ML20247G684

Proposed Tech Specs,Modifying TSs 3.3.1.1 & 3.3.2.1 to Restrict Time Most Reactor Protection or Esfa Channels Can Be in Bypass Position to 48 Hours,From Indefinite Period of Time

ML20247G684
Person / Time
Site:	Millstone
Issue date:	05/14/1998
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20247G670	List: B16186, Application for Amend to License DPR-65,modifying TSs 3.3.1.1 & 3.3.2.1 to Restrict Time Most Reactor Protection or Esfa Channels Can Be in Bypass Position to 48 Hours,From Indefinite Period of Time ML20247G684
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NUDOCS 9805200319
Download: ML20247G684 (72)

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Docket No. 50-336 B16186 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protective and Engineered Safety Feature Actuation System instrumentation Discussion of Proposed Changes May 1998 9905200319 990514 PDR ADOCK 05000336 P

PDR

U. S. Nucirr R:gulttory Commission B16186/ Attachment 1/Page 1 Proposed Revision to Technical Specifications Reactor Protective and Engineered Safety Feature Actuation System Instrumentation Discussion of Proposed Changes introduction Northeast Nuclear Energy Company (NNECO) hereby proposes to amend Operating License DPR-65 by incorporating the attached proposed changes into the Technical Specifications of Millstone Unit No. 2.

The proposed changes modify Technical Specifications 3.3.1.1, " Reactor Protective Instrumentation" and 3.3.2.1, " Engineered Safety Feature Actuation System Instrumentation" to restrict the time most reactor protection or engineered safety feature actuation channels can be in the bypass position to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, from an indefinite period of time.

NNECO is also proposing to modify the Technical Specification action requirement for the loss of turbine load reactor trip function, the channel calibration requirements for the loss of turbine load reactor trip function and the wide range logarithmic neutron flux monitors, add a note to exclude neutron detectors from channel calibration requirements, correct a reference to a surveillance requirement, and correct errors contained on Technical Specification Page 2-4.

Desian Basis and Licensina Basis The Reactor Protection System (RPS) protects the reactor core from clad failure or fuel melting, and protects the Reactor Coolant System (RCS) pressure boundary by initiating a reactor shutdown if certain plant parameter values are exceeded. The RPS i

uses four redundant channels to monitor each of these plant parameters.

Each channel is physically and electrically independent from the other three. If any two of

)

the four channels (2 of 4 coincidence) monitoring a plant parameter exceed the allowable value for that parameter, a reactor trip is initiated.

The Engineered Safety Feature Actuation System (ESFAS) continuously monitors plant parameters for accident indications. The ESFAS will actuate plant systems to protect the public from the release of radioactivity if these plant parameters indicate an accident has occurred. The ESFAS uses four redundant channels to monitor each of l

these plant parameters. Each channel is physically and electrically independent from l

the other three. If any two of the four channels (2 of 4 coincidence) monitoring a plant parameter exceed the allowable value for that parameter, the required plant system (s) is(are) actuated to mitigate the consequences of the accident. (This discussion does not apply to the ESFAS automatic closure of the containment purge isolation valves on high radiation whicn uses a 1 of 4 actuation logic.)

U. S. Nucl=r Regul: tory Commission B16186/ Attachment 1/Page 2 Currently, the Millstone Unit No. 2 Technical Specifications allow one of the four RPS or ESFAS channels, monitoring a given parameter, to be bypassed for an indefinite period of time. When in the bypass position, the RPS or ESFAS trip logic requires two of the remaining three channels (2 of 3 coincidence) to sense the plant parameter exceeding the allowable value, before a protection action is initiated.

For older CE plants, designed and built to earlier, less stringent channel isolation / separation requirements, there is a remote possibility a failure in one RPS or ESFAS channel could propagate to a second channel. With two channels inoperable due to this one failure, and a third channel bypassed, there would only be one operable RPS or ESFAS channel. The one operable channel would not be able to initiate a protective action, if required, since at least two trip signals are needed. Newer CE plants are not susceptible to this because they have been built to more stringent channel isolation and separation design requirements (e.g., location and physical separation of sensors, instrument lines, and transmitters; electric cables routed through conduit and covered cable trays in containment, etc.).

The NRC required, in a letter dated March 31,1982,' that older CE plants, including Millstone Unit No. 2, either submit a Technical Specification change to remove the ability to bypass an RPS or ESFAS channel for an indefinite period of time, or provide sufficient justification that the respective plant is not susceptible to this failure mechanism. In a letter dated June 25,1984, Millstone Unit No. 2 provided justification for continued operation with the ability to bypass an RPS or ESFAS channel for an indefinite period of time.

As a result of the review of the RPS and ESFAS conducted during the 10CFR50.54(f) response project, it has been determined that Millstone Unit No. 2 did not provide adequate justification to allow continued use of indefinite bypass operation for an RPS or ESFAS channel. Therefore, NNECO proposes to modify the Millstone Unit No. 2 Technical Specifications 3.3.1.1 and 3.3.2.1 to remove the capability for indefinite bypass operation for all reactor trip channels except high pressurizer pressure, and for all engineered safety feature (ESF) actuation channels except containment purge valve isolation and containment sump recirculation.

All automatic reactor trip functions and ESFs were evaluated to determine if the proposed change to the associated action statement (Table 3.3-1 Action 2 and Table 3.3-3 Action 2) could result in equipment actuations that would be detrimental to plant R. A. Clark letter to W. G. Counsil, Modification of NRC Position Conceming Reactor Protection System inoperable Channel Condition at Millstone Nuclear Power Station, Unit No. 2, dated March 31,1982.

W. G. Counsil letter to J. R. Miller, " Millstone Nuclear Power Station, Unit No. 2, Reactor Protection and Engineered Safeguards System Actuation Logic," dated June 25,1984.

' U. S. Nuclear Regulatory Commission B16186/ Attachment 1/Page 3 safety. The only automatic reactor trip function or ESF of potential concern is the high pressurizer pressure reactor trip function. In addition to initiating a reactor trip when 2 of 4 channels sense a high pressurizer pressure condition, this function will also open both pressurizer power operated relief valves (PORVs). Opening both PORVs, due to the failure of two pressurizer pressure channels -(one channel placed in trip by Technical Specification action statement and then a subsequent failure of another channel), would not be desirable.

However, this would not place the plant in an unanalyzed condition (FSAR Section 14.6.1 analyzes the inadvertent opening of both PORVs), the release of reactor coolant can be terminated by closure of the PORV block valves from the control room, and the Emergency Operating Procedures provide guidance on how to address this situation. Since there is not sufficient justification to exclude the high pressurizer pressure reactor trip function from the change to address isolation / separation issues, but allowing operation to continue for an indefinite period of time with a pressurizer pressure high reactor trip channel in the tripped condition is not desirable, an action statement will be added to Technical Specification 3.3.1.1 that will allow a pressurizer high pressure RPS channel to be inoperable for a maximum of 30 days. If the channel is not repaired within this 30 day period, a plant shutdown will be required. During this 30 day time period, the inoperable pressunzer pressure high reactor protection channel will be maintained in the bypassed condition, instead of the tripped condition as proposed for the other RPS trip functions. This will minimize the risk of initiating a loss of primary coolant due to the inadvertent actuation of the PORVs, as well as the risk associated with fault propagation between channels.

