVs is restricted to three and four RCS loops unisolated.

The RHR suction relief valves have been qualified for all mass injection transients for any combination of isolated loops. In addition, the heat injection transients not prohibited by the Technical Specifications have also been considered in the qualification of the RHR suction relief valves.

Figure 3.4-4a and Figure 3.4-4b present the PORV setpoints for COPPS. The setpoints are staggered so only one valve opens during ".a low temperature I overpressure transient. Setting both valves to the values of Figure 3.4-4a and Figure 3.4-4b within the tolerance allowed for the calibration accuracy, ensures that the iso-6ti'emal P/T limits will not be exceeded for the analyzed isothermal I events.

When a PORV is opened, the release of coolant will cause the pressure increase to slow and reverse. As the PORV releases coolant, the RCS pressure decreases until a reset pressure is reached and the valve is signaled to close. The pressure continues to decrease below the reset pressure as the valve closes.

MILLSTONE - UNIT 3 B 3/4 4-16a Amendment No. 0, FY, 797, 197

REACTOR COOLANT SYSTEM 04n) August 27, 2001 BASES OVERPRESSURE PROTECTION SYSTEMS (continued)

APPLICABLE SAFETY ANALYSIS Safety analyses (Ref. 5) demonrstrate that the reactor vessel is adequately I protected against exceeding the P/T limits for the analyzed isothermal events.

In MODES 1, 2, AND 3, and in MODE 4, with RCS cold leg temperature exceeding 226°F, the pressurizer safety valves will provide RCS overpressure protection in I the ductile region. At 226°F and below, overpressure prevention is provided by I two means: (1) two OPERABLE relief valves, or (2) a depressurized RCS with a sufficiently sized RCS vent, consistent with ASME Section XI, Appendix G for I temperatures less than RTNDT + 50"F. Each of these means has a limited I overpressure relief capability.

The required RCS temperature for a given pressure increases as the reactor vessel material toughness decreases due to neutron embrittlement. Each time the Technical Specification curves are revised, the cold overpressure protection must be re-evaluated to ensure its functional requirements continue to be met using the RCS relief valve method or the depressurized and vented RCS condition.

Transients capable of overpressurizing the RCS are categorized as either mass or heat input transients, examples of which follow:

Mass Input Transients

a. Inadvertent safety injection; or

b. Charging/letdown flow mismatch Heat Input Transients

a. Inadvertent actuation of Pressurizer heaters;

b. Loss of RHR cooling; or

c. Reactor coolant pump (RCP) startup with temperature asymmetry within the RCS or between the RCS and steam generators.

The Technical Specifications ensure that mass input transients beyond the operability of the cold overpressure protection means do not occur by renderipg all Safety Injection Pumps and all but one centrifugal charging pump incapable of injecting into the RCS whenever any RCS cold leg is < 226°F.

The Technical Specifications ensure that energy addition transients beyond the operability of the cold overpressure protection means do not occur by limiting reactor coolant pump starts. LCO 3.4.1.4.1, "Reactor Coolant Loops and Coolant Circulation - Cold Shutdown - Loops Filled," LCO 3.4.1.4.2, "Reactor Coolant MILLSTONE - UNIT 3 B 3/4 4-18 Amendment No. 797, 197

REACTOR COOLANT SYSTEM n,*.n August 27, 2001 BASES OVERPRESSURE PROTECTION SYSTEMS (continued)

Loops and Coolant Circulation - Cold Shutdown - Loops Not Filled," and LCO 3.4.1.3, "Reactor Coolant Loops and Coolant Circulation - Hot Shutdown" limit starting the first reactor coolant pump such that it shall not be started when any RCS loop wide range cold leg temperature is < 226°F unless the secondary side water temperature of each steam generator is < 50°F above each RCS cold leg temperature. The restrictions ensure 'the potential energy addition to the RCS from the secondary side of the steam generators will not result in an RCS overpressurization event beyond the capability of the COPPS to mitigate. The COPPS utilizes the pressurizer PORVs and the RHR relief valves to mitigate the limiting mass and energy addition events, thereby protecting the isothermal reactor vessel beltline P/T limits. The restrictions will ensure the reactor vessel will be protected from a cold overpressure event when starting the first RCP. If at least one RCP is operating, no restrictions are necessary to start additional RCPs for reactor vessel protection. In addition, this restriction only applies to RCS loops and associated components that are not isolated from the reactor vessel.

The RCP starting criteria are based on the equipment used to provide cold overpressure protection. A maximum temperature differential of 50°F between the steam generator secondary sides and RCS cold legs will limit the potential energy addition to within the capability of the pressurizer PORVs to mitigate the transient. The RHR relief valve are also adequate to mitigate energy addition transients constrained by this temperature differential limit, provided all RCS cold leg temperature are at or below 150°F. The ability of the RHR relief valves to mitigate energy addition transients when RCS cold leg temperature is above 1500F has not been analyzed. As a result, the temperature of the steam generator secondary sides must be at or below the RCS cold leg temperature if the RHR relief valves are providing cold overpressure protection and the RCS cold leg temperature is above 150°F.

MILLSTONE - UNIT 3 B 3/4 4-19 Amendment No. 7 7, 197

REACTOR COOLANT SYSTEM August 27, 200f BASES OVERPRESSURE PROTECTION SYSTEMS (continued)

The cold overpressure transient analyses demonstrate that either one relief valve or the depressurized RCS and RCS vent can maintain RCS pressure below limits when RCS letdown is isolated and only one centrifugal charging pump is operating.

Thus, the LCO allows only one centrifugal charging pump capable of injecting when cold overpressure protection is required.

The cold overpressure protection enabling temperature is conservatively established at a value < 226°F based on the criteria provided by ASME Section XI, I Appendix G.

PORV Performance The analyses show that the vessel is protected against non-ductile failure when 1 the PORVs are set to open at the values shown in Figures 3.4-4a and 3.4-4b within the tolerance allowed for the&calibration accuracy. The curves are derived by analyses for both three and four RCS loops unisolated that model the performance of the PORV cold overpressure protection system (COPPS), assuming the limiting mass and heat transients of one centrifugal charging pump injecting into the RCS, or the energy addition as a result of starting an RCP with temperature asymmetry between the RCS and the steam generators. These analyses consider pressure overshoot beyond the PORV opening setpoint resulting from signal processing and I valve stroke times.

The PORV setpoints in Figures 3.4-4a and 3.4-4b will be updated when the P/T limits conflict with the cold overpressure analysis limits. The P/T limits are periodically modified as the reactor vessel material toughness decreases due to neutr6n embrittlement. Revised limits are determined using neutron fluence projections and the results of testing of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.9.1, "Pressure/Temperature Limits

- Reactor'Coolant System (Except the Pressurizer)," discuss these evaluations.

The PORVs are considered active components. Thus, the failure of one PORV is assumed to represent the worst case, single active failure.

RHR Suction Relief Valve Performance The RHR suction relief valves do not have variable pressure and temperature lift setpoints as do the PORVs. Analyses show that one RHR suction relief valve with a setpoint at or between 426.8 psig and 453.2 psig will pass flow greater than that required for the limiting cold overpressure transient while maintaining KCS pressure less than the isothermal P/T limit curve. Assuming maximum relief flow requirements during the limiting cold overpressure event, an RHR suction relief valve will maintain RCS pressure to < 110% of the nominal lift setpoint.

Although each RHR suction relief valve is a passive spring loaded device, which meets single failure criteria, its location within the RHR System precludes meeting single failure criteria when spurious RHR suction isolation valve or RHR suction valve closure is postulated. Thus the loss of an RHR suction relief MILLSTONE - UNIT 3 B 3/4 4-20 Amendment No. JX7, 197

REACTOR COOLANT SYSTEM oh, August 27, 2001 BASES OVERPRESSURE PROTECTION SYSTEMS (continued) valve is the worst case single failure. Also, as the RCS P/T limits are revised to reflect change in toughness in the reactor vessel materials, the RHR suction relief valve's analyses must be re-evaluated to ensure continued accommodation of the design bases cold overpressure transients.

RCS Vent Performance With the RCS'depressurized, analyses show a vent size of > 2.0 square inches is l capable of mitigating the limiting cold overpressure transient. The capacity of this vent size is greater than the flow of the limiting transient, while maintaining RCS pressure less than the maximum pressure on the isothermal P/T limit curve.

The RCS vent size will be re-evaluated for compliance each time the isothermal P/T limit curves are revised.

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

The RCS vent satisfies Criterion 2 of IOCFR50.36(c)(2)(ii).

MILLSTONE'- UNIT 3 B 3/4 4-21 Amendment No. JR7, 197

REACTOR COOLANT SYSTEM August 27, 2001 BASES OVERPRESSURE PROTECTION SYSTEMS (continued)

LCO This LCO requires that cold overpressure protection be OPERABLE and the maximum mass input be limited to one charging pump. Failure to meet this LCO could lead to the loss of low temperature overpressure mitigation and violation of the reactor vessel isothermal P/T limits as a result of an operational transient.

