ML052560443

License Amendment Request (LBDCR 05-MP3-1006), Temperature Requirement for the Reactivity Control System Rod Drop Time Test

ML052560443
Person / Time
Site:	Millstone
Issue date:	09/13/2005
From:	Hartz L Dominion, Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
(LBDCR 05-MP3-1006, 05-401, NUREG-1431
Download: ML052560443 (26)

Text

Dominion Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, Virginia 23060 Web Address: www.dom.com September 13, 2005 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 33650 Serial No.05-401 MPS Lic/MAE RO Docket No.

50-423 License No.

NPF-49 DOMINION NUCLEAR CONNECTICUT. INC.

MILLSTONE POWER STATION UNIT 3 TEMPERATURE REQUIREMENT FOR THE REACTIVITY CONTROL SYSTEM ROD DROP TIME TEST LICENSE AMENDMENT REQUEST (LBDCR 05-MP3-0061 Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) hereby requests to amend Operating License NPF-49 for Millstone Power Station Unit 3 (MPS3). The enclosed license amendment request proposes to revise Technical Specification (TS) 3/4.1.3.4, Reactivity Control Systems, Rod Drop Time, Limiting Condition For Operation (LCO) a., by reducing the temperature at which the shutdown and control rod cluster control assemblies (RCCA) drop tests are performed from greater than or equal to 551 QF,rr to greater than or equal to 500QF. The associated TS Bases will be updated to address the proposed changes. The approval of this change will allow greater flexibility in refueling outage scheduling. This change is consistent with the temperature at which the shutdown and control RCCA drop tests are performed in Surveillance Requirement (SR) 3.1.4.3 of NUREG 1431, Standard Technical Specifications -

Westinghouse Plants, published June 2004.

The proposed amendment does not involve a significant impact on public health and safety and does not involve a Significant Hazards Consideration pursuant to the provisions of 10 CFR 50.92.

The Site Operations Review Committee and the Management Safety Review Committee have reviewed and concurred with the determinations. contains description of the proposed TS change and the Significant Hazards Consideration. contains the TS marked-up pages and contains the retyped pages. Attachment 4 contains the marked-up pages of the TS bases for information only. MPS3 TS bases are controlled in accordance with TS Section 6.1 8, Technical Specification Bases Control program.

We request issuance of this amendment no later than December 30, 2006, with the amendment to be implemented within 60 days of issuance to support spring 2007 refueling outage.

Serial No.05-401 Docket No. 50-423 Temperature Requirement For Rod Drop Test Page 2 of 4 In accordance with 10 CFR 50.91 (b), a copy of this license amendment request is being provided to the State of Connecticut.

If you have any questions or require additional information, please contact Mr. Paul R.

Willoughby at (804) 273-3572 Very truly yours, W

Leslie N. Hartz Vice President - Nuclear Engineering Attach men ts:

1. Evaluation of Proposed License Amendment

2. Marked-Up Pages

3. Re-typed Pages

4. Bases Marked-up Pages

5. Millstone Unit 3 Rod Drop Data Commitments made in this letter: None.

Serial No.05-401 Docket No. 50-423 Temperature Requirement For Rod Drop Test Page 3 of 4 cc:

U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1 41 5 Mr. G. F. Wunder Project Manager U.S. Nuclear Regulatory Commission One White Flint North 1 1555 Rockville Pike Mail Stop 08-B-1A Rockville, MD 20852-2738 Mr. S. M. Schneider NRC Senior Resident Inspector Millstone Power Station

Serial No.05-401 Docket No. 50-423 Temperature Requirement For Rod Drop Test Page 4 of 4 COMMONWEALTH OF VIRGINIA

)

1 COUNTY OF HENRICO 1

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Leslie N. Hartz, who is Vice President - Nuclear Engineering, of Dominion Nuclear Connecticut, Inc. She has affirmed before me that she is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of her knowledge and belief.

I,

Acknowledged before me this /3* day of My Commission Expires:

_e/t,2005.