The current action statement for the failure of a containment purge channel (Table 3.3-3 Action 3) will not be changed. This ESF function will occur if 1 of 4 containment radiation channels exceed the actuation setpoint. Therefore, this ESF function is not susceptible to the potential failure mechanisms addressed by this change.

l The current action statement for the failure of a containment sump recirculation channel (Table 3.3-3 Action 4) will not be changed. This ESF function will occur if 2 of 4 refueling water storage tank (RWST) level channels decrease to the actuation setpoint.

Action 4 requires a failed channel to be placed in the bypass, not trip, position. This is appropriate. If the failed channel is placed in the trip position, and a second channel l

fails, an inadvertent sump recirculation actuation signal (SRAS) would be generated. If i

l this were to occur during the initial phases of a loss r) coolant accident (LOCA), a low i

L pressure safety injection pump, required for LOCA mitigation, could be stopped by the SRAS. In addition, in the event of a LOCA or main steam line break, this inadvertent SRAS could open the containment sump recirculation motor operated valves before the RWST inventory is injected into the RCS. If the action statement for this ESF function is changed, a second ESF channel failure could result in an inadvertent SRAS, which would place Millstone Unit No. 2 in an unanalyzed condition (inadvertent SRAS is not an analyzed accident in the FSAR).

The resultant equipment actuations could be detrimental to accident mitigation and plant safety. Therefore, the current action statement (Table 3.3-3 Action 4) will not be changed. The relationship between an

f U. S. Nuciner R:gulttory Commission l

B16186/ Attachment 1/Page 4 l

inadvedent SRAS and Technical Specification action statements is discussed by l

Combustion Eng:neering infobulletin No. 97-02.' The current action statement for the failure of a containment sump recirculation channel (Table 3.3-3 Action 4) was approved by Technical Specification Amendment 179.d Most of the proposed changes are consistent with the Calvert Cliffs model RPS and ESFAS Technical Specifications provided in Enclosure 3 of the NRC conespondence j

dated April 16, 1981.5 The use of the Calvert Cliffs RPS and ESFAS Technical Specifications as a model for changes to the Millstone Unit No. 2 Technical l

Specifications (Option 1) is acceptable, and will not require any further analyses or review by either NNECO or the NRC, as stated in the NRC correspondence dated March 31,1982.*

i Description of Proposed Chanaes Each of the proposed changes is discussed below. Additional background information is included, as necessary, to explain the changes.

Related changes are grouped together. ' However, the marked up pages contained in Attachment 3 are sequenced in numerical order by page number.

RPS/ESFAS Indefinite Bypass The proposed changes to Technical Specifications 3.3.1.1, " Reactor Protective instrumentation," and 3.3.2.1,

" Engineered Safety Feature Actuation System Instrumentation," will restrict the time an RPS or ESFAS actuation channel can be in the bypass position. The proposed changes modify the action statements for Technical Specifications 3.3.1.1 and 3.3.2.1 to require a failed RPS, except the pressurizer pressure high pressure reactor protection channel, or ESFAS actuation channel to be placed in the tripped condition within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of the failure, instead of allowing the failed channel to remain in the bypass position for an indefinite period of time. This change will also allow a second channel to be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, instead of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, provided one of the inoperable channels is placed in the tripped i

condition. Most of these changes are consistent with the Calvert Cliffs RPS and ESFAS Technical Specifications model, as discussed in the Design Basis and l

I Combustion Engineering infobulletin No. 97-02, " Spurious Recirculation Actuation Signal," dated May 23,1997.

G. S. Vissing letter to John F. Opeka, issuance of Amendment 179, dated October 7, 1994.

R. A. Clark letter to W. G. Counsil, Evaluation of the Reactor Protection System inoperable Channel Condition at Millstone Nuclear Power Station, Unit No. 2, dated April 16,1981.

R. A. Clark letter to W. G. Counsil, Modification of NRC Position Conceming Reactor Protection System inoperable Channel Condition at Millstone Nuclear Power Station, Unit No. 2, dated March 31,1982.

i

U. S. Nucicar Rcgulctory Commission B16186/ Attachment 1/Page 5 Licensing Basis section of this attachment, and the new, improved Standard Technical Specifications (STS) for Combustion Engineering plants (NUREG-1432).

1.

Action Statement 2 of Table 3.3-1, " Reactor Protective instrumentation," will be modified to restrict the time an inoperable RPS trip channel can be in the bypass position. This change will require the inoperable channel to be placed in the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, regniess of power level. (The current requirement is immediately 5 5% power, and no time limit specified E 5%

power.) The inoperable channel may remain in the bypassed condition for a maximum of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. After 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the channel must either be restored to operable status, or placed in the tripped condition. Also, one additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, instead of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, provided one of the inoperable channels is placed in the tripped condition.

This change is consistent with the Calvert Cliffs model and NUREG-1432.

However, the words in the Calvert Cliffs model "while performing tests and maintenance on that channel" will not be added. This does not change the

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intent to allow one channel in trip and one channel in bypass for a finite amount of time. Also, no distinction in action requirements will be made based on power level (current wording) or any reference to operating mode (Calvert Cliffs model),

and an exception to Technical Specification 3.0.4 will not be added.

2.

Action 5 will be added to Table 3.3-1 and a "5" will be added to Functional Unit 4 (Pressurizer Pressure High) in this table.

Action 2 will be removed from Functional Unit 4. This change will require the inoperable channel to be placed in the bypassed condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, regardless of power level. (The current requirement is immediately 5 5% power, and no time limit specified 2 5% power.)

The inoperable channel may remain in the bypassed condition for a maximum of 30 days. Also, one additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, instead of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, provided one of the inoperable channels is placed in the tripped condition. This new action statement will require the plant to be placed in Mode 3, where the high pressurizer pressure reactor trip function is not needed, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> if a pressurizer high pressure reactor protection channel is inoperable for greater than 30 days.

If the failed pressurizer high pressure reactor protection channel was placed in the tripped condition after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, as is proposed for most of the RPS and ESFAS trip functions, the high failure of a second pressurizer high pressure reactor protection channel would result in inadvertent PORV actuation.

Therefore, the proposed change will leave a failed pressurizer high pressure reactor protection channel in the bypassed condition.

To limit the risk associated with this configuration, a 30 day restriction on plant operation with an inoperable pressurizer high pressure reactor protection channel will be added. If the inoperable channel is not restored with 30 days, a plant shutdown to Mode 3

U. S. Nucle r Regulatory Commission B16186/ Attachment 1/Page 6 will be required. By restricting the time this configuration will be allowed, the associated risk will be limited to an acceptable value.

3.

Action Statement 2 of Table 3.3-3, " Engineered Safety Feature Actuation System Instrumentation," will be modified to restrict the time an inoperable ESFAS trip channel can be in the bypass position. This change will require the inoperable channel to be placed in the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, regardless of pressurizer pressure. (The current requirement is immediately <

1750 psia, and no time limit specified > 1750 psia.) The inoperable channel may remain in the bypassed condition for a maximum of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. After 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the channel must either be restored to operable status, or placed in the tripped condition. Also, one additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, instead of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, provided one of the inoperable channels is placed in the tripped condition.

This change is consistent with the Calvert Cliffs model and NUREG-1432.