To limit the mass input capability, the LCO requires a maximum of one centrifugal charging pump capable of injecting into the RCS.

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

I. Two OPERABLE PORVs; or A PORV is OPERABLE for cold overpressure protection when its block valve is' open, its lift setpoint is set to the nominal-setpoints provided for both three and four loops unisolated by Figure 3.4-4a or 3.4-4b and when the surveillance requirements are met.

2. Two OPERABLE RHR suction relief valves; or An RHR suction relief valve is OPERABLE for cold overpressure protection when its isolation valves from the RCS are open and when its setpoint is at or between 426.8 psig and 453.2 psig, as verified by required testing.

3. One OPERABLE PORV and one OPERABLE RHR suction relief valve; or

4. A depressurized RCS and an RCS vent.

An RCS vent is OPERABLE when open with an area of > 2.0 square inches.

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

MILLSTONE - UNIT 3 B 3/4' 4-22 Amendment No. W7, 197

3/4.7 PLANT SYSTEMS Jy-34a~4ýýý BASES 3/4.7.1 TURBTNE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensures that the Secondary System pressure will be limited to within 110% (1305 psig) of Its .design pressure of 1185 psig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a Turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of

-condenser heat. sink (i.e.; no steam byp.ss to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The design minimum total relieving capacity for all valves on all of the steam lines is 1.579 X 107 lbs/h which is 105% of the ar steam flow of 1.504 X 107 lbs/h at 100% RATED THERMAL POWER. A minimum of two OPERABL safety va ves per steam generator ensures t a su icient relieving capacity is available for the all L POWER restriction in Table 3.7-2.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in Secondary Coolant System steam flow and THERMAL POWER required by the reduced Reactor trip settings of the Power Range Neutron Flux channels. The Reactor Trip Setpoint reductions are derived on the following bases:

(wbf9M

,K where:

Rio - Safety Analysis power range high neutron flux setpoint, percent Q - Nominal NSSS power rating of the plant (including reactor coolant pump heat), Mwt K - Conversion factor, 947.82 1Btufsec) hi- heat of vaporization for steam at the highest MSSV opening pressure including tolerance (+/- 3%) and accumulation, as appropriate, Btu/lbm N - Number of loops in plant KIjLSOHE - 1UHIT 3 3/4 7-2 Amendment No-jM

Docket No. 50-423 B18814 Attachment 4 Millstone Power Station, Unit No. 3 License Basis Document Change Request 3-13-02 Elimination of 'N-I' Loop Operation from Technical Specifications Retyped Pages

INDEX SAFFTV IMTMS ANf ITMTTINA SAFFTY SYSTFM SFTTTNGS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE ..... .............. 2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT .... ............. 2-2 II FIGURE 2.1-2 DELETED .......... ..................... * . 2-3 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS ......... 2-4 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 2-5 RASFS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE . . . . . . . . . . . . . . . . . . . . . . . . . B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE ................ B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS ......... B 2-3 MILLSTONE - UNIT 3 iii 0959

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY ...... ......... ........................ 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - MODES I AND 2 .... ............ .. 3/4 1-1 Shutdown Margin - MODES 3, 4, AND 5 LOOPS FILLED . . . 3/4 1-3 FIGURE 3.1-1 REQUIRED SHUTDOWN MARGIN FOR MODE 3 .... ........ 3/4 1-4 II FIGURE 3.1-2 DELETED ...... ..... ...................... 3/4 1-5 FIGURE 3.1-3 REQUIRED SHUTDOWN MARGIN FOR MODE 4 .... ........ 3/4 1-6 FIGURE 3.1-4 REQUIRED SHUTDOWN MARGIN FOR MODE 5 WITH RCS LOOPS FILLED ..... ............... .. 3/4 1-7 Shutdown Margin - Cold Shutdown Loops Not Filled . . . . . . . . . . . . . . . . . . .

3/4 1-8 FIGURE 3.1-5 5 REQUIRED SHUTDOWN MARGIN FOR MODE 5 W ITH RCS LOOPS DRAINED .... ........... 3/4 1-9 Moderator Temperature Coefficient .... 3/4 1-10 Minimum Temperature for Criticality . . . 3/4 1-12 3/4.1.2 BORATION SYSTEMS DELETED ...... .................. 3/4 1-13 DELETED ...... .................. 3/4 1-14 DELETED ...... ..................

• 3/4 1-15 DELETED ...... ..................

• 3/4 1-16 DELETED ...... .................. 3/4 1-17 DELETED .... ............. ........ 3/4 1-18 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height . . . . . . . . . . . . . . . . . 3/4 1-20 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD . . . 3/4 1-22 Position Indication Systems - Operating . . . 3/4 1-23 MILLSTONE 0959

- UNIT 3 iv Amendment No. gq, P9, FF, 797,

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE DELETED .................................................. 3/4 1-24 Rod Drop Time ............................................ 3/4 1-25 Shutdown Rod Insertion Limit ............................. 3/4 1-26 Control Rod Insertion Limits ............................. 3/4 1-27 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE .................................... 3/4 2-1 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - FQ(Z) ..................... 3/4 2-5 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR ................................................... 3/4 2-19 3/4.2.4 QUADRANT POWER TILT RATIO ................................ 3/4 2-24 3/4.2.5 DNB PARAMETERS ........................................... 3/4 2-27 TABLE 3.2-1 DNB PARAMETERS ........................................ 3/4 2-28 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION ...................... 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION ................... 3/4 3-2 TABLE 3.3-2 DELETED TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ............................................. 3/4 3-10 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION .......................................... 3/4 3-15 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION .......................................... 3/4 3-17 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS ........................... 3/4 3-26 MILLSTONE - UNIT 3 v Amendment No. g, p, p, gl, 0959 77

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-13 DELETED TABLE 4.3-9 DELETED 3/4.3.4 DELETED 3/4.3.5 SHUTDOWN MARGIN MONITOR ..... ........... . . . .. . 3/4 3-82 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation ...... ................ ... 3/4 4-1 Hot Standby ...... ......... ....................... 3/4 4-2 Hot Shutdown ....... ......... ....................... 3/4 4-3 Cold Shutdown - Loops Filled ....... ............... ... 3/4 4-5 Cold Shutdown - Loops Not Filled ..... ............. .. 3/4 4-6 Loop Stop Valves ..... ..... ..................... ... 3/4 4-7 I Isolated Loop Startup ...... ..... ................... 3/4 4-8 3/4.4.2 SAFETY VALVES ...... ... ....................... .. 3/4 4-9 DELETED ...... ..... .......................... ... 3/4 4-10 3/4.4.3 PRESSURIZER Startup and Power Operation ...... ................ .. 3/4 4-11 FIGURE 3.4-5 PRESSURIZER LEVEL CONTROL ........ ............... 3/4 4-11a Hot Standby ...... ..... ........................ .. 3/4 4-11b 3/4.4.4 RELIEF VALVES ...... ... ....................... .. 3/4 4-12 3/4.4.5 STEAM GENERATORS ..... ..... ..................... ..3/4 4-14 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION ...... ................ .. 3/4 4-19 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION .... ............ .. 3/4 4-20 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems ...... ................. .. 3/4 4-21 Operational Leakage ...... ... .................... .. 3/4 4-22 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES . . 3/4 4-24 3/4.4.7 DELETED ...... ....... .......................... ..3/4 4-25 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS .... ........ 3/4 4-26 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS ....... ....... ....................... 3/4 4-27 3/4.4.8 SPECIFIC ACTIVITY ...... ... ..................... ..3/4 4-28 MILLSTONE - UNIT 3 vii Amendment No. 7*,

0960 7 70,

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE FIGURE 3.4-1 DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >IiCi/gram DOSE EQUIVALENT 1-131 ............... 3/4 4-30 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM . . . . . . . . . . . . . . . . . . . . . . 3/4 4-31 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ............ 3/4 4-33 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 10 EFPY .............. 3/4 4-34 FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 10 EFPY .............. 3/4 4-35 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE ................ 3/4 4-36 Pressurizer . . . . . . . . . . . . . . . . . . . . 3/4 4-37 Overpressure Protection Systems .......... 3/4 4-38 FIGURE 3.4-4a High Setpoint PORV Curve For the Cold Overpressure Protection System ................. 3/4 4-40 1 FIGURE 3.4-4b Low Setpoint PORV Curve For the Cold Overpressure Protection System ................. 3/4 4-41 I 3/4.4. 10 DELETED . . . . . . . . . . . . . . . . . . . . . . 3/4 4-42 3/4.4.11 DELETED . . . . . . . . . . . . . . . . . . . . . . 3/4 4-43 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS ..................... 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350°F . 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 350:F ........ 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK ............. 3/4 5-9 3/4.5.5 pH TRISODIUM PHOSPHATE STORAGE BASKETS ........ 3/4 5-10 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity ................ 3/4 6-1 Containment Leakage ................. 3/4 6-2 Containment Air Locks ................ 3/4 6-5 Containment Pressure ................. 3/4 6-7 MILLSTONE - UNIT 3 viii Amendment No. g, 07, ?Y 77X, 0960