(SEAL)

Serial No.05-401 Docket No. 50-423 ATTACHMENT 1 PROPOSED REVISION TO TECHNICAL SPECIFICATIONS (LBDCR 05-MP3-006)

TEMPERATURE REQUIREMENT FOR THE REACTIVITY CONTROL SYSTEM ROD DROP TIME TEST EVALUATION OF PROPOSED LICENSE AMENDMENT MILLSTONE POWER STATION UNIT 3 DOMINION NUCLEAR CONNECTICUT, INC.

Serial No.05-401 Docket No. 50-423 Rod Drop Time Test Page 1 of 9 PROPOSED REVISION TO TECHNICAL SPECIFICATIONS (LBDCR 05-MP3-006)

TEMPERATURE REQUIREMENT FOR THE REACTIVITY CONTROL SYSTEM ROD DROP TIME TEST EVALUATION OF PROPOSED LICENSE AMENDMENT 1.O DESCRIPTION

3.0 BACKGROUND

3.1 3.2 Reason for Proposed Amendment Description of The Reactivity Control System

4.0 TECHNICAL ANALYSIS

4.1 4.2 Safety Summary Details of the Proposed Amendment

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory RequirementsKriteria

6.0 ENVIRONMENTAL CONSIDERATION

7.0 PRECEDENTS

Serial No.05-401 Docket No. 50-423 Rod Drop Time Test Page 2 of 9 1.O DESCRIPTION Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) hereby requests to amend Operating License NPF-49 for Millstone Power Station Unit 3 (MPS3). The enclosed license amendment request proposes to revise Technical Specification (TS) 3/4.1.3.4, Reactivity Control Systems, Rod Drop Time, Limiting Condition For Operation (LCO) a., by reducing the temperature at which the shutdown and control rod cluster control assemblies (RCCA) drop tests are performed from greater than or equal to 551 pF, to greater than or equal to 5OOQF. The associated TS bases will be updated to address the proposed changes.

2.0 PROPOSED CHANGE

TS 3/4.1.3.4, Reactivity Control Systems, Rod Drop Time, LCO states:

The individual full-length (shutdown and control) rod drop time from the fully withdrawn position shall be less than or equal to 2.7 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a. Tavg greater than or equal to 55I0F, and

b. All reactor coolant pumps operating.

DNC is proposing to change the LCO, as follows:

The individual full-length (shutdown and control) rod drop time from the fully withdrawn position shall be less than or equal to 2.7 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a. Tavg greater than or equal to 5OO0F, and

b. All reactor coolant pumps operating.

3.0 BACKGROUND

3.1 Description of The Reactivity Control System The OPERABILITY (i.e., trippability) of the shutdown and control RCCAs is an initial assumption in all safety analyses that assume RCCA insertion upon reactor trip.

There are 61 RCCAs (control rods) in the reactor core with 193 fuel assemblies. The 61 control rods are divided into five (5) shutdown banks and four (4) control banks.

Shutdown banks provide the necessary reserve negative reactivity, when fully inserted]

to ensure that the reactor is shutdown ( kff

< 1.0) in the event of a reactor trip.

When

Serial No.05-401 Docket No. 50-423 Rod Drop Time Test Page 3 of 9 the reactor is operating, the control banks are used to add or remove negative reactivity to control T,,

by partial insertion into the core or by partial withdrawal from the core.

Verification of RCCA drop times allows the operator to determine that the maximum RCCA drop time permitted is consistent with the assumed RCCA drop time used in the safety analysis. After reactor vessel head removal and replacement during refueling outages, measuring RCCA drop times prior to reactor criticality ensures that the reactor internals and RCCA drive mechanism will not interfere with RCCA motion or RCCA drop time. It also verifies that no degradation in these systems that would adversely affect RCCA motion or drop time has occurred.

3.2 Reason for Proposed Amendment During planning for refueling outages, DNC determined that it is possible to schedule the RCCA drop testing when the reactor coolant temperature reaches 500°F. However, TS 3.1.3.4.a currently requires that the RCCA drop test be performed at 551°F or greater. The RCCA drop testing is a critical path item, and revising the TS and its bases provides operational flexibility by permitting RCCA drop testing to be performed concurrently with other refueling outage tasks performed with reactor coolant greater than or equal to 500°F.