However, the words in the Calvert Cliffs model "while performing tests and maintenance on that channel" will not be added. This does not change the intent to allow orie channel in trip and one channel in bypass for a finite amount of time. Also, no distinction in action requirements will be made based on pressurizer pressure (current wording), and an exception to Technical Specification 3.0.4 will not be added.

4.

The Bases for Technical Specifications 3/4.3.1 and 3/4.3.2, " Protective and Engineered Safety Features (ESF) instrumentation," will be expanded to discuss the proposed changes to the action requirements for the failure of an RPS or ESFAS trip channel.

A discussion of the trip logic for containment purge isolation on high containment radiation, and the associated action statement, will be added to emphasize the differences between this, and other ESFAS actuations. A discussion of the impact of surveillance testing on RPS operability

. will also be added.

Turbine Trip Function

- Action Statement 3 for an inoperable RPS turbine trip channel will be changed to Action Statement 2. ' Action Statement 2 is used by most of the RPS trip functions required in Modes 1 and 2.

Even though the loss of turbine load reactor trip function is not credited in any safety analysis, it is important for plant operation because it minimizes the pressure and temperature response of the RCS following.a loss of turbine load.

Therefore, it should be covered by the same, more restrictive, action statement as most of the other RPS trip functions. A channel calibration will also be required every 18 months for the RPS turbine trip function, to be consistent with most of the other RPS trip functions. These changes are consistent with NUREG-1432.

l l

U. S. Nuclear Regulatory Commission I

B16186/ Attachment 1/Page 7 1.

The Action Statement in Table 3.3-1 for a channel failure of Loss of Turbine -

Hydraulic Fluid Pressure - Low, Functional Unit 10, will be changed from "3" to "2." This change will make the required action consistent with most of the other RPS trip functions.

2.

The requirements of Action Statement 3 of Table 3.3-1 will be deleted and replaced by the words "NOT USED." Action Statement 3 is no longer necessary since the required action for an RPS turbine trip channel failure will be changed to Action Statement 2.

l 3.

The channel calibration requirement in Table 4.3-1, " Reactor Protective Instrumentation Surveillance Requirements," for Loss of Turbine - Hydraulic Fluid Pressure - Low, Functional Unit 10, will be changed from "N.A." to "R."

This change will make the channel calibration requirement consistent with most l

of the other RPS trip functions.

l Neutron Flux Detectors The proposed changes will require a channel calibration of the wide range logarithmic l

neutron flux monitors to be performed every 18 months.

Even though this instrumentation is not associated with any reactor trip function, it does have channel check and channel functional test requirements. Therefore, it should have a channel calibration requirement. The proposed changes will also exclude the wide range logarithmic neutron flux detectors and the linear power range reactor protection flux detectors from the channel calibration requirement. These changes are consistent with NUREG-1432.

1.

A "(5)" will be added to the quarterly channel calibration requirement for Power Level - High Nuclear Power, Functional Unit 2a, in Table 4.3-1. The statement

"(5) - Neutron detectors are excluded from the CHANNEL CALIBRATION." will be added to the table notation for Table 4.3-1.

This note will exclude the neutron detectors from the channel calibration requi ement because the detectors are passive devices with minimal drift, and because of the difficulty in simulating a meaningful signal.

2.

The channel calibration requirement in Table 4.3-1 for Wide Range Logarithmic Neutron Flux Monitor, Functional Unit 11, will be changed from "N.A." to "R(5)."

This change will require a channel calibration, consistent with the requirement to l

perform a channel check and channel function test, and will exclude the neutron l

detectors from the channel calibration requirement, as previously discussed.

3.

The. channel calibration requirement in Table 4.3-6, " Remote Shutdown Instrumentation Surveillance Requirements," for Wide Range Logarithmic Neutron Flux, will be changed from "N.A." to "R*." This change to add a l

_________..______.______E___.

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U. S. Nucirr R guintory Commission B16186/ Attachment 1/Page 8 i

requirement to perform a channel calibration is consistent with the current requirement to perform a channel check. The statement " Neutron detectors are excluded from the CHANNEL CALIBRATION." will be added to Table 4.3-6.

)

This note will exclude the neutron detectors from the channel calibration i

l requirement, as previously discussed.

I i

Miscellaneous items i

The proposed changes will correct errors inadvertently introduced on Technical Specification Page 2-4, Table 2.2-1, " Reactor Protective instrumentation Trip Setpoint Limits," when replacement pages were submitted with license amendment requests.

1 The proposed changes will also correct a reference to a surveillance requirement contained in an action statement.

j i

1.

A "(1)" will be added to Reactor Coolant Pump Speed - Low, Functional Unit 4, in Table 2.2-1, " Reactor Protective Instrumentation Trip Setpoint Limits," Page 2-4.

This will reference the note that states this trip may be bypassed below 5%

l power, provided the bypass is automatically removed at or above 5% power.

l This bypass capability currently exists in the design of the Millstone Unit No. 2 j

RPS, and is the same bypass feature referenced for Reactor Coolant Flow -

1 Low, Functional Unit 3. Both of these reactor trip functions provide protection for i

a reduction in RCS flow.

l A reference to Note 1 for Underspeed - Reactor Coolant Pumps, Functional Unit 11, Page 2-5, was contained in the original NNECO submittal to the NRC dated l

March 2,1979.7 A reference to Note 1 was also contained in a subsequent i

NNECO submittal to the NRC dated March 23,1979,8 which added the setpoint I

for the underspeed trip. A reference to Note 1 for Underspeed - Reactor Coolant Pumps, Functional Unit 11, Page 2-5, was contained in License Amendment No.

52 to Facility Operating License DPR-65.'

In a letter dated August 29,1980,' NNECO proposed to revise the setpoint for the reactor coolant pump underspeed trip. As part of this submittal, NNECO W. G. Counsil letter to U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power Station, Unit No. 2, Proposed Revisions to Techn: cal Specifications," dated March 2, 1979.

W. G. Counsil letter to U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power j

l Station, Unit No. 2, Reactor Coolant Pump Speed Sensing System," dated March 23, 1979.

I R. W. Reid letter to W. G. Counsil, License Amendment No. 52 to Facility Operating l

License DPR-65, dated May 12,1979.

W. G. Counsil letter to U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power l

I Station, Unit No. 2, Proposed Revisions to Technical Specifications," dated August 29, 1980.

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I U. S. Nucl=r Regulatory Commission B16186/ Attachment 1/Page 9 proposed to revise Table 2.2-1 by changing Underspeed - Reactor Coolant Pumps, Functional Unit 11, Page 2-5, to Reactor Coolant Pump Speed - Low, Functional Unit 4, Page 2-4. The replacement pages contained in this submittal did not contain a reference to Note 1 for the underspeed trip. The elimination of the reference to Note.1 was not discussed in the NNECO submittal. It appears that the reference to Note 1 was inadvertently removed when the wording was changed. The reference to Note 1 for Reactor Coolant Pump Speed - Low, Functional Unit 4, Page 2-4, was not contained in License Amendment No. 61 to Facility Operating License DPR 65."

- The elimination of the reference to Note 1 was an administrative error, it was not discussed or justified in either the proposed change to Technical Specifications or the Safety Evaluation Report for License Amendment No. 61.

The addition of this note, which was inadvertently removed, will not result in any technical change to the Millstone Unit No. 2 RPS.