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Air Temperature ............ S. . . . . .. . 3/4 6-9 Containment Structural Integrity .... S. . . . . .. . 3/4 6-10 Containment Ventilation System ....... ... . .... . 3/4 6-11 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Quench Spray System .... ... ..... . 3/4 6-12 Recirculation Spray System .......... . . . .. . 3/4 6-13 3/4.6.3 CONTAINMENT ISOLATION VALVES ...... ... ... .. . 3/4 6-15 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Monitors ........... ... . . . . . . 3/4 6-16 Electric Hydrogen Recombiners ..... ... . .. . . . 3/4 6-17 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM Steam Jet Air Ejector ......... S. . . . . . . 3/4 6-18 3/4.6.6 SECONDARY CONTAINMENT Supplementary Leak Collection and Release System . . 3/4 6-19 Secondary Containment ............... 3/4 6-22 Secondary Containment Structural Integrity ...... ................. 3/4 6-23 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves . . . . . . . . . . . . . . . . . . . 3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES 3/4 7-2 I I

TABLE 3.7-2 DELETED 3/4 7-2 MILLSTONE - UNIT 3 ix Amendment No. X, fl, 97, ?P, 70, 0960 7x, 719,

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9.6 REFUELING MACHINE ................... 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS ........ 3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level . . . . . . . . . . . . . . . . . . . 3/4 9-8 Low Water Level . . . . . . . . . . . . . . . . . . . . 3/4 9-9 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM .... 3/4 9-10 3/4.9.10 WATER LEVEL - REACTOR VESSEL ............. 3/4 9-11 3/4.9.11 WATER LEVEL - STORAGE POOL .............. 3/4 9-12 3/4.9.12 FUEL BUILDING EXHAUST FILTER SYSTEM ... ........... 3/4 9-13 3/4.9.13 SPENT FUEL POOL - REACTIVITY ............. 3/4 9-16 3/4.9.14 SPENT FUEL POOL - STORAGE PATTERN ........... 3/4 9-17 FIGURE 3.9-1 MINIMUM FUEL ASSEMBLY BURNUP VERSUS NOMINAL INITIAL ENRICHMENT FOR REGION 1 4-OUT-OF-4 STORAGE CONFIGURATION ........ ... .................... 3/4 9-18 FIGURE 3.9-2 REGION I 3-OUT-OF-4 STORAGE FUEL ASSEMBLY LOADING SCHEMATIC ........ ... ....................... 3/4 9-19 FIGURE 3.9-3 MINIMUM FUEL ASSEMBLY BURNUP VERSUS NOMINAL INITIAL ENRICHMENT FOR REGION 2 STORAGE CONFIGURATION .... 3/4 9-20 FIGURE 3.9-4 MINIMUM FUEL ASSEMBLY BURNUP AND DECAY TIME VERSUS NOMINAL INITIAL ENRICHMENT FOR REGION 3 STORAGE CONFIGURATION ........ ... .................... 3/4 9-21 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN . . . . . . . . . . . . . . . . . . . . 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS 3/4 10-2 I 3/4.10.3 PHYSICS TESTS . . . . . . . . . . . . . . . . . . . . . 3/4 10-4 3/4.10.4 REACTOR COOLANT LOOPS ....... .................. 3/4 10-5 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN ......... 3/4 10-6 3/4.11 DELETED 3/4.11.1 DELETED 3/4.11.2 DELETED 3/4.11.3 DELETED MILLSTONE - UNIT 3 xii Amendment 7p, gy, IFJ, 7Jp, 0961

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tav.g) shall not exceed the limits shown in Figure 2.1-1. I APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2750 psia be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

MILLSTONE 0962

- UNIT 3 2-1 Amendment No. 777,

680 660 LL C.,. 640 1-620 600 580 560 0 0.2 0.4 0.6 0.8 1.0 1.2 FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT I MILLSTONE - UNIT 3 2-2 Amendment No. Fq, 0962

This page intentionally left blank.

MILLSTONE - UNIT 3 2-3 Amendment No. go, 0962

TABLE 2.2-1 o- e REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

=.)

NOMINAL m FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE C 1. Manual Reactor Trip N.A. N.A.

-- 4

2. Power Range, Neutron Flux

a. High Setpoint 109% of RTP** < 109.6% of RTP** I

b. Low Setpoint 25% of RTP** 25.6% of RTP**

3. Power Range, Neutron Flux, 5% of RTP** with 5.6% of RTP** with High Positive Rate a time constant a time constant

> 2 seconds 2 seconds

4. Deleted

5. Intermediate Range, 25% of RTP** < 27.4% of RTP**

Neutron Flux (71

6. Source Range, Neutron Flux l X 10+ 5 cps < 1.06 x 10+5 cps

7. Overtemperature AT See Note 1 See Note 2 I

CD rt

• • RTP = RATED THERMAL POWER 0

NQ w1-

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOMINAL FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE I-C-

8. Overpower AT See Note 3 See Note 4

9. Pressurizer Pressure-Low 1900 psia > 1897.6 psia

10. Pressurizer Pressure-High 2385 psia < 2387.4 psia

11. Pressurizer Water Level-High 89% of instrument < 89.3% of instrument span span

12. Reactor Coolant Flow-Low 90% of loop > 89.8% of loop design flow* design flow*

13. Steam Generator Water 18.1% of narrow > 17.8% of narrow Level Low-Low range instrument range instrument I', span span

!1

14. General Warning Alarm N.A. N.A.

15. Low Shaft Speed - Reactor 92.4% of rated > 92.2% of rated Coolant Pumps speed speed 1

• Minimum Measured Flow Per Loop = 1/4 of the RCS Flow Rate Limit as listed in Section 3.2.3.1.a I

- 1.

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOMINAL o

(C C) FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE I

(I,

-I 0

z 16. Turbine Trip m

a. Low Fluid Oil Pressure 500 psig > 450 psig C

z

-I b. Turbine Stop Valve 1% open > 1% open Closure

17. Safety Injection Input N.A. N.A.

from ESF

18. Reactor Trip System Interlocks

a. Intermediate Range I x 10.10 amp > 9.0 x 10-11 amp Neutron Flux, P-6 T

4~ b. Low Power Reactor Trips Block, P-7

1) P-10 input (Note 5) 11% of RTP** < 11.6% of RTP**

2) P-13 input 10% RTP** Turbine < 10.6% RTP** Turbine Impulse Pressure Impulse Pressure Equivalent Equivalent ir

c. Power Range Neutron 37.5% of RTP** < 38.1% of RTP**

Flux, P-8

• • RTP = RATED THERMAL POWER

• 1.4

o--* TABLE 2.2-1 (Continued)

I-REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS m

NOMINAL z FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE

-=4

d. Power Range Neutron 51% of RTP** < 51.6% of RTP**

Flux, P-9

e. Power Range Neutron 9% of RTP** > 8.4% of RTP**

Flux, P-1O (Note 6)

19. Reactor Trip Breakers N.A. N.A.

20. Automatic Trip and Interlock N.A. N.A.

Logic T'

!o

21. DELETED I CL aM 0

,t V44

• • RTP = RATED THERMAL POWER

TABLE 2.2-1 (Continued)

TABLE NOTATIONS NOTE 1: OVERTEMPERATURE AT T )T(l+ :S)K K (1+t4S) (T- TI) +K3 (P -P') -f (AI)

Where: AT is measured Reactor Coolant System AT, OF; ATo is loop specific indicated AT at RATED THERMAL POWER, OF; (l+,rs)

(1+r 2 S) is the function generated by the lead-lag compensator on measured AT; r1 and T2 are the time constants utilized in the lead-lag compensator for AT, T, Ž 8 sec, 72 3 sec; K, < 1.20 2 > 0.02456/°F; IK (1+r 4 s)

(1+,5 s) is the function generated by the lead-lag compensator for Tav,;

74 and r. are the time constants utilized in the lead-lag compensator for T.v, 74 20 sec, T 5 4 sec; T is measured Reactor Coolant System average temperature, °F; T' is loop specific indicated Tavg at RATED THERMAL POWER, < 587.1°F; K3 > 0.001311/psi P is measured pressurizer pressure, psia;

"* P' is nominal pressurizer pressure, Ž 2250 psia; s is the Laplace transform operator, sec 1 ;

2.1 SAFETY LIMITS RASFS 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB. This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions.

The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux and is indicative of the margin to DNB.

The DNB design basis is as follows: uncertainties in the WRB-1 or WRB-2 correlations, plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, and computer codes are considered statistically such that there is at least a 95 percent probability with 95 percent confidence level that DNBR will not occur on the most limiting fuel rod during Condition I and II events. This establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties. In addition, margin has been maintained in the design by meeting safety analysis DNBR limits in performing safety analyses.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure, and average temperature below which the calculated DNBR is no less than the design DNBR value or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

These curves are based on an enthalpy hot channel factor, F NH , of 1.70 (includes measurement uncertainty) and a reference cosine with a peak of A.55 for axial power shape. An allowance is included for an increase in FA H at reduced power based on the expression:

FANH = 1.70 [1 + 0.3 (l-P)]

where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming axial imbalance is within the limits ofI F (delta I) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits.