4.0 TECHNICAL ANALYSIS

4.1 Details of the Proposed Amendment The RCCA drop test is intended to provide verification that the actual RCCA drop times are consistent with the RCCA drop times assumed in the safety analysis. The RCCA drop test ensures that the reactor internals and the RCCA drive mechanisms do not interfere with RCCA motion or increase the RCCA drop time, and that no degradation in the system has occurred that would adversely affect the operability of the RCCAs.

The current requirement, to perform the RCCA drop test when the average reactor coolant temperature is greater than or equal to 55IQF, ensures that the measured RCCA drop times will be representative of the conditions that exist at reactor full power operation. The value of 551QF for RCCA drop testing is identical to the MPS3 minimum required temperature for criticality, and RCCA drop testing at 551 QF demonstrates ope rabi Ii ty at operating temperature.

During the evolution of the Westinghouse Standard Technical Specifications (NUREG-1431) an average reactor coolant temperature of greater than or equal to 5OOQF was determined to adequately simulate operating conditions for RCCA drop tests. Data obtained by DNC from the MPS3 RCCA drop time testing during initial (Cycle 1) plant startup support the NUREG-1 431 determination. These data demonstrate that RCCA drop times increase with decreasing reactor coolant temperature, principally because of

Serial No.05-401 Docket No. 50-423 Rod Drop Time Test Page 4 of 9 the increased water density and viscosity at the lower temperatures. Additionally, during the review of a Florida Power and Light license amendment request (Docket Nos. 50-250 and 50-251, submittal dated March 12, 2001) to lower the Turkey Point Units 3 and 4 RCCA drop test temperature requirements, the Nuclear Regulatory Commission (NRC) reviewed Turkey Point Units 3 and 4 initial startup RCCA drop data.

The NRCs safety evaluation (Docket Nos. 50-250 and 50-251, Amendment 208, dated March 12, 2001) states that the data show that there is a slight increase in RCCA drop time as reactor coolant temperature is decreased. provides a tabulation of the RCCA drop times measured during Cycle 1 startup. The parameters for the initial tests were as follows:

Cycle 1, 100% flow, cold conditions (i.e. Tavg =14!jQF, Reactor Coolant System (RCS) pressure = 390 psia) 0 Cycle 1, 100% flow, hot conditions (i.e. Tavg = 557QF, RCS pressure =

2250 psia).

The increase in RCCA drop time due to decrease in temperature from 557QF to 145QF was 0.099 second as demonstrated by RCCA drop results in Cycle 1. This general result was expected, since the viscosity of the water was lower for the hot condition. It is assumed that the Cycle 1 result is typical of current cycles. Additionally, a factor of safety of 1.5 is applied to the increase in RCCA drop time to account for uncertainties.

Therefore, the increase in RCCA drop time due to temperature decrease is conservatively estimated to be 0.1 5 second.

Measured RCCA drop times taken during MPS3, Cycle 10 startup were less than 1.6 seconds, and measuring the RCCA drop time at 500QF is expected to increase this time by less than 0.15 seconds. This results in a conservative drop time estimate at 500QF of approximately 1.75 seconds.

There is sufficiently large margin between the estimated RCCA drop time and the 2.7-second limit in the TS and the 2.19 seconds for surveillance testing acceptance criteria (plant specific seismic allowance of 0.51 seconds). Therefore, there is no change in the acceptance criteria for RCCA drop time.

The RCCA drop time assumption in the safety analysis is not changed, and consequently, the analysis results are not affected.

Based on the above, the available margin in the measured RCCA drop test will accommodate the slight increase in drop times as a result of performing the test at a lower temperature.

4.2 Safety Summary DNC is proposing to change the temperature at which the shutdown and control RCCA drop tests are performed from greater than or equal to 551 QF, to greater than or equal

Serial No.05-401 Docket No. 50-423 Rod Drop Time Test Page 5 of 9 to 5OOQF. This change does not alter any of the assumptions used in the safety analyses, nor will it cause any safety system parameters to exceed their acceptance limit.