Page 2-4 has been revised by subsequent license amendment requests as discussed below. Attachment 5 contains the history for Technical Specification Pages 2-4 and 2-5, from License Amendment No. 52 to License Amendment 61.

3 2.

The trip setpoint for Power Level - High, Four Reactor Coolant Pumps Operating, Functional Unit 2, in Table 2.2-1 will be modified by adding ", and a maximum of <E 106.6% of RATED THERMAL POWER."

The proposed change will correct an error on Technical Specification Page 24.

The maximum power level - high trip setpoint, four reactor coolant pumps operating, was inadvertently omitted by NNECO in license amendment requests dated November 15,1988' and February 1,1989. These license amendment requests included an incorrect Technical Specification Pape 2-4. This led to the error on Page 2-4 when License Amendment No.139' to Facility Operating License DPR-65 was issued.

l R. A. Clark letter to W. G. Counsil, License Amendment No. 61 to Facility Operating License DPR-65, dated October 6,1980.

12 E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power Station, Unit No. 2, Proposed License Amendment Change, Cycle 10 Reload, (TAC No.

i l

68360)," dated November 15,1988.

'8 E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power Station, Unit No. 2, Proposed Revision to Technical Specifications, Reduced Reactor Coolant System Flow Rate," dated February 1,1989, f

G. S. Vissing letter to E. J. Mroczka, " Issuance of Amendment (TAC NO. 68360),"

l dated March 20,1989.

t l

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U. S. Nucirr Regulatory Commission B16186/ Attachment 1/Page 10 in License Amendment No. 61 to Facility Operating License DPR-65, the trip setpoint for power level - high, four reactor coolant pumps operating, was changed to "s 9.6% above THERMAL POWER, with a minimum setpoint of s14.6% of RATED THERMAL POWER, and a maximum of s106.6% of RATED THERMAL POWER." Additionally, a corresponding change was made to the

" Power Level - High" subsection of Bases Section 2.2.1.

License Amendments 79, 90, and 113 only changed the footnote (*) on Page 2-4, they did not change the trip setpoint for power level - high, four reactor coolant pumps operating.

License Amendment No.139 changed the steam generator pressure - low setpoint and the footnote (*) on Page 2-4. These were the only changes to Page 2-4 that were requested in the corresponding license amendment requests dated November 15,1988 and February 1,1989. However, when License Amendment No.139 was issued, Page 2-4 contained an additional change that inadvertently j

deleted the maximum power level - high trip setpoint, four reactor coolant pumps operating (s106.6% of RATED THERMAL POWER). The elimination of the maximum trip setpoint was an administrative error and was not discussed or l

justified in the Safety Evaluation Report for License Amendment No.139.

The replacement pages for Technical Specification Page 2-4 submitted in the license amendment requests dated November 15,1988, and February 1,1989 were incorrect. This led to the issuance of the replacement page containing the error in License Amendment No.139. The maximum trip setpoint for the power level - high, four reactor coolant pumps operating, should not have been deleted. The maximum trip setpoint is still addressed in the " Power Level - High" subsection of Bases Section 2.2.1. This subsection states:... The trip setpoint has a maximum value of 106.6% of Rated Thermal Power..."

Subsequently, Page 2-4 has been revised by License Amendment No.148 and License Amendment No.199.'7 Neither of these License Amendments changed the power level - high trip setpoint, four reactor coolant pumps operating, and the omission of the maximum trip setpoint was not detected. contains the history for Technical Specification Page 2-4 from License Amendment No. 61 to License Amendment No.199.

R. A. Clark letter to W. G. Counsil, License Amendment No. 61 to Facility Operating l

License DPR-65, dated October 6,1980.

G. S. Vissing letter to E. J. Mroczka, " Issuance of Amendment (TAC NO. 77063),"

dated October 12,1990.

D. G. Mcdonald letter to T. C. Feigenbaum, " Issuance of Amendment (TAC NO.

M94466)," dated July 2,1996.

U. S. Nucliar R:gul tory Commission B16186/ Attachment 1/Page 11 This issue was disec,vered during a review of the Technical Specifications for the Millstone Unit 2. Improved Technical Specifications project. It was documented by Adverse Condition Report (ACR) M2-96-0210. A deportability evaluation was performed. This error _ was determined to be not reportable because the surveillance procedure was not changed as a result of Technical Specification Amendment No.139. Therefore,.the maximum power level - high trip setpoint, four reactor coolant pumps operating, was always verified to be < 106.6%.

3.

. The surveillance requirement (Specification 4.3.2.1) referenced in Action 4.b.1 of '

Table 3.3-3 does not currently exist in the Millstone Unit No. 2 Technical Specifications. _ The reference will be changed to Specification 4.3.2.1.1.

Safety Summary The proposed changes will:

1.

Restrict the time most'of the reactor protection or engineered safety feature actuation channels can be in the bypass position to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, from an indefinite period of time.

2.

Allow a second channel to be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, instead of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, provided one of the inoperable channels is placed in the tripped condition.

3.

Apply a more restrictive action statement to the loss of turbine load reactor trip function.

4.

Require a channel calibration every 18 months for the loss of turbine load reactor trip function and for the wide range logarithmic neutron flux monitors.

1 5.

Exclude neutron detectors from the channel calibration requirement.

6.

Correct a reference to a surveillance requirement and errors on Technical Specification Page 2-4.

7.

Expand the Bases of the Technical Specifications to explain how the RPS or ESFAS are affected by the proposed changes and to discuss the impact of surveillance testing on RPS operability.

Restricting the time a reactor protection or engineered safety feature actuation channel can be in the bypass position to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, from an indefinite period of time, has no effect on how the RPS or the ESFAS operates.

This change does not reduce 1

operability or surveillance requirements for any reactor protection or engineered safety

U. S. Nucl=r R gulatory Commission B16186/ Attachment 1/Page 12 feature actuation channel. Therefore, the RPS and ESFAS will continue to function as designed to mitigate design basis accidents. This proposed change is consistent with the Calvert Cliffs RPS and ESFAS Technical Specifications model provided in of the NRC correspondence dated April 16, 1981,'8 and with NUREG -

1432.

i Allowing a failed pressurizer pressure high RPS channel to be placed in bypass for a maximum of 30 days will minimize the risk of inadvertent opening of both PORVs if a second pressurizer high pressure reactor protection channel failed while the first channel was in the tripped condition.

This will also minimize the risk of fault propagation between channels associated with the issue of channel isolation / separation. Therefore, the RPS will continue to function as before.

Millstone Unit No. 2 is currently allowed to remove a second RPS or ESFAS channel from service, provided one of the inoperable channels is placed in the tripped condition. This change will continue to allow a second channel to be removed from service. However, the length of time this configuration is allowed, will be increased from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The increase in time to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> may be a benefit to plant safety by reducing the probability of inadvertent protection system actuations and by providing greater flexibility in performing maintenance and/or testing of an inoperable channel. Even though the length of time has been increased, the RPS and ESFAS will continue to function as before.

(Consistent with the Calvert Cliffs model and NUREG-1432.)

Applying a more restrictive action statement to the loss of turbine load reactor trip function does not reduce the operability requirements for this reactor protection function. Therefore, the RPS will continue to function as before. (Consistent with NUREG-1432.)

Requiring a channel calibration every 18 months for the loss of turbine load reactor trip function and for the wide range logarithmic neutron flux monitors does not reduce the operability req'uirements for these reactor protection functions. Channel calibrations are being performed on these functions, even though they are not currently required by Technical Specifications.