MILLSTONE - UNIT 3 B 2-1 Amendment No. f?,

0964

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) functional capability at the specified trip setting is required for those anticipatory or diverse reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates a turbine trip signal whenever reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.

Manual Reactor Trip The Reactor Trip System includes manual Reactor trip capability.

Power Range, Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip setting. The Low Setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.

The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.

Power Range, Neutron Flux, High Positive Rate The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.

Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for all rod ejection accidents.

MILLSTONE - UNIT 3 B 2- 4 Amendment No. 771, 7I ,

0965

LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Range, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon trolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-1O becomes active. No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Trip System.

Overtemperature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors, and pressure is within the range between the Pressurizer High and Low Pressure trips. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1. Although a direction of conservatism is identified for the Overtemperature AT reactor trip function K? and K3 gains, the gains should be set as close as possible to the values contained in Note 1 to ensure that the Overtemperature AT setpoint is consistent with the assumptions of the safety analyses.

Overpower AT The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT MILLSTONE - UNIT 3 B 2-5 Amendment No. 11, P, 7gj, 0965

LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Trip System Interlocks (Continued)

P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or more reactor coolant loops. On decreasing power, the P-8 automatically blocks the above listed trips.

P-9 On increasing power, P-9 automatically enables Reactor trip on Turbine trip. On decreasing power, P-9 automatically blocks Reactor trip on Turbine trip.

P-10 On increasing power, P-1O provides input to P-7 to ensure that Reactor Trips on low flow in more than one reactor coolant loop, reactor coolant pump low shaft speed, pressurizer low pressure and pressurizer high level are active when power reaches 11%. It also allows the manual block of the Intermediate Range trip and the Low Setpoint Power Range trip; and automatically blocks the Source Range trip and deenergizes the Source Range high voltage power.

On decreasing power, P-10 resets to automatically reactivate the Intermediate Range trip and the Low Setpoint Power Range trip before power drops below 9%. It also provides input to reset P-7.

P-13 Provides input to P-7.

MILLSTONE - UNIT 3 B 2-8 Amendment No.  ?ý,

0966

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1.1 BORATION CONTROL SHUTDOWN MARGIN - MODES 1 AND 2 LIMITING CONDITION FOR OPERATION 3.1.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% Ak/k. I APPLICABILITY: MODES I and 2*.

ACTION:

With the SHUTDOWN MARGIN less than 1.3% Ak/k, immediately initiate and con tinue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.3% Ak/k:

a. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod(s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s);

b. When in MODE I or MODE 2 with Keff greater than or equal to 1 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6;

c. When in MODE 2 with Keff less than 1, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;

d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specification 4.1.1.1.2, with the control banks at the maximum insertion limit of Specification 3.1.3.6; and

• See Special Test Exceptions Specification 3.10.1.

MILLSTONE 0967

- UNIT 3 3/4 1-1 Amendment No. g, ;;7,

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - MODES 3, 4 AND 5 LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.1.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to the limits shown in Figures 3.1-1, 3.1-3 and 3.1-4.* I APPLICABILITY: MODES 3, 4 and 5 ACTION:

With the SHUTDOWN MARGIN less than the required value, immediately initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.2.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the required value:

a. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod(s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s); and

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following factors:

1) Reactor Coolant System boron concentration,

2) Control rod position,

3) Reactor Coolant System average temperature,

4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration.

4.1.1.1.2.2 Valve 3CHS-V305 shall be verified closed and locked at least once per 31 days.

• Additional SHUTDOWN MARGIN requirements, if required, are given in Specification 3.3.5.

MILLSTONE - UNIT 3 0968 3/4 1-3 Amendment No. ý9, 117, l9,

o'-*

0 =

r-cor I I (2050,3.495) rn 4 (2500,3.495)

-e 2" 3.5 3

z I ....

2.5 2

(JJ z I ------- 1-s-a 1.5 eI il/! 4 ,)'1 0 (o,1.300) - tfI12 uua. 1 -

1 CD

0. M 1 CD

(-I. 3: 0.5 0 U) 2,000 2,500 0 1,000 1,500

(] 500 (ppm)

RCS CRITICAL BORON CONCENTRATION I

REQUIREDFIGURE 3.1-1FOR MODE 3 SHUTDOWN I

This page intentionally left blank.

MILLSTONE - UNIT 3 3/4 1-5 Amendment No. , ,

0968

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued) c) A power distribution map is obtained from the movable incore detectors and FQ(Z) and FNAH are verified to be within their limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; and d) THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85%

of RATED THERMAL POWER.

c. With more than one rod trippable but inoperable due to causes other than addressed by ACTION a. above, POWER OPERATION may continue provided that:

1. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the remainder of the rods in the bank(s) with the inoperable rods are aligned to within +/-12 steps of the inoperable rods while maintaining the rod sequence and insertion limits of Specification 3.1.3.6. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, and

2. The inoperable rods are restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

d. With more than one rod misaligned from its group step counter demand height by more than +12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the rod position deviation monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 92 days.

MILLSTONE 0969

- UNIT 3 3/4 1-21 Amendment No. X, P, 7?7,

REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS - OPERATING ITMTTTN( CnNNTTTNN FNR nPFRATTnN 3.1.3.2 The Digital Rod Position Indication System and the Demand Position Indication System shall be OPERABLE and capable of determining the control rod positions within +/-12 steps.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. With a maximum of one digital rod position indicator per bank inoperable:

1. Determine the position of the nonindicating rod(s) indirectly by the movable incore detectors at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and immediately after any motion of the nonindicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b. With a maximum of one demand position indicator per bank inoperable:

1. Verify that all digital rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.3.2.1 Each digital rod position indicator shall be determined to be OPERABLE by verifying that the Demand Position Indication System and the Digital Rod Position Indication System agree within 12 steps at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the rod position deviation monitor is inoperable, then compare the Demand Position Indication System and the Digital Rod Position Indication System at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.1.3.2.2 Each of the above required digital rod position indicator(s) shall be determinded to be OPERABLE by verifying that the digital rod position indicators agree with the demand position indicators within 12 steps when exercised over the full-range of rod travel at least once per 24 months.

MILLSTONE - UNIT 3 3/4 1-23 Amendment No. 0, p, 7, 0970 7,

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE ITMTTTNG mNNlTTTNN FnR NPFRATTON 3.2.1.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:

a. The limits specified in the CORE OPERATING LIMITS REPORT (COLR) for Relaxed Axial Offset Control (RAOC) operation, or

b. Within the target band about the target flux difference during base load operation, specified in the COLR.

APPLICABILITY: MODE I above 50% RATED THERMAL POWER*.

ACTION:

a. For RAOC operation with the indicated AFD outside of the applicable limits specified in the COLR,

1. Either restore the indicated AFD to within the COLR specified limits within 15 minutes, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b. For base load operation above APLND with the indicted AFD outside of the applicable target band about the target flux differences:

1. Either restore the indicated AFD to within the COLR specified target band within 15 minutes, or

2. Reduce THERMAL POWER to less than APLND of RATED THERMAL POWER and discontinue base load operation within 30 minutes.

c. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR.

• See Special Test Exception 3.10.2 MILLSTONE - UNIT 3 3/4 2-1 Amendment No. -T, P, 0971

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MILLSTONE 0972

- UNIT 3 3/4 2-3 Amendment No. g, P,

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MILLSTONE 0972

- UNIT 3 3/4 2-4 Amendment No. g, f,

POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - FQ.

LIMITING CONDITION FOR OPERATION 3.2.2.1 FQ(Z) shall be limited by the following relationships:

F RTP F0 (Z) -.a K(Z) for P > 0.5 P

F RTP FO(Z) < Q K(Z) for P

• 0.5 0.5 QFTP = the FQ limit at RATED THERMAL POWER (RTP) provided in the core operating limits report (COLR).

Where: P = THERMAL POWER and RATED THERMAL POWER K(Z) = the normalized FQ(Z) as a function of core height specified in the COLR.

APPLICABILITY: MODE 1.

ACTION:

With FQ(Z) exceeding its limit:

a. For RAOC operation with Specification 4.2.2.1.2.b not being satisfied or for base load operation with Specification 4.2.2.1.4.b not being satisfied:

(1) Reduce THERMAL POWER at least 1%. for each 1% FQ(Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower AT Trip setpoints have been reduced at least 1% for each 1% FQ(Z) exceeds the limit, and MILLSTONE - UNIT 3 3/4 2-5 Amendment No. g, fo, g, 110, 0972 7709

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MILLSTONE 0973

- UNIT 3 3/4 2-12 Amendment No. g, f, YP, 119,

This page intentionally left blank.