The proposed change does not affect the revisions to plant procedures, which were made to address Westinghouse Nuclear Safety Advisory Letter, NSAL-00-016 (Rod Withdrawal from Subcritical Protection in Lower Modes, issued in 2000). NSAL-00-016 indicates that the core should be borated to an all-rods-out condition, when the control rod system is capable of rod withdrawal and the power range trip is not operable, to prevent core criticality from being reached during a postulated uncontrolled rod/bank withdrawal event. Revisions to plant procedures were made such that the shutdown and control banks are incapable of being withdrawn in modes 3, 4 and 5, unless:

The Power Range Neutron Flux - low setpoint trip function is reinstated at a RCS T

greater than or equal to 55loF, or The core is borated to an all-rods-out condition such that criticality is precluded during a postulated uncontrolled rodhank withdrawal accident.

Therefore, based on the above discussion, the proposed change will have no adverse effect on plant safety. Additionally, these changes can be made without adverse impact to plant operations or to the health and safety of the public.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration DNC is proposing to change the temperature at which the shutdown and control RCCA drop tests are performed from greater than or equal to 551 QF,r to greater than or equal to 5OOQF.

DNC has evaluated whether or not a Significant Hazards Consideration (SHC) is involved with the proposed changes by addressing the three standards set forth in 10 CFR 50.92(c) as discussed below.

Criterion 1 :

Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

DNC is proposing to change the temperature at which the shutdown and control RCCA drop tests are performed from greater than or equal to 551 *F, to greater than or equal

Serial No.05-401 Docket No. 50-423 Rod Drop Time Test Page 6 of 9 to 5OOQF. The proposed change does not modify any plant equipment and does not impact any failure modes that could lead to an accident. Additionally, the proposed change has no effect on the consequence of any analyzed accident since the change does not affect the function of any equipment credited for accident mitigation. Based on this discussion, the proposed amendment does not increase the probability or consequences of an accident previously evaluated.

Criterion 2:

Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not modify any plant equipment and there is no impact on the capability of existing equipment to perform its intended functions.

No system setpoints are being modified and no changes are being made to the method in which plant operations are conducted. No new failure modes are introduced by the proposed change.

The proposed amendment does not introduce accident initiators or malfunctions that would cause a new or different kind of accident.

As noted above, the proposed change does not affect the revisions to plant procedures, which were made to address Westing house Nuclear Safety Advisory Letter, NSAL 01 6 (Rod Withdrawal from Subcritical Protection in Lower Modes, issued in 2000).

Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3:

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The TS change does not involve a significant reduction in margin because the acceptance criterion for the RCCA drop time will not change. The proposed change will reduce the minimum RCCA drop test temperature from greater than or equal to 551QF to greater than or equal to 5OOQF. This will slightly increase the measured test RCCA drop time. However, the measured test RCCA drop time is required to remain within the current TS limit of 2.7 seconds and the 2.1 9 seconds for surveillance testing acceptance criteria (plant specific seismic allowance of 0.51 seconds). The proposed change does not affect any of the assumptions used in the accident analysis, nor does it affect any operability requirements for equipment important to plant safety. Therefore, the margin of safety is not impacted by the proposed amendment.

Serial No.05-401 Docket No. 50-423 Rod Drop Time Test Page 7 of 9 In summary, DNC concludes that the proposed amendment does not represent a significant hazards consideration under the standards set forth in 10 CFR 50.92(c).

5.2 Applicable Regulatory RequirementsKriteria DNC is proposing to change the temperature at which the shutdown and control RCCA drop tests are performed from greater than or equal to 551*F, to greater than or equal to 500QF.

10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants, (GDC) contains the following GDCs, which are applicable to the proposed amendment:

GDC 10, Reactor design, which requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

GDC 26, Reactivity control system redundancy and capability, which requires that two independent reactivity control systems of different design principles be provided. GDC 26 also requires that: (1) one of the systems shall use control RCCAs, preferably including a positive means for inserting the RCCAs, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions, such as stuck RCCAs, specified acceptable fuel design limits are not exceeded; (2) the second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded; and (3) one of the systems shall be capable of holding the reactor core subcritical under cold conditions.