Therefore, the RPS will continue to function as before.

(Consistent with NUREG-1432.)

Excluding the neutron detectors from the channel calibration requirement is acceptable because the detectors are passive devices with minimal drift, and because of the difficulty in simulating a meaningful signal. In addition, slow changes in the sensitivity of the linear power range flux detectors is compensated for by performing the daily R. A. Clark letter to W. G. Counsil, Evaluation of the Reactor Protection System Inoperable Channel Condition at Millstone Nuclear Power Station, Unit No. 2, dated April 16,1981.

U. S. Nucinr Regulttory Commission B16186/ Attachment 1/Page 13 calorimetric. calibration and the monthly calibration using the incore detectors.

Therefore, the RPS will continue to function as before. (Consistent with NUREG-1432.)

Changing the surveillance requirement (Specification 4.3.2.1) referenced. in Action 4.b.1 of Table. 3.3-3 to Specification 4.3.2.1.1 will correct a reference to an item that does not currently exist in the Millstone Unit No. 2 Technical Specifications. This will j

have no effect on ESFAS operation. Therefore, the ESFAS will continue to function as before.

Correcting the errors on Technical Specification Page 2-4 have no affect on how the plant is operated. Adding a reference to the reactor coolant pump low speed reactor i

trip function to a note that states this trip may be bypassed when < 5% power,'and that j

the bypass must be automatically removed when 15% power will not affect this reactor j

trip function. This bypass capability currently exists in the design of the Millstone Unit

' No. 2 RPS, and is the same bypass feature referenced for the reactor coolant flow low -

reactor trip function.

Both of these reactor trip functions provide protection for a reduction in RCS flow. The reference to this bypass feature for the underspeed' trip was contained in License Amendment No. 52, and inadvertently omitted in the NNECO submittal that resulted in the issuance of License Amendment No. 61.

License Amendment No. 61 also changed the power level high trip setpoint to the current value.

The power level high trip'setpoint was inadvertently modified in the NNECO submittals that resulted in the issuance of License Amendment No.139. The setpoint change caused by this error was not requested by NNECO. No surveillance procedure that would have been affected by this error was changed when Technical Specification l

Amendment No.139 was issued. Therefore, the maximum power level - high trip setpoint, four reactor coolant pumps operating, was always verified to be < 106.6%.

The addition of the reference-to the note and the setpoint change will not result in any technical change to the Millstone Unit No. 2 RPS. Therefore, the RPS will continue to function as before.

Expanding the Bases of the Technical Specifications to provide a discussion of how the RPS and ESFAS are affected by the proposed changes, the effect the action statements have on the operation of the RPS and ESFAS, and to discuss the impact of

. surveillance testing on RPS operability will have no effect on equipment operation.

Therefore, the RPS and ESFAS will continue to function as before.

The proposed changes have no effect on how the RPS and ESFAS function to mitigate the consequences of design basis accidents. Therefore, there is no significant impact on the public health and safety.

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Docket No. 50-336 B16186 i

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Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protective and Engineered Safety Feature Actuation System Instrumentation Significant Hazards Consideration i

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U. S. Nucle *r Regulatory Commission B16186/ Attachment 2/Page 1 Proposed Revision to Technical Specifications -

Reactor Protective and Engineered Safety Feature Actuation System instrumentation.

Significant Hazards Consideration Significant Hazards Consideration In accordance with 10CFR50.92, NNECO 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 10CFR50.92(c) are not compromised. The proposed changes do not involve an SHC because the changes i

L would not.

l 1.

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

The proposed change to restrict the time.most of the reactor protection or engineered safety feature actuation channels can be in the bypass position to 48

{

hours, from an indefinite period of time, has no effect on the design.of the j

Reactor Protection System (RPS) or the Engineered Safety Feature Actuation

.i System (ESFAS), and does not affect.how these systems operate. In addition, this will minimize the susceptibility of these systems to the remote ' possibility of L

fault propagation between channels. The pressurizer high pressure reactor protection channels will not be required to be placed in the tripped condition

. after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. A failed pressurizer high pressure channel will be allowed to i

i remain in the bypassed condition for up to 30 days. If the failed pressurizer high pressure channel was placed in the tripped condition, and then a high failure of another pressurizer high pressure channel occurred, the reactor would trip and -

both pressunzer power operated relief valves (PORVs) would open, resulting in an undesired loss of primary coolant. Limiting the time that a failed pressurizer high pressure reactor protection channel can be in bypass to 30 days will minimize

' the risk of the inadvertent opening of both PORVs, as well as the risk associated with fault propagation between channels.

These systems will still function as designed to mitigate design basis accidents. Therefore, this change does not significantly increase the probability or consequences of an accident previously evaluated.

The proposed change to increase the time a second RPS or ESFAS channel can be removed from service (from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />), provided one of the inoperable channels is placed in the tripped condition, has no_ effect on the design of the RPS or ESFAS and does not affect how these systems operate.

These systems will still function as designed to mitigate design basis accidents.

However, one of the proposed changes will allow two pressurizer pressure reactor protection channels to be removed from service (one channel in the tripped condition and one channel in the bypassed condition) for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> instead of the current 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time limit. With a pressurizer pressure channel in the tripped

U. S. Nucirr Regulatory Commission B16186/ Attachment 2/Page 2 condition, the high failure of a second pressunzer pressure channel would initiate a reactor trip, open both pressunzer PORVs, and cause an undesired loss of primary coolant. Thus, this change will increase the probability of occurrence of a previously evaluated accident (FSAR Section 14.6.1 - Inadvertent Opening of a Pressurized Water Reactor Pressurizer Pressure Relief Valve). However, since this configuration will only be allowed for an additional 46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br />, the increase in the probability of occurrence of a previously evaluated accident will be limited to an acceptable value. Therefore, this change dr#ss not significantly increase the probability or consequences of an accident previously evaluated.

The proposed change to apply a more restrictive action statement to the loss of turbine load reactor trip function has no effect on the design of this trip function and does not affect how this trip function operates. Also, this trip function is not assumed to operate to mitigate any design basis accident. Therefore, this change does not significantly increase the probability or consequences of an accident previously evaluated.

The proposed change to require a channel calibration every 18 months for the loss of turbine load reactor trip function and for the wide range logarithmic neutron flux monitors has no effect on the design of either the loss of turbine load reactor trip function or the wide range logarithmic neutron flux monitors.

Also, neither of these are assumed to operate to mitigate any design basis accident. Therefore, this change does not significantly increase the probability or consequences of an accident previously evaluated.

The proposed change to exclude the neutron detectors from the channel calibration requirement has no effect on the design of the neutron detectors and has no significant effect on how these detectors operate. The detectors are passive devices with minimal drift. In addition, slow changes in the sensitivity of the linear power range flux detectors is compensated for by performing the daily calorimetric calibration and the monthly calibration using the incore detectors.

These detectors will still function as designed to mitigate design basis accidents.

Therefore, this change does not significantly increase the probability or l

consequences of an accident previously evaluated.

l The proposed change to correct the surveillance requirement referenced in an j.

action statement has no effect on the design of the ESFAS and does not affect how this system operates. The ESFAS will still function as designed to mitigate

(

design basis accidents. Therefore, this change does not significantly increase the probability or consequences of an accident previously evaluated.