MILLSTONE - UNIT 3 3/4 2-13 Amendment No. YP, 170, 0973

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MILLSTONE 0973

- UNIT 3 3/4 2-14 Amendment No. #f, f, Yg,

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MILLSTONE 0973

- UNIT 3 3/4 2-15 Amendment No. 77, 99, PP, 9, 779,

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MILLSTONE 0973

- UNIT 3 3/4 2-16 Amendment No. 77, M PP, M,

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MILLSTONE - UNIT 3 3/4 2-17 Amendment No. 77, 9, M, 9, 0973 170,

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MILLSTONE - UNIT 3 3/4 2-18 Amendment No. 77, 0, P, 99, 0973

POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3.1 The indicated Reactor Coolant System (RCS) total flow rate and FNH shall be maintained as follows:

a. RCS total flow rate > 371,920 gpm, and

b. FN F1 P [1.0 + PFH (1.0 - P)]

Where:

1) P = THERMAL POWER RATED THERMAL POWER

2) FH = Measured values of FNH obtained by using the movable incore detectors to obtain a power distribution map. The measured value of FaH should be used since Specification 3.2.3.1b.

takes into consideration a measurement uncertainty of 4% for incore measurement,

3) F RTP = The FNlimit at RATED THERMAL POWER in the CORE OPERATING LIMITS REPORT (COLR),

4) PFAH - The power factor multiplier for FNprovided in the COLR, and

5) The measured value of RCS total flow rate shall be used since uncertainties of 2.4% for flow measurement have been included in Specification 3.2.3.1a.

APPLICABILITY: MODE 1.

ACTION:

With the RCS total flow rate or FAH outside the region of acceptable operation:

a. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:

1. Restore the RCS total flow rate and FAH to within the above limits, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

MILLSTONE - UNIT 3 3/4 2-19 Amendment No. 11, g, 9, 719, 0973

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MILLSTONE - UNIT 3 3/4 2-22 Amendment No. 17, 9, P, 0974

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MILLSTONE 0974

- UNIT 3 3/4 2-23 Amendment No. 77, P9, 79, J0,

TABLE 3.2-1 o r-J

= DNB PARAMETERS

--4 0

z PARAMETER LIMITS Indicated Reactor Coolant System Tavg < 591.1°F Indicated Pressurizer Pressure > 2218 psia*

0o C+

• Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.

TABLE 3.3-1 o

-J

'-4 REACTOR TRIP SYSTEM INSTRUMENTATION

()

-4 0

MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE m

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION C 1, 24*, 5*

z'-4 1. Manual Reactor Trip 2 I 2 3*, 1

-4 2 1 2 11 (A)

2. Power Range, Neutron Flux 2

a. High Setpoint 4 3 1, 2 2

b. Low Setpoint 4 2 3 1###, 2 2

3. Power Range, Neutron Flux 4 2 3 1, 2 2 High Positive Rate

4. Deleted

5. Intermediate Range, Neutron Flux 2 1 2 1###, 2 3 (A)

6. Source Range, Neutron Flux 2##

(A)

a. Startup 2 2 3*, 4*, 5* 4

b. Shutdown 2 1 2 11

7. Overtemperature AT 4 2 3 1, 2 6 I I

8. Overpower AT 4 2 3 1, 2 6

9. Pressurizer Pressure--Low 4 2 3 I** 6 (1)

0. 10. Pressurizer Pressure--High 4 2 3 1, 2 6 (1)

CD 6 3 2 2 1**

r9 11. Pressurizer Water Level--High z

0

• -1

TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION Ir r")

--4 MINIMUM CHANNELS 0

TOTAL NO. CHANNELS APPLICABLE MODES m TO TRIP OPERABLE ACTION OF CHANNELS FUNCTIONAL UNIT z

12. Reactor Coolant Flow--Low 2/loop 2/loop 1 6 I

a. Single Loop (Above P-8) 3/loop 2/loop 1 6 I 3/loop 2/loop in

b. Two Loops (Above P-7 and two oper below P-8) ating loops 2/stm. gen. 3/stm. gen. 1, 2 6 (1)

13. Steam Generator Water 4/stm. gen.

Level--Low-Low I

14. Low Shaft Speed--Reactor 3 1** 6 Coolant Pumps 4-1/pump 2

15. Turbine Trip 3 2 126 2 4 1"***

Cj.) a. Low Fluid Oil Pressure 4 4

b. Turbine Stop Valve Closure

16. Deleted

17. Reactor Trip System Interlocks

a. Intermediate Range 2 8 Neutron Flux, P-6 2 1

b. Low Power Reactor Trips Block, P-7 1 8 P-1O Input 4 2 3

= or 2 8 0 2 1 1 P-13 Input

-Q

-'-4

Co TABLE 3.3-1 (Continued)

-4 I

1 REACTOR TRIP SYSTEM INSTRUMENTATION

-I 0

z MINIMUM m

TOTAL NO. CHANNELS CHANNELS APPLICABLE C FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

'-4

-I

17. Reactor Trip System Interlocks (Continued)

c. Power Range Neutron 4 2 3 1 8 Flux, P-8 2 3 8

d. Power Range Neutron 4 Flux, P-9

e. Power Range Neutron 4 2 3 1,2 8 Flux, P-10 (A,

1, 2 2 1 2 10, 13 (A, 18. Reactor Trip Breakers( ) 2 1 3*, 4*, 5* 11 2 2 1 2 1, 2 5 13A

19. Automatic Trip and Interlock 2 2 1 2 11 Logic

20. DELETED I

21. DELETED r1 z

- "Q

TABLE 3.3-1 (Continued)

TABLE NOTATIONS

• When the Reactor Trip System breakers are in the closed position and the Control Rod Drive System is capable of rod withdrawal.

• • Above the P-7 (At Power) Setpoint.

• • • Above the P-9 (Reactor Trip/Turbine Trip Interlock) Setpoint.

1. 1. Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

1. 1. 1. Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

(1) The applicable MODES and ACTION statements for these channels noted in Table 3.3-3 are more restrictive and, therefore, applicable.

(2) Including any reactor trip bypass breakers that are racked in and closed for bypassing a reactor trip breaker.

ACTION STATEMENTS ACTION I - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1, and

c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85%

of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Specification 4.2.4.2.

MILLSTONE - UNIT 3 3/4 3-5 Amendment No. 97, M, 9, 9

0976

TABLE 4.3-1 (Continued)

I REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS z TRIP

'I ACTUATING MODES WHICH FOR ANALOG CHANNEL DEVICE OPERATIONAL SURVEILLANCE C OPERATIONAL ACTUATION LOGIC TEST IS REQUIRED CHANNEL CHANNEL TEST

-I CHECK CALIBRATION TEST FUNCTIONAL UNIT N.A.

N.A. M(7, 11)

N.A. N.A.

18. Reactor Trip Breaker M(7) 1, 2, 3*,

19. Automatic Trip and N.A. N.A. N.A. N.A. 4*, 5*

Interlock Logic

20. DELETED N.A. 1,2 ,3*,

N.A. RM 15)

N.A. N.A.

21. Reactor Trip Bypass Breaker

22. DELETED IVA I',,.

(C)

"I,,,1

o =C TABLE 3.3-3 (Continued) 00 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION r"

ENGINEERED SAFETY MINIMUM I- TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

5. Turbine Trip and Feedwater Isolation

a. Automatic Actuation 2 1 2 1, 2 25 Logic and Actuaion Relays

b. Steam Generator 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2, 3 20, 21 Water Level - in each in any oper in each High-High (P-14) operating ating loop operating (j

loop loop

c. Safety Injection 2 1 2 1,2 22 0 Actuation Logic

d. Tav, Low Coincident 1Tave/l oop I Tave in 1 T.. in 1, 2 20 with P-4 any two any three loops loops

'1.

-~3

TABLE 3.3-4 o 3=

-i ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS rrlr NOMINAL I- FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE z 5. Turbine Trip and Feedwater C

Isolation

a. Automatic Actuation Logic N.A. N.A.

Actuation Relays

b. Steam Generator Water 80.5% of narrow < 80.8% of narrow Level--High-High (P-14) range instrument range instrument span. span.

c. Safety Injection Actuation See Item 1. above for all Safety Injection Trip Logic Setpoints and Allowable Values.

d. Tave Low Coincident with 564°F > 563.6°F Reactor Trip (P-4) r3 6. Auxiliary Feedwater CL, 0,

a. Manual Initiation N.A. N.A.

b. Automatic Actuation Logic N.A. N.A.

CD and Actuation Relays

c. Steam Generator Water 0

Level--Low-Low

1) Start Motor-Driven 18.1% of > 17.8% of narrow Pumps narrow range range instrument span.

instrument span.

INSTRUMENTATION 3/4.3.5 SHUTDOWN MARGIN MONITOR LIMITING CONDITION FOR OPERATION 3.3.5 Two channels of Shutdown Margin Monitors shall be OPERABLE

a. With a minimum count rate as designated in the CORE OPERATING LIMITS REPORT (COLR), or

b. If the minimum count rate in Specification 3.3.5.a cannot be met, then the Shutdown Margin Monitors may be made operable with a lower minimum count rate, as specified in the COLR, by borating the Reactor Coolant System above the requirements of Specification 3.1.1.1.2 or 3.1.1.2. The additional boration shall be:

1. A minimum of 150 ppm above the SHUTDOWN MARGIN require ments of Figure 3.1-1 (Mode 3), or

2. A minimum of 350 ppm above the SHUTDOWN MARGIN require ments of Figure 3.1-3 (Mode 4), Figure 3.1-4 (Mode 5 - RCS loops filled) and Figure 3.1-5 (Mode 5 - RCS loops drained).