GDC 27, Combined reactivity control systems capability, which requires that the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions, and with appropriate margin for stuck RCCAs, the capability to cool the core is maintained.

GDC 28, Reactivity limits, which requires that the reactivity control systems be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than

Serial No.05-401 Docket No. 50-423 Rod Drop Time Test Page 8 of 9 limited local yielding, nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. GDC 28 also requires that these postulated reactivity accidents shall include consideration of RCCA ejection (unless prevented by positive means),

RCCA dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.

Measured RCCA drop times taken during MPS3, Cycle 10 startup were less than 1.6 seconds, and measuring the RCCA drop time at 5OOQF is expected to increase this time by less than 0.15 seconds to become approximately 1.75 seconds. There is sufficient margin between the estimated RCCA drop time and the 2.7-second limit in the TSs.

There is no change in the acceptance criteria for RCCA drop time. Therefore, the above listed criteria are not impacted by the proposed change.

In 10 CFR 50.36, the US. Nuclear Regulatory Commission (NRC or the Commission) established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, technical specifications are required to include items in the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in a plants technical specifications.

The proposed change ensures that Technical Specification 3/4.1.3.4, Reactivity Control Systems, Rod Drop Time, Limiting Condition For Operation (LCO) a., will continue to satisfy the requirements of 10 CFR 50.36.

6.0 ENVl RONMENTAL CONS1 DERATION DNC has determined that the proposed amendment would change requirements with respect to use of a facility component located within the restricted area, as defined by 10 CFR 20, or it would change inspection or surveillance requirements. DNC has evaluated the proposed change and has determined that the change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released off site, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(~)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 PRECEDENTS A Florida Power and Light license amendment request (Docket Nos. 50-250 and 50-251) was submitted on March 12, 2001 to lower the Turkey Point Unit 3 and 4 RCCA

Serial No.05-401 Docket No. 50-423 Rod Drop Time Test Page 9 of 9 drop test temperature requirements. The NRC approved Florida Power and Light license amendment request on May 7,2001 (Amendments 214 and 208).

A similar request was made by Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant Unit 1 (Docket No. 50-315) on June 25, 2004. The NRC approved this request on December 20,2004 (Amendment 284).

Serial No.05-401 Docket No. 50-423 ATTACHMENT 2 PROPOSED REVISION TO TECHNICAL SPECIFICATIONS (LBDCR 05-MP3-006)

TEMPERATURE REQUIREMENT FOR THE REACTIVITY CONTROL SYSTEM ROD DROP TIME TEST MARKED-UP PAGE MILLSTONE POWER STATION UNIT 3 DOMINION NUCLEAR CONNECTICUT, INC.

REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full-length (shutdown and control) rod drop time from the f u l l y withdrawn position shall be less than o r equal t o 2.7 seconds from beginning of decay of stationary gripper coil voltage t o dashpot entry with:

a.

Tavg greater than o r equal t o

b.

All reactor coolant pumps operating.

APPLICABILITY:

MODES 1 and 2.

ACT ION :

a.

With the drop time of any full-length rod determined t o exceed the above limit, restore the rod drop time t o within the above limit prior t o proceeding t o MODE 1 or 2.

b.

With the rod drop times within limits but determined with three reactor cool ant pumps operating, operation may proceed provided THERMAL POWER i s restricted t o less than o r equal t o 65% of RATED THERMAL POWER w i t h the reactor coolant stop valves i n the nonoperat i ng 1 oop cl osed.

SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement p r i o r t o reactor c r i t i c a l i t y :

a.

For a l l rods following each removal of the reactor vessel head,

b.

For specifically affected individual rods following any maintenance on o r modification t o the Control Rod Drive System which could affect the drop time of those specific rods, and c.

A t l e a s t once per 24 months.