The proposed change to add a reference to the reactor coolant pump low speed reactor trip function to a note that states this trip may be bypassed when < 5%

l power, and that the bypass must be automatically removed when > 5% power will not affect this reactor trip function. This bypass capability currently exists in j

the design of the Millstone Unit No. 2 RPS, and is the same bypass feature l

referenced for the reactor coolant flow low reactor trip function. Both of these

- U. S. Nucl:cr R:gulatory Commission B16186/ Attachment 2/Page 3 reactor trip functions provide protection for a reduction in RCS flow.

The addition of this note will not result in any technical change to the Millstone Unit No. 2 RPS. The RPS will continue to function as before. Therefore, this change does not significantly increase the probability or consequences of an accident previously evaluated.

The pruposed diange to correct the power level high trip setpoint on Technical 1

I Specification Page 2-4 will not result in any change to the actual plant setpoint for this RPS trip function. As a result of this proposed change, the setpoint liste,d on Page 2-4 will agree with the setpoint previously approved by the NRC, and currently used by the RPS. The change has no effect on the design of the RPS and does not affect how this system operates. The RPS will still function as designed to mitigate design basis accidents. Therefore, this change does not significantly increase the probability or consequences of an accident previously evaluated.

The information added to the Bases of the Technical Specifications to provide a discussion of how the RPS and ESFAS are affected by the proposed changes, the effect the action statements have on the operation of the RPS ar d ESFAS, and to discuss the impact of surveillance testing on RPS operability will have no effect on equipment operation. The RPS and ESFAS will continue to function as designed to mitigate design basis accidents. Therefore, this change does not significantly increase the probability or consequences of an accident previously j

evaluated.

Thus, this License Amendment Request does not impact the probability of an accident previously evaluated nor does it involve a significant increase in the 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 l

type of equipment will be installed) or require any new or unusual operator actions.

They 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. They will not alter I

l assumptions made in the safety analysis and licensing basis. The RPS and the l

ESFAS will still functim as designed to mitigate design basis accidents.

Therefore, these 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 safety analysis assumption.

The proposed changes do not

U. S. Nucinr Rigulatory Commission B16186/ Attachment 2/Page 4 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 RPS or ESFAS function. 'In addition, l

most of the changes are consistent with the Calvert Cliffs RPS and ESFAS Technical Specifications model provided in Enclosure 3 of the NRC correspondence dated April 16, 1981,' and with the new, improved Standard Technical Specifications (STS) for Combustion Engineering plants (NUREG-1432).

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

The NRC hss provided guidance conceming the application of standards in 10CFR50.92 by providing certain examples (March 6, 1986, 51 FR 7751) of amendments that are considered nat hkely to involve an SHC. The changes proposed herein to correct the errors (Power Level - High and Reactor Coolant Pump Speed -

Low) on Technical Specification Page 2-4 and the surveillance requirement referenced in an action statement are enveloped by example (i), a purely administrative change to i

Technical Specifications.

The changes proposed herein to remove the ability to operate with one reactor protection or engineered safety feature actuation channel in the bypass position for an indefinite period of time, to apply a more restrictive action statement to the loss of turbine load reactor trip function, and to require a channel calibration every 18 months for the loss of turbine load reactor trip function and for the wide range logarithmic neutron flux monitors are enveloped by example (ii), a change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications. All other changes proposed herein are not enveloped by a specific example.

As described above, this License Amendment Request does not significantly impact the probability of an accident previously evaluated, does not involve a significant increase in the consequences of an accident previously evaluated, does not create the possibility of a new or different kind of accident from any accident previously evaluated, and does not result in a significant reduction in a margin of safety. Therefore, NNECO has concluded that the proposed changes do not involve an SHC.

R. A. Clark letter to W. G. Counsil, Evaluation of the Reactor Protection System inoperable Channel Condition at Millstone Nuclear Power Station, Unit No. 2, dated April 16,1981.

l Docket No. 50-336 B16186 i

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Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protective and Engineered Safety i

Feature Actuation System instrumentation Marked Up Pages i

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.AJO CffA A'G G ~

i, 3/4.3 INSTRUMENTATION fog Tw A,g/r/g 7,y.%

3/4.3.1 REACTOR PROTECTIVE INSTRUMEMATIO](

ON4V LINITING CONDITION FOR OPERATION 3.3.1.1 bypasses of Table 3.3-1 shall be OPERABLE.As a minimum, the reactor pro l

APPLICABILITY:

As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS

~

i 4.3'.1.1.1 Each reactor protective instrumentation channel shall be 1

demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-1.

C' 4.3.1.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per la months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip l

function shall be demonstrated to be within its limit at least once per 18 l

months.

Neutron detectors are exempt from response time testing.

Each test l

shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels

• column of Table 3.3-1.

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TABLE NOTATION

• With the protective system trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.

(a) Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 15% of RATED THERMAL POWER.

(b) Trip may be manually bypassed b'e1o'w 780 psia when 'all CEAs are fully

~

inserted; bypass shall be automatically removed at or above 780 psia.

(c) Trip may be bypassed below 15%,of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 215% of RATED THERMAL POWER.

(d) Trip does not need to be operable if all the control rod drive mechanisms are de-energized or if the RCS boron concentration is greater than or equal to the refueling : concentration of Specification 3.9.1 (e)

Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

(f) AT Fower input to trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 2 5% of RATED THERMAL POWER.

(

ACTION STATEMENTS ACTION 1 -

With the number of channels OPERABLE one less than required by the Minimum Chanels 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 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and/or open the protective system trip breakers.

ACTION 2 -

With the numb ~er of OPERABLE channels one less than the Total Number of Channels and with the "EP."E POWER livel.

a.

5. 37. of KAltD lHLRMAL POWLM, immeolateiy piace i.he ins >Eribl e 4he""^l in th: bypt!sSd c0nditiGn,^ si5tGri t!.2,.n0pera 3le

-shunel to OPERA 0LE staius prior i.u iiiu casing THEPu".AL POWER abov: 5% Gf RATED THETdiAL HiwER.

-b.

1. 5% of RATED THEFFid FGWER, operation may continue "ith the

-inopor abie cnanMi~ is. U.e byp;;;:d ::nditicn, provided the following conditions are satisfied:

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MILLSTONE - UNIT 2 3/4 3-4 AmendmentNo.J,M,77,J#,h/

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INSERT A - Paae 3/4 3 4 l

a.

The inoperable channel is placed in either the bypassed or tripped condition l

within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The inoperable channel shall either be restored to OPERABLE status, or placed in the tripped condition, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

b.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, all functional units receiving an input from the inoperable channel are also declared inoperable, and the appropriate actions are taken for the affected functional units.

c.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided one of the inoperable channels is placed in the tripped condition.

1 l

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TABLE 3.3-1 (Continued)

ACTION STATEMENTS 1

I 1.

All f etional unit receiving an in t from th ypassed cha 1 are also plac in the b assed condi n.

2.

The Minim hannels OPERABL equirement is met; h er, one addition channel may be-remove from service for p to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surv lance testing p Specification 4.

1.1 pr ided one of the operable channels s

t p aced in the trip d condition.

j ACTION 3 With the number OPERABLE channel one less than the Total Number Channels and wit the THERMAL POWER level:

a.