APPLICABILITY: MODES 3*, 4, and 5.

ACTION:

a. With one Shutdown Margin Monitor inoperable, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

b. With both Shutdown Margin Monitors inoperable or one Shutdown Margin Monitor inoperable for greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, immediately suspend all operations involving positive reactivity changes via dilution and rod withdrawal. Verify the valves listed in Specifica tion 4.1.1.2.2 are closed and secured in position within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least once per 14 days thereafter.** Verify comp liance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1.2 or 3.1.1.2, as applicable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

• The shutdown margin monitors may be blocked during reactor startup in accordance with approved plant procedures.

"**The valves may be opened on an intermittent basis under administrative controls as noted in Surveillance 4.1.1.2.2.

MTttSTONF - UNIT 3 3/4 3-82 Amendment No. ,

G 0980

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION I ITMTTING COMNITTNN FNR NPFRATTON 3.4.1.1 Four reactor coolant loops shall be OPERABLE and in operation. I I

APPLICABILITY: MODES 1 and 2.*

ACTION:

With less than the above required reactor coolant loops in operation, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

rIIRVFTI IANCF RFQIITRFMFNTS 4.4.1.1 The above required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• See Special Test Exceptions Specification 3.10.4.

MILLSTONE - UNIT 3 3/4 4-1 Amendment No.

0981

REACTOR COOLANT SYSTEM LOOP STOP VALVES I 1MTT7N( CONDTTml FflR nPFRATTnN removed from the 3.4.1.5 Each RCS loop stop valve shall be open and the power valve operator.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

power

a. With power available to one or more loop stop valve operators, removeSTANDBY from the loop stop valve operators within 30 minutes or be in HOT hours.

within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 valve(s) b."I With one or more RCS loop stop valves closed, maintain thewithin the closed and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.1.5 Verify each RCS loop stop valve is open and the power removed valve operator at least once per 31 days.

from the I

All required actions of Action Statement 3.4.1.5.b shall be completed whenever this action is entered.

,,, *iu,,,slT 'AIA a-7 Amendment No.

MIILL3uIUr- - u"11 *I-T I

• 0982

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES I TMTTTN CnNNTTTNN FNR OPFRATTON 3.7.1.1 All main steam line Code safety valves shall be OPERABLE with lift settings as specified in Table 3.7-3.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With one or more main steam line Code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

qIIRVF1I IANrF RFQIITRFMFNT!

4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5.

MILLSTONE - UNIT 3 0983 3/4 7-1 Amendment No. 97,

TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES I MAXIMUM NUMBER OF INOPERABLE MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR (PERCENT OF RATED THERMAL POWER)

I 65 2 46 3 28 TABLE 3.7-2 DELETED I

MILLSTONE 0983

- UNIT 3 3/4 7-2 Amendment No. jol,

SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS I

ITMTTTNG -ONfITTTAN FR nPFRATITN 3.10.2.1 The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and

b. The limits of Specifications 3.2.2.1 and 3.2.3.1 are maintained specified in Specification and determined at the frequencies 4.10.2.1.2 below.

APPLICABILITY: MODE 1.

ACTION:

With any of the limits of Specification 3.2.2.1 or 3.2.3.1 being exceeded while the are requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1.1, and 3.2.4 suspended, either:

a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2.1 and 3.2.3.1, or

b. Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SIIRVFTI IANCF RFOIIIRFMFNTq to 85%

4.10.2.1.1 The THERMAL POWER shall be determined to be less than or equal of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.2.1.2 The Surveillance Requirements of the below listed specifications shall be performed at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during PHYSICS TESTS:

a. Specifications 4.2.2.1.2 and 4.2.2.1.3, and

b. Specification 4.2.3.1.2.

3/4 10-2 Amendment No.

MILLSTONE - UNIT 3 0984

This page intentionally left blank.

MILLSTONE - UNIT 3 3/4 10-3 Amendment No.

0984

POWER DISTRIBUTION LIMITS N

1iANIM-AXIAL FLUX DIFFERENCE (Continued)

(2) APLND (for base load operation). Penalty deviation minutes for base load operation are not accumulated based on the short period of time during which operation outside of the target band is allowed.

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flow rate, and nuclear enthalpy rise hot channel factor ensure that: (I) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200°F ECCS acceptance criteria limit.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to ensure that the limits are maintained provided:

a. Control rods in a single group move together with no individual rod insertion differing by more than +12 steps, indicated, from the group demand position;

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6;

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and

d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

FNAH will be maintained within its limits provided Conditions a. through

d. above are maintained. The relaxation of FNAH as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

The FNAH as calculated in Specification 3.2.3.1 is used in the various accident analyses where FNH influences parameters other than DNBR, e.g., peak clad temperature, and thus is the maximum "as measured" value allowed.

MILLSTONE - UNIT 3 B 3/4 2-3 Amendment No. g, P9, 0985

POWER DISTRIBUTION LIMITS RAF5 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

Margin is maintained between the safety analysis limit DNBR and the design limit DNBR. This margin is more than sufficient to offset any rod bow penalty and transition core penalty. The remaining margin is available for plant design flexibility.

When an FQ measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance.

The heat flux hot channel factor, FQ(Z), is measured periodically using the incore detector system. These measurements are generally taken with the core at or near steady state conditions. Using the measured three dimensional power distributions, it is possible to derive FQM(Z), a computed value of FQ(Z).

However, because this value represents a steady state condition, it does not include the variations in the value of FQ(Z) that are present during nonequilibrium situations.

To account for these possible variations, the steady state limit of FQ(Z) is adjusted by an elevation dependent factor appropriate to either RAOC or base load operation, W(Z) or W(Z)BL, that accounts for the calculated worst case transient conditions. The W(Z) and W(Z)BL, factors described above for normal operation are specified in the COLR per Specification 6.9.1.6. Core monitoring and control under nonsteady state conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and control rod insertion. Evaluation of the steady state FQ(Z) limit is performed in Specification 4.2.2.1.2.b and 4.2.2.1.4.b while evaluation nonequilibrium limits are performed in Specification 4.2.2.1.2.c and 4.2.2.1.4.c.

When RCS flow rate and FNAH are measured, no additional allowances are necessary prior to comparison with the limits of the Limiting Condition for Operation. Measurement errors of 2.4% for RCS total flow rate and 4% for FN H have been allowed for in determination of the design DNBR value.

The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate indicators. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non conservative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi will be added if venturis are not inspected and cleaned at least once for 18 months. Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

MILLSTONE - UNIT 3 B 3/4 2-4 Amendment No. 77, M 779, 0985

POWER DISTRIBUTION LIMITS RASFE HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the accept able region of operation defined in Specification 3.2.3.1.

3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during power operation.

The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in FQ is depleted. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.

The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc tion of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on FQ is reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.

3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR greater than the design limit throughout each analyzed transient. The indicated T.,9 value of 591.1°F MILLSTONE 0985

- UNIT 3 B 3/4 2-5 Amendment No. 77, M, P9,

POWER DISTRIBUTION LIMITS KitI-N DNB PARAMETERS (Continued) and the indicated pressurizer pressure value is 2218 psia. The calculated values of the DNB related parameters will be an average of the indicated values for the operable channels.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

Measurement uncertainties have been accounted for in determining the parameter limits.

MILLSTONE 0985

- UNIT 3 B 3/4 2-6 Amendment No. 17, P,

INSTRUMENTATION BASES 3/4 3.5 SHUTDOWN MARGIN MONITOR The Shutdown Margin Monitors provide an alarm that a Boron Dilution Event may be in progress. The minimum count rate of Specification 3/4.3.5 and the SHUTDOWN MARGIN requirements of Figures 3.1-1, 3.1-3, 3.1-4, and 3.1-5 ensure that at least 15 minutes are available for operator action from the time of the Shutdown Margin Monitor alarm to total loss of shutdown margin. By borating an additional 150 ppm above the SHUTDOWN MARGIN required by Figure 3.1-1, or 350 ppm above the SHUTDOWN MARGIN required by Figure 3.1-3, 3.1-4, or 3.1-5, lower values of minimum count rate are accepted.

Shutdown Margin Monitors

Background:

The purpose of the Shutdown Margin Monitors (SMM) is to annunciate an increase in core subcritical multiplication allowing the operator at least 15 minutes response time to mitigate the consequences of the inadvertent addition of unborated primary grade water (boron dilution event) into the Reactor Coolant System (RCS) when the reactor is shut down (Modes 3, 4, and 5).

The SMMs utilizes two channels of source range instrumentation (GM detectors).

Each channel provides a signal to its applicable train of SMM. The SMM channel uses the last 600 or more counts to calculate the count rate and updates the measurement after 30 new counts or I second, whichever is longer.