MILLSTONE - UNIT 3 3/4 1-25 Amendment No. 99, JZp, 286;,

Serial No.05-401 Docket No. 50-423 ATTACHMENT 3 PROPOSED REVISION TO TECHNICAL SPECIFICATIONS (LBDCR 05-MP3-006)

TEMPERATURE REQUIREMENT FOR THE REACTIVITY CONTROL SYSTEM ROD DROP TIME TEST RE-TYPED PAGE MILLSTONE POWER STATION UNIT 3 DOMINION NUCLEAR CONNECTICUT, INC.

REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 withdrawn position shall be less than or equal to 2.7 seconds &om beginning of decay of stationary gripper coil voltage to dashpot entry with:

The individual full-length (shutdown and control) rod drop time fkom the fully

a.

Tavg greater than or equal to 500°F, and

b.

All reactor coolant pumps operating.

APPLICABILITY:

MODES 1 and2.

ACTION:

a.

With the drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

b.

With the rod drop times within limits but determined with three reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 65% of RATED THERMAL POWER with the reactor coolant stop valves in the nonoperating loop closed.

SURVEILLANCE REQUIREMENTS 4.1.3.4 prior to reactor criticality:

The rod drop time of full-length rods shall be demonstrated through measurement

a.

For all rods following each removal of the reactor vessel head,

b.

For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and

c.

At least once per 24 months.

MILLSTONE - UNIT 3 314 1-25 Amendment No. a, 442,2436,

Serial No.05-401 Docket No. 50-336 ATTACHMENT 4 PROPOSED REVISION TO TECHNICAL SPECIFICATIONS (LBDCR 05-MP3-006)

TEMPERATURE REQUIREMENT FOR THE REACTIVITY CONTROL SYSTEM ROD DROP TIME TEST BASES MARKED-UP PAGES MILLSTONE POWER STATION UNIT 2 DOMINION NUCLEAR CONNECTICUT, INC.

August 27,2001 3/4.1.3 -CONTROL-The specifi&$ions of this section ensure that: (I) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod misalignment 04 associated accident analyses are limited, OPERABILITY of the control rod position indicators is xequird to determine contml rod positions and thereby ensure compliance with the control rnLSToNE - UNIT 3 B 314 1-3 Amendment No. $2, @,'a, 44-3, M3, i64j 197

LBDCR 3-10-02 LBDCR 3-9-02

. August 27,2002 September 4,2002 BASES rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within k12 steps at 24,48,120, and fully withdrawn position for the Control Banks and 18,210, and full withdrawn position for the Shutdown Banks provides assurances that Digital Rod Position Indication System does not indicate the actual shut wn rod position between 18 ste s and 210 steps, only points in the indicated ranges are picked for verification of agreement with B emanded position.

i3 the Digital Rod Position In B icator is operating correctly over the fidl fan e of indication, Since the The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met, Misalignment of a rod requires measurement of peaking factors and a restriction in T H E W POWER. These restrictions provide assurance of fuel rod integrity dwing continued operation, In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the redts rod drop time used in remain valid during future operation.

with all reactor The maximum rod drop t h e restriction is consistent with the safety analyses. Measurement with Tayg greater than or equal coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.

used in the FSAR accident analysis. A rod drop time was calculated to validate the Technical S ecification limit. This calculation accounted for all uncertainties, including a plant specific seismic rod drop time, the acceptance criteria for surveillance testing is 2.19 seconds (References 4 and 5).

The required rod drop time of I 2.7 seconds specified in Technical Specification 3.1.3.4 is a P lowance of 0.5 1 seconds. Since the seismic allowance should be removed when verifying the actual Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is inoperable. These verification fiequencies are adequate for assuring that the applicable LCOs are satisfied.

The Digital Rod Position Indication (DRPI) System is defined as follows:

Rod position indication as displayed on DRPI display panel (MB4), or Rod position indication as displayed by the Plant Process Computer System With the above definition, LCO, 3.1.3.2, ACTION a. is applicable with either DRPI display panel or the plant process computer points OPERABLE.