% of RATED THE POWER, immediately lace the geT noperable channe in the bypassed co 1 tion, restore uSED the inoperable annel.to OPERABLE atus prior to increasing T RMAL POWER above 5 of RATED THERMAL POWER.

)

b.

> 5%

RATED THERMAL P

, power operation m con nue.

ACTION 4 With the number of channels OPERABLE one less.than required by the Minimum Channels OPERABLE requirement, imediately verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, and at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter.

INSE2 MILLSTONE - UNIT 2 3/4 3-5

)

INSERT E - Paae 3/4 3-5 ACTION 5 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may continue provided the following conditions are satisfied; a.

The inoperable channel is placed in the bypassed condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The inoperable channel shall be restored to OPERABLE status within 30 days or the plant shall be placed in Mode 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, all functional units receiving an input from the inoperable channel are also declared inoperable, and the appropriate actions are taken for the affected functional units.

c.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided one of the inopacable channels is placed in the tripped condition.

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1 wteen 2, un TABLE 4.3-1 (Continued)

TABLE NOTATION With reactor trip breaker closed.

k i j

(1)

If not performed in previous 7 days.

l' l

(2) i Heat balance only, above 15% of RATED THERMAL POWER; 8

adjust " Nuclear Power Calibrate" potentiometers to make I

nuclear power signals agree with calorimetric calculation.

14 During PHYSICS TESTS, these daily calibrations of nuclear power and AT power may be suspended provided these calibra-tions are performed upon reaching'each major test power

]

plateau and prior to proceeding to the next major test power plateau.

~'

(3)

Above 15% of RATED THERMAL POWER, recalibrates the excore B

detectors which monitor the AXIAL SHAPE INDEX by using the incore detectors or restrict THERMAL POWER during subsequent operations to < 90% of the maximum allowed THERMAL POWER level with the existing Reactor Coolant Pump combination.

'~~

(4)

Above 15% of RATED THERMAL POWER, adjust " AT Pwr Calibrate"

(-

potentiometers to null " Nuclear Pwr - AT Pwr".

During PHYSCIS TESTS, these daily calibrations of nuclear power and 14 AT power may be susper.ded provided these calibrations are performed upon reaching each major test power plateau and prior to proceeding to the next major test power plateau.

j (5) -

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l le MILLSTONE - UNIT 2 3/4 3-9

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fog y,vw,ggw June 10,1996 INSTRUMENTATION ON@

3 /4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The engineered safety feature actuation system instrumentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.

l APPLICABILITY: As shown in Table 3.3-3.

ACTION:

a.

With an engineered safety feature actuation system instru-mentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, either adjust the trip setpoint to be consistent with the value specified in the Trip Setpoint column of Table 3.3-4 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or declare the channel inoperable and take the ACTION shown in Table 3.3-3.

i b.

With an engineered safety feature actuation system instru-mentation channel inoperable, take the ACTION shown in Table 3.3-3.

g SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each engineered safety feature acutation system instrumen-tation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-2.

4.3.2.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affects by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

4 4

Q,...

j MILLSTONE - UNIT 2 3/4 3-10 Amendment No.198 0247 k

~

Harch 1, 1979

• AC C/MM*E ~

INSTRUMENTATION N I M 8Ad7'7ev O/Ul f SURVEILLANCE REQUIREMENTS (Continued) l 4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of'each ESF i

l function'shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one channel per function i

such that all channels are tested at.least once every N times 18. months l

where H is the total number of redundant channels in a specific ESF tanction as shown in the " Total,N,0..of Chan.ntis". Column of Table 3.3-3.

l 4.3.2.1.4 The trip value shall be such that.the containment purge l

effluent shall not result in calculated concentrations of radioactivity offsite in excess of 10 CFR Part 20, Appendix B. Table II.6 Forghe purposes of calculating this trip value, a x/Q = 5.8 x 10 sec/m phall be used when the systgm is a}igned to purge through the building vent and a X/Q = 7.5 x 10- sec/m shall be used when the system is aligned l

to purge through the Unit 1 stack, the gaseous and aprticulate (Half i

l Lives gre'ater than 8 days) radioactivity shall be asusmed to be Xe-133 l

and Cs-137,5respectively. However, the setpoints shall be no greater than 5 x 10 cpm.

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Millstone Unit No. 2 3/4 3-15 Amendment No. 77. H. 777.198 ama

January i7. i^^4 TABLE 3.3-3 frontinued)

TABLE NOTATION (a) Trip function may be bypassed when pressurizer pressure is < 1750 psta; bypass shall be automatically removed when pressurizer pressure is 11750 psia.

(b) An SIAS signal is first necessary to enable CSAS logic.

~

(c) Trip function may be bypassed below 600 psia; bypass shall be automatically removed at or above 500 psia.

(d) Deleted I

(e) Trip may be bypassed during testing pursuant to special Test Exception 3.10.3.

ACTION STATEMENTS I

i ACTION 1 -

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 COLD SHUTDOWN within the next

{

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

ACTION 2 -

With the number of OPERABLE channels one less than the Total

(

Number of Channels =d ::ith th; prn::rt:Or prn::=:

1700 yain; immeciaisis ylse the ti.eperetle et==:1 t r.-

th: typ::::d unditis; re; tere the is.eperible cher.r.ei t:

CPETRE status prior to increasing me pressuritur pr::::= :,t::: 1750 p;ie.

=t.

1 1700 dune;p;ie, operation may continuetith

+'t he; rtble in the byy...e e..dities, provided the following conditions are satisfied:

-1.

All !=;tienel unit....eivi v.o in.; fr the v

by; = ::d d - :1 ;re ei;; pie;ed is the typese.d

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aditis.

(I4SE.R T A

63 i

MILLSTONE. UNIT 2 3/4 3-16 Amendment No. JJJ,17Jeh 0210

l l

lNSERT B - Pace 3/4 3-16 l

l l

a.

The inoperable channel is placed in either the bypassed or tripped condition l

within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The inoperable channel shall either be restored to OPERABLE i

status, or placed in the tripped condition, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

1 b.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, all functional units receiving an input from the inoperable channel are also declared inoperable, and the appropriate actions are taken for the affected functional units.

c.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided l

one of the inoperable channels is placed in the tripped condition.

l r

l l

l l

- Jar.uari 17,1000 TABLE 3.3-3 (Continuedi 1

ACTION 3 -

With less than the minimum channels OPERABLE the containment purge valves are to be maintained closed.

ACTION 4 -

With the number of OPERABLE channels one less than the Total Number of Channels and with the pressurizer pressure:

a.

< 1750 psia: insnediately place the inoperable channel in the bypassed condition; restore the inoperable channel to OPERABLE status prior to increasing the pressurizer pressure above 1750 psia.

b.

1 1750 psia, operation may continue with the inoperable channel in the bypassed condition, provided the following condition is satisfied:

1.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service 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

.2.1 provided B,QIli of the inoperable channels re placed in the bypassed condition.

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l MILLSTONE - UNIT 2 3/43-17 AmendmentNo.177.J/),M/

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- Ane 10,190s 3/4.3 INSTRUMENTATION BASES

~y 3/4.3.1 AND 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF) INSTRUMENTATION The OPERABILITY of the protective and ESF instrumentation systems and bypasses ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof exceeds its setpoint, 2) the specified coincidence logic is maintained,

3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available for protective and ESF purposes from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundance and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.