Each channel has 20 registers that hold the counts (20 registers X 30 count =

600 counts) for averaging the rate. As the count rate decreases, the longer it takes to fill the registers (fill the 30 count minimum). As the instrument's measured count rate decreases, the delay time in the instrument's response increases. This delay time leads to the requirement of a minimum count rate for OPERABILITY.

During the dilution event, count rate will increase to a level above the normal steady state count rate. When this new count rate level increases above the instrument's setpoint, the channel will alarm alerting the operator of the event.

Applicable Safety Analysis The SMM senses abnormal increases in the source range count per second and alarms the operator of an inadvertent dilution event. This alarm will occur at least 15 minutes prior to the reactor achieving criticality. This 15 minute window allows adequate operator response time to terminate the dilution, FSAR Section 15.4.6.

LCO LCO 3.3.5 provides the requirements for OPERABILITY of the instrumentation of the SMMs that are used to mitigate the boron dilution event. Two trains are required to be OPERABLE to provide protection against single failure.

MILLSTONE - UNIT 3 B 3/4 3-7 Amendment No. If, 0986

3/4.4 REACTOR COOLANT SYSTEM RA SF 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The purpose of Specification 3.4.1.1 is to require adequate forced flow rate for core heat removal in MODES I and 2 during all normal operations and anticipated transients. Flow is represented by the number of reactor coolant pumps in operation for removal of heat by the steam generators. To meet safety analysis acceptance criteria for DNB, four reactor coolant pumps are required at rated power. An OPERABLE reactor coolant loop consists of an OPERABLE reactor coolant pump in operation providing forced flow for heat transport and an OPERABLE steam generator in accordance with Specification 3.4.5. With less than the required reactor coolant loops in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

In MODE 3, three reactor coolant loops, and in Mode 4, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, in MODE 3 a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., the Control Rod Drive System is not capable of rod withdrawal.

In MODE 4, if a bank withdrawal accident can be prevented, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (any combination of RHR or RCS) be OPERABLE.

In MODE 5, with reactor coolant loops filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two RHR loops or at least one RHR loop and two steam generators be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE.

In MODE 5, during a planned heatup to MODE 4 with all RHR loops removed from operation, an RCS loop, OPERABLE and in operation, meets the requirements of an OPERABLE and operating RHR loop to circulate reactor coolant. During the heatup there is no requirement for heat removal capability so the OPERABLE and operating RCS loop meets all of the required functions for the heatup condition. Since failure of the RCS loop, which is OPERABLE and operating, could also cause the associated steam generator to be inoperable, the associated steam generator cannot be used as one of the steam generators used to meet the requirement of LCO 3.4.1.4.1.b.

The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The restrictions on starting the first RCP in MODE 4 below the cold overpressure protection enable temperature (226°F), and in MODE 5 are provided to prevent RCS pressure transients. These transients, energy additions due to the differential temperature between the steam generator secondary side and the RCS, can result in pressure excursions which could challenge the P/T limits.

MILLSTONE - UNIT 3 B 3/4 4-1 Amendment No. 7, 9, 99, 797, 0987 ,

3/4.4 REACTOR COOLANT SYSTEM BASES (Continued)

The RCS will be protected against overpressure transients and will not exceed the reactor vessel isothermal beltline P/T limit by restricting RCP starts based on the differential water temperature between the secondary side of each steam generator and the RCS cold legs. The restrictions on starting the first RCP only apply to RCPs in RCS loops that are not isolated. The restoration of isolated RCS loops is normally accomplished with all RCPs secured. If an isolated RCS loop is to be restored when an RCP is operating, the appropriate temperature differential limit between the secondary side of the isolated loop steam generator and the in service RCS cold legs is applicable, and shall be met prior to opening the loop isolation valves.

The temperature differential limit between the secondary side of the steam generators and the RCS cold legs is based on the equipment providing cold overpressure protection as required by Technical Specification 3.4.9.3. If the pressurizer PORVs are providing cold overpressure protection, the steam generator secondary to RCS cold leg water temperature differential is limited to a maximum of 500 F. If any RHR relief valve is providing cold overpressure protection and RCS cold leg temperature is above 150°F, the steam generator secondary water temperature must be at or below RCS cold leg water temperature.

If any RHR relief valve is providing cold overpressure protection and RCS cold leg temperature is at or below 150°F, the steam generator secondary to RCS cold leg water temperature differential is limited to a maximum of 50°F.

Specification 3.4.1.5 The reactor coolant loops are equipped with loop stop valves that permit any loop to be isolated from the reactor vessel. One valve is installed on each hot leg and one on each cold leg. The loop stop valves are used to perform maintenance on an isolated loop. Operation in MODES 1-4 with a RCS loop stop valve closed is not permitted except for the mitigation of emergency or abnormal events. If a loop stop valve is closed for any reason, the required actions of this specification must be completed. To ensure that inadvertent closure of a loop stop valve does not occur, the valves must be open with power to the valve operators removed in MODES 1, 2, 3 and 4.

The safety analyses performed for the reactor at power assume that all reactor coolant loops are initially in operation and the loop stop valves are open. This LCO places controls on the loop stop valves to ensure that the valves are not inadvertently closed in MODES 1, 2, 3 and 4. The inadvertent closure of a loop stop valve when the Reactor Coolant Pumps (RCPs) are operating will result in a partial loss of forced reactor coolant flow. If the reactor is at rated power at the time of the event, the effect of the partial loss of forced coolant flow is a rapid increase in the coolant temperature which could result in DNB with subsequent fuel damage if the reactor is not tripped by the Low Flow reactor trip. If the reactor is shutdown and a RCS loop is in operation removing decay heat, closure of the loop stop valve associated with the operating loop could also result in increasing coolant temperature and the possibility of fuel damage.

The loop stop valves have motor operators. If power is inadvertently restored to one or more loop stop valve operators, the potential exists for accidental closure of the affected loop stop valve(s) and the partial loss of forced reactor coolant flow. With power applied to a valve operator, only the interlocks prevent the valve from being operated. Although operating MILLSTONE - UNIT 3 B 3/4 4-1a Amendment No. 7, 9, M, W*7, 0987 797, 70,

3/4.4 REACTOR COOLANT SYSTEM procedures and interlocks make the occurrence of this event unlikely, the prudent action is to remove power from the loop stop valve operators. The time period of 30 minutes to remove power from the loop stop valve operators is sufficient considering the complexity of the task.

Should a loop stop valve be closed in MODES 1 through 4, the affected valve must be maintained closed and the plant placed in MODE 5. Once in MODE 5, the isolated loop may be started in a controlled manner in accordance with LCO 3.4.1.6, "Reactor Coolant System Isolated Loop Startup." Opening the closed loop stop valve in MODES I through 4 could result in colder water or water at a lower boron concentration being mixed with the operating RCS loops resulting in positive reactivity insertion. The time period provided in Action 3.4.1.5.b allows time for borating the operating loops to a shutdown boration level such that the plant can be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The allowed action times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

Surveillance Requirement 4.4.1.5 is performed at least once per 31 days to ensure that the RCS loop stop valves are open, with power removed from the loop stop valve operators. The primary function of this Surveillance is to ensure that power is removed from the valve operators, since Surveillance Requirement 4.4.1.1 requires verification every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that all loops are operating and circulating reactor coolant, thereby ensuring that the loop stop valves are open. The frequency of 31 days ensures that the required flow is available, is based on engineering judgement, and has proven to be acceptable.

Operating experience has shown that the failure rate is so low that the 31 day frequency is justified.

Specification 3.4.1.6 The requirement to maintain the isolated loop stop valves shut with power removed ensures that no reactivity addition to the core could occur due to the startup of an isolated loop. Verification of the boron concentration in an isolated loop prior to opening the first stop valve provides a reassurance of the adequacy of the boron concentration in the isolated loop.

RCS Loops Filled/Not Filled:

In MODE 5, any RHR train with only one cold leg injection path is sufficient to provide adequate core cooling and prevent stratification of boron in the Reactor Coolant System.

The definition of operability states that the system or subsystem must be capable of performing its specified function(s). The reason for the operation of one reactor coolant pump (RCP) or one RHR pump is to:

• Provide sufficient decay heat removal capability

• Provide adequate flow to ensure mixing to:

• Prevent stratification

• Produce gradual reactivity changes due to boron concentration changes in the RCS The definition of "Reactor coolant loops filled" includes a loop that is filled, swept, and vented, and capable of supporting natural circulation heat transfer. This allows the non-operating RHR loop to be removed from service MILLSTONE - UNIT 3 B 3/4 4-lb Amendment No. 7, M, P, 797, 0987 1 97l,707

3/4.4 REACTOR COOLANT SYSTEM BASES (Continued) while filling and unisolating loops as long as steam generators on the operable reactor coolant loops are available to support decay heat removal. Any loop being unisolated is not OPERABLE until the loop has been swept and vented. The process of sweep and vent will make the previously OPERABLE loops inoperable and the requirements of LCO 3.4.1.4.2, "Reactor Coolant System, Cold Shutdown Loops Not Filled," are applicable. When the RCS has been filled, swept and vented using an approved procedure, all unisolated loops may be declared OPERABLE.