The plant process computer may be utilized to satisfy DRPI System requirements which Technical Specification SR 4.1.3.2.1 determines each digital rod position indicator to be meets LCO 3.1.3.2, in requiring diversity for determining digital rod position indication.

OPERABLE by verifying the Demand Position Indication System and the DRPI System agree within 12 steps at least once each 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, except during the time when the rod position deviation monitor is inoperable, MILLSTONE - UNIT 3 B 314 1-4 Amendment No. 60, 1

R R

CONTROL SYS-BASES LBDCR 04-MP3-001 2/13/04 then compare the Demand Position Indication System and the DRPI System at least once each 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

The Rod Deviation Monitor is generated only from the DRPI panel at MB4. Therefore, when rod position indication as displayed by the plant process computer is the only available indication, then perform SURVEILLANCE REQUIREMENTS every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Technical Specification SR 4.1.3.2.1 determines each digital rod position indicator to be OPERABLE by verifying the Demand Position Indication System and the DRPI System agree within 12 steps at least once each 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, except during the time when the rod position deviation monitor is inoperable, then compare the Demand Position Indication System and the DRPI System at least once each 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

The Rod Deviation Monitor is generated only fiom the DRPI panel at MB4. Therefore, when rod position indication as displayed by the plant process computer is the only available indication, then perform SURVEILLANCE REQUIREMENTS every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Additional surveillance is required to ensure the plant process cornputex indications are in agreement with those displayed on the DRPI. This additional SWRvEILlLANCE REQUIREMENT is as follows:

Each rod position indication as displayed by the plant process computer shall be determined to be OPBRABLE by verifying the rod position indication as displayed on the DRPI display panel agrees with the rod position indication as displayed by the plant process computer at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The rod position indication, as displayed by DRPI display panel (MB4), is a non-QA system, calibrated on a refueling interval, and used to implement T/S 3.1.3.2. Because the plant process computer receives field data from the same source as the DRPI System (MB4), and is also calibrated on a reheling interval, it fully meets all requirements specified in T/S 3.1.3.2 for rod position.

Additionally, the plant process'computer provides the same type and level of accuracy as the DRPI System (MB4). The plant process computer does not provide any alarm or rod position deviation monitoring as does DRPI display panel (MB4).

I MILLSTONE - UNIT 3 B 314 1-5 Amendment No. 60,

pEACTIVITY CONTROL S YSTEMS BASES February 24,2005 (Continued)

For Specification 3.1 -3.1 ACTIONS b. and c., it is incumbent upon the plant to verify the trippability of the inoperable control rod(s). Trippability is defined in Attachment C to a letter dated December 21,1984, from E. P. Rahe (Westinghouse) to C. 0. Thomas (NRC). This may be by verification of a control system failure, usually electrical in nature, or that the failure is associated with the control rod stepping mechanism, In the event the plant is unable to verify the rod@) trippability, it must be assumed to be untrippable and thus falls under the requirements of ACTION a. Assuming a controlled shutdown from 100% RATED THERMAL POWER, this allows approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for this verification.

For LCO 3.1.3.6 the control bank insertion limits are specified in the CORE OPERATING LIMITS REPORT (COLR). These insertion limits are the initial assumptions in safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions, assumptions of available SHUTDOWN MARGIN, and initial reactivity insertion rate.

The applicable I&C calibration procedure (Reference 1.) being current indicates the associated circuitry is OPERABLE.

There are conditions when the Lo-Lo and Lo alarms of the RIL Monitor are limited below the FUL specified in the COLR. The RIL Monitor remains OPERABLE because the lead control rod bank still has the Lo and Lo-Lo alarms greater than or equal to the RIL.

When rods are at the top of the core, the Lo-Lo alarm is limited below the RIL to prevent spurious alarms. The FUL is equal to the Lo-Lo alarm until the adjustable upper limit setpoint on the FUL Monitor is reached, then the alarni remains at the adjustable upper limit setpoint. When the RIL is in the region above the adjustable upper limit setpoint, the Lo-Lo alarm is below the RIL.

References :

1.