One cold leg injection isolation valve on an RHR train may be closed without considering the train to be inoperable, as long as the following conditions exist:

• CCP temperature is at or below 95°F

• Initial RHR temperature is below 184°F

• The single RHR cold leg injection flow path is not utilized until a minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown

• CCP flow is at least 6,600 gpm

• RHR flow is at least 2,000 gpm In the above system lineup, total flow to the core is decreased compared to the flow when two cold legs are in service. This is acceptable due to the substantial margin between the flow required for cooling and the flow available, even through a slightly restricted RHR train.

The review concerning boron stratification with the utilization of the single injection point line, indicates there will not be a significant change in the flow rate or distribution through the core, so there is not an increased concern due to stratification.

Flow velocity, which is high, is not a concern from a flow erosion or pipe loading standpoint. There are no loads imposed on the piping system which would exceed those experienced in a seismic event. The temperature of the fluid is low and is not significant from a flow erosion standpoint.

The boron dilution accident analysis, for Millstone Unit 3 in MODE 5, assumes a full RHR System flow of approximately 4,000 gpm. Westinghouse analysis, Reference (1), for RHR flows down to 1,000 gpm, determined adequate mixing results. As the configuration will result in a RHR flow rate only slightly less then 4,000 gpm there is no concern in regards to a boron dilution accident.

The basis for the requirement of two RCS loops OPERABLE is to provide natural circulation heat sink in the event the operating RHR loop is lost. If the RHR loop were lost, with two loops swept and vented and two loops air bound, natural circulation would be established in the two swept loops.

Natural circulation would not be established in the air bound loops. Since there would be no circulation in the air bound loops, there would be no mechanism for the air in those loops to be carried to the vessel, and subsequently into the swept loops rendering them inoperable for heat sink requirements.

MILLSTONE - UNIT 3 B 3/4 4-1c Amendment No.

0987

3/4.4 REACTOR COOLANT SYSTEM BASES (Continued)

The LCO is met as long as at least two reactor coolant loops are OPERABLE and the following conditions are satisfied:

One RHR loop is OPERABLE and in operation, with exceptions as allowed in Technical Specifications; and Either of the following:

• An additional RHR loop OPERABLE, with exceptions as allowed in Technical Specifications; or

• The secondary side water level of at least two steam generators shall be greater than 17% (These are assumed to be on OPERABLE reactor coolant loops)

When the reactor coolant loops are swept, the mechanism exists for air to be carried into previously OPERABLE loops. All previously OPERABLE loops are declared inoperable and an additional RHR loop is required OPERABLE as specified by LCO 3.4.1.4.2 for loops not filled. When the RCS has been filled, swept, and vented using an approved procedure, all unisolated loops may be declared OPERABLE.

ISOLATED LOOP STARTUP The below requirements are for unisolating a loop with all four loops isolated while decay heat is being removed by RHR and to clarify prerequisites to meet T/S requirements for unisolating a loop at any time.

With no RCS loops operating, the two RHR loops referenced in Specification 3.4.1.4.2 are the operating loops. Starting in MODE 4 as referenced in Specification 3.4.1.3, the RHR loops are allowed to be used in place of an operating RCS loop. Specification 3.4.1.4.2 requires two RHR loops OPERABLE and at least one in operation. Ensuring the isolated cold leg temperature is within 20'F of the highest RHR outlet temperature for the operating RHR loops within 30 minutes prior to opening the cold leg stop valve is a conservative approach since the major concern is a positive reactivity addition.

SR 4.4.1.6.1: When in MODE 5 with all RCS loops isolated, the two RHR loops referenced in LCO 3.4.1.4.2 shall be considered the OPERABLE RCS loops.

ISOLATED LOOP STARTUP (Continued)

The isolated loop cold leg temperature shall be determined to be within 20°F of the highest RHR outlet temperature for the operating RHR loops within 30 minutes prior to opening the cold leg stop valve.

Surveillance requirement 4.4.1.6.2 is met when the following actions occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to opening the cold leg or hot leg stop valve:

• An RCS boron sample has been taken and analyzed to determine current boron concentration

• The SHUTDOWN MARGIN has been determined using OP 3209B, "Shutdown Margin" using the current boron concentration determined above

• For the isolated loop being restored, the power to both loop stop valves has been restored MILLSTONE - UNIT 3 B 3/4 4-1d Amendment No.

0987

3/4.4 REACTOR COOLANT SYSTEM BASES (Continued)

Surveillance 4.4.1.6.2 indicates that the reactor shall be determined subcritical by at least the amount required by Specifications 3.1.1.1.2 or 3.1.1.2 for MODE 5 or Specification 3.9.1.1 for MODE 6 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of opening the cold leg or hot leg stop valve.

Specification 3.1.1.1.2 requires the SHUTDOWN MARGIN to be as shown in Figure 3.1-2 for three loop operation. Figure 3.1-2 is for three loop operation in MODE 3. The other figures, as used by this specification, require four loop operation, so cannot be used to determine the required SHUTDOWN MARGIN for the MODE 5 loops isolated condition.

Specification 3.1.1.2 requires the SHUTDOWN MARGIN to be as shown in Figure 3.1-5 or Figure 3.1-4 with CVCS aligned to preclude boron dilution.

This specification is for loops not filled and therefore is applicable to an all loops isolated condition.

Specification 3.9.1.1 requires Keff of 0.95 or less, or a boron concentration of greater than or equal to 2,600 ppm in MODE 6.

Specification 3.1.1.1.2 or 3.1.1.2 for MODE 5, both require boron concentration to be determined at least once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. SR 4.1.1.1.2.1.b.2 and 4.1.1.2.1.b.1 satisfy the requirements of Specifications 3.1.1.1.2 and 3.1.1.2 respectfully. Specification 3.9.1.1 for MODE 6 requires boron concentration to be determined at least once each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. S.R.4.9.1.1.2 satisfy the requirements of Specification 3.9.1.1.

References:

1. Letter NEU-94-623, dated July 13, 1994; Mixing Evaluation for Boron Dilution Accident in Modes 4 and 5, Westinghouse HR-59782.

2. Memo No. MP3-E-93-821, dated October 7, 1993.

MILLSTONE - UNIT 3 B 3/4 4-1e Amendment No.

0987

REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (continued)

Violating the LCO limits places the reactor vessel outside of the bounds of the analyses and can increase stresses in other RCPB components. The consequences depend on several factors, as follows:

a. The severity of the departure from the allowable operating P/T regime or the severity of the rate of change of temperature;

b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and

c. The existences, sizes, and orientations of flaws in the vessel material.

APPLICABILITY The RCS P/T limits LCO provides a definition of acceptable operation for prevention of nonductile failure of ferritic RCS components using ASME Section XI, Appendix G, as modified by Code Case N-640 and the additional requirements of IOCFR50, Appendix G (Ref. 1). The P/T limits were developed to provide requirements for operation during heatup or cooldown (MODES 3, 4, and 5) or ISLH testing, in keeping with the concern for nonductile failure. The limits do not apply to the Pressurizer.

During MODES I and 2, other Technical Specifications provide limits for operation that can be more restrictive than or can supplement these P/T limits. LCO 3.2.5, "DNB Parameters"; LCO 3.2.3.1, "RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor"; LCO 3.1.1.4, "Minimum Temperature for Criticality"; and Safety Limit 2.1, "Safety Limits," also provide operational restrictions for pressure and temperature and maximum pressure. Furthermore, MODES 1 and 2 are above the temperature range of concern for nonductile failure, and stress analyses have been performed for normal maneuvering profiles, such as power ascension or descent.

ACTIONS Operation outside the P/Trlimits must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Allowed Outage Times (AOTs) 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 MILLSTONE 0988

- UNIT 3 B 3/4 4-10 Amendment No. 797, Wf7,

3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensures that the Secondary System pressure will be limited to within 110% (1305 psig) of its design pressure of 1185 psig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a of Turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure on Code, 1971 Edition. The design minimum total relieving capacity for all valves all of the steam lines is 1.579 X 10 lbs/h which is 105% of the total secondary 7

steam flow of 1.504 X 107 lbs/h at 100% RATED THERMAL POWER.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in Secondary Coolant System steam flow and THERMAL POWER required by the reduced Reactor trip settings of the Power Range Neutron Flux channels. The Reactor Trip Setpoint reductions are derived on the following bases:

1

=(1001/) (wKhfgN Hi K where:

Hio = Safety Analysis power range high neutron flux setpoint, percent Q = Nominal NSSS power rating of the plant (including reactor coolant pump heat), Mwt K = Conversion factor, 947.82 (Btu/sec)

Mwt hfg = heat of vaporization for steam at the highest MSSV opening pressure including tolerance (+/- 3%) and accumulation, as appropriate, Btu/lbm N Number of loops in plant MILLSTONE 0989

- UNIT 3 B 3/4 7-1 Amendment No. 197,