2.

3.

4.

5.

IC 3469N08, Rod Control Speed, Insertion Limit; and ControI TAVE Auctioneered/Deviation Alarms.

Letter NS-OPLS-OPL-1-9 1-226, (Westinghouse Letter NEU-9 1 -563), dated April 24, 199 1, Millstone Unit 3 Technical Requirements Manual, Appendix 8.1, CORE OPERATMG LIMITS REPORT.

Westinghouse Letter NEU-97-298, Millstone Unit 3 - RCCA Drop Time, dated November 13, 1997.

Westinghouse Letter 98NEU-G-0060, Millstone Unit 3 - Robust Fuel Asseiiibly (Design Report) and Generic SECL, dated October 2, 1998.

MILLSTONE - UNIT 3 B 314 1-6 Aiiiendiiieiit No.

LBDCR NO. 04-MP3-015

Serial No.05-401 Docket No. 50-336 ATTACHMENT 5 PROPOSED REVISION TO TECHNICAL SPECIFICATIONS (LBDCR 05-MP3-006)

TEMPERATURE REQUIREMENT FOR THE REACTIVITY CONTROL SYSTEM ROD DROP TIME TEST MILLSTONE UNIT 3 ROD DROP DATA MILLSTONE POWER STATION UNIT 3 DOMINION NUCLEAR CONNECTICUT, INC.

Serial No.05-401 Docket No. 50-423 Rod Drop Time Test Page 1 of 2 Millstone Unit 3 RCCA Drop Time Data Cycle 1 : Initial core/

Drop Time Testing Difference in 100% RCS Flow 100% RCS Flow Drop Times Rod Core Cold Conditions Hot Conditions Cl(cold) - Cl(hot)

Bank Location DroD Time (msec) Drot, Time (msec) 0 SB-A SB-B SB-C SB-D SB-E DO2 B12 M14 PO4 H04 B04 D14 P12 M02 H12 GO3 cog J13 NO7 DO8 C07 G13 N 09 J 03 M 08 E03 c11 L13 NO5 C05 El3 N11 LO3 A07 G15 R09 JO1 1 5 6 1492 1506 1508 1514 1492 1488 1496 1494 1508 1486 1498 1480 1500 1496 1492 1504 1498 1494 1494 1492 1512 1502 1512 1476 1496 1498 1494 1486 1494 1498 1480 1412 1402 1416 1422 1404 1274 1398 1394 1418 1408 1396 1400 1376 1416 1410 1400 1398 1402 1398 1406 1398 1396 1388 1398 1394 1402 1402 1392 1398 1402 1396 1406 88 90 90 86 110 21 8 90 102 76 100 90 98 1 04 a4 86 92 106 96 96 88 94 116 114 114 82 94 96 102 88 92 102 74 Cycle 1 : Initial core

Serial No.05-401 Docket No. 50-423 Rod Drop Time Test Page 2 of 2 Drop Time Testing Difference in 100% RCS Flow 100% RCS Flow Drop Times Rod Core Cold Conditions Hot Conditions Cl(cold) - Cl(hot)

Bank Location Droo Time (msecl Droo Time (msecl lmsec)

CB-A H06 F08 H10 KO8 E05 E l 1 L11 LO5 CB-B F02 B10 K14 PO6 B06 F14 P10 KO2 CB-C H02 B08 H14 PO8 F06 F10 K10 KO6 CB-D DO4 M12 D12 M04 H08 Average Times 1558 1542 1488 1494 1510 1508 1498 1498 1504 1490 1506 1500 1480 1482 1496 1482 1488 1498 1498 1519 1490 1496 1492 1492 1480 1482 1488 1498 1508 1497 1406 1386 1364 1408 1392 1410 1400 1366 1420 1418 1 422 1410 1400 1398 1408 1386 1420 1400 1398 1396 1402 1404 1392 1402 1392 1406 1382 1402 1398 1399 152 156 124 86 118 98 98 132 84 72 84 90 80 84 88 96 68 98 100 1 23 88 92 100 90 88 76 106 96 110 99