4 4 xset rh, e

The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum p

frequencies p eff tent to demonstrate this capability.

4

\\Q Kmtser b The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable. The Reactor Protective and Engineered Safety Feature response times are contained in the Millstone Unit No. 2 Technical Requirements Manual. Changes to the Technical Requirements Manual require a 10CFR50.59 review as well as a review by the Plant ' Operations Review Committee.

The containment airborne radioactivity monitors (gaseous and particulate)

.are provided to initiate closure of the containment purge valves upon detection of high radioactivity levels in the containment. Closure of these valves prevents excessive amounts of radioactivity from being released to the environs in the event of an accident.

i 7Xe achab~ logie & 1Arr k is.1 ad of % Ack. Sale-ed 3 1

or %b/c 3.3-3 adbener w9wa6/c awe-Qwy ebonnels, o

NILLSTONE - UNIT 2 B 3/4 3-1 AmendmentNo.JF7,177.(([

0251

l l

lNSERT C - Pace B 3/4 3-1 Action Statement 2 of Tables 3.3-1 and 3.3-3 requires an inoperable Reactor l

Protection System (RPS) or Engineered Safety Feature Actuation System (ESFAS) l channel to be placed in the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

The inoperable channel may remain in the bypassed condition for a maximum of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

While in the bypassed condition, the affected functional unit trip coincidence will be 2 out of 3. After 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the channel must either be declared OPERABLE, or placed in I

the tripped condition. If the channel is placed in the tripped condition, the affected functional unit trip coincidence will become 1 out of 3. One additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided one of the inoperable channels is placed in the tripped condition.

Action Statement 2 of Tables 3.3-1 and 3.3-3 is modified by a note stating the provisions of Technical Specification 3.0.4 are not applicable. This note was added to allow the changing of MODES even though two channels are inoperable, one channel bypassed and one channel tripped. MODE changes in this configuration are allowed to permit maintenance and testing on one of the inoperable channels.

In this configuration, the affected RPS or ESFAS functional unit trip coincidence is 1 out of 2, and the probability of a common cause failure affecting both of the OPERABLE channels during the 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is remote.

Action Statement 5 of Table 3.3-1 will require a failed pressurizer high pressure reactor protection channel to be placed in the bypassed condition within one hour, if the failed channel is not restored to OPERABLE status within 30 days, the plant will be shut down to MODE 3. The 30 day restriction on plant operation with an inoperable pressurizer high pressure reactor protection channel is necessary since the failed channel is placed in the bypassed condition instead of the tripped condition because of the potential inadvertent opening of both pressurizer power operated relief valves (PORVs) if a second pressurizer high pressure reactor protection channel failed while the first channel was in the tripped condition.

i INSERT D - Pace B 3/4 3-1 The surveillance testing verifies OPERABILITY of the Reactor Protection System (RPS) by overlap testing of the four interconnected modules: measurement channels, I

bistable trip units,/ RPS logic, and reactor trip circuit breakers. When testing the measurement channels or bistable trip units that provide an automatic reactor trip l

function, the associated RPS channel will be removed from service, declared j

inoperable, and Action Statement 2 of Technical Specification 3.3.1.1 entered. When testing the RPS logic (matrix testing), the individual RPS channels will not be affected.

1 Each parameter within each RPS channel supplies three contacts to make up the 6

)

different logic ladders / matrices (AB, AC, AD, BC, BD, and CD). During matrix testing, only one logic matrix is tested at a time. Since each RPS channel supplies 3 different logic ladders, testing one ladder matrix at a time will not remove an RPS channel from the overall logic matrix. Therefore, matrix testing will not remove an RPS channel from service or make the RPS channel inoperable. It is not necessary to enter an action statement while performing matrix testing. This also applies when testing the reactor j

trip circuit breakers since this test will not remove an RPS channel from service or i

make the RPS channelinoperable, i

10/7/94 IN_STRUMENTATION T:

fcR 2^Ho+v+rzen avcy BASES I

1 3/4.3.1 AND 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF)

I INSTRUMENTATION (Continued) l The maximum allowable trip value for these monitors corresponds to I

calculated concentrations at the site boundary which would not exceed the concentrations listed in 10 CFR Part 20, Appendix B, Table II. Exposure for a year to the concentrations in 10 CFR Part 20, Appendix B.

Table corresponds to l

a total body dose to an individual of 500 mrem which is well below the guidelines l

of 10 CFR Part 100 for an individual at any point on the exclusion area boundary for two hours.

Detennination of the monitor's trip value in count.s per minute, which is the actual instrument response, involves several factors including:

1) the atmospheric dispersion (x/Q), 2) isotopic composition of the sample, 3) sample flow rate, 4) sample collection efficiency, 5) counting efficiency,*and 6)*the background radiation level at the detector.

The x/Q of 5.8 x 10 sec/m is the highest annual average x/Q estimated for the site boundary (0.48 miles in the NE sector) for vent releases from the containment and 7.5 x 10* sec/m' is the highest annual average x/Q estimated for an off-site location (3 miles in the NNE sector) for releases from the Unit I stack.

This calculation also assumes that the isotopic composition is xenon-133 for gaseous radioactivity and cesium-137 for particulate radioactivity (Half Lives greater than 8 days).

The upper limit of 5 x 10' cpm is approximately 90 percent of full instrument scale.

SRAS teate Modification Action Statement 4 of Table 3.3-3, which applies only to the SRAS logic, specifies that during surveillance testing the second inoperable channel must also be placed in the bypassed condition. For the SRAS logic, placing the second inoperable channel in the tripped condition (as in Action Statement 2) could result in the false generation of a SRAS signal due to an addi.tional failure which causes a trip signal in either of the remaining channels at the onset of a LOCA.

The false generation of the SRAS signal leads to unacceptable consequences for LOCA mitigation.

With Action Statement 4, during the two-hour period when two channels are bypassed, no additional failure can result in the false generation of the SRAS signal. However, an additional failure that prevents a trip of.either of the two remaining channels may prevent the generation of a true SRAS signal while in this Action Statement.

If no SRAS is generated at the appropriate time, operating proceduresinstruct the operator to ensure that the SRAS actuation occurs when the refueling water storage tank level decreases.

Due to the limited period of vulnerability, and the existence of operator requirements to manually initiate an SRAS if an automatic initiation does not occur, this risk is considered,

acceptable.

Il 8D l Trias MILLSTONE - UNIT 2 B 3/4 3-2 Amendment No. U7 179 om l

1

Docket No.,9336 li}li19fl Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protective and Engineered Safety Feature Actuation System Instrumentation Retyped Pages i

May 1998

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TABLE 3.3-1 (Continued) o TABLE NOTATION

• With the protective synem trip breakers in the closed position and the CEA drive system capt*,le of CEA withdrawal.

(a) Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 15% of RATED THERMAL POWER.

(b) Trip may be manually bypassed below 780 psia when all CEAs are fully inserted; bypass shall be automatically removed at or above 780 psia.

(c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed k5en THERMAL POWER is 1 15% of RATED THERMAL POWER.

(d) Trip does not need to be operable if all the control rod drive mechanisms are de-energized or if the RCS boron concentration is greater than or equal to the refueling concentration of Specification 3.9.1.

(e) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

(f) AT Power input to trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 15% of RATED THERMAL POWER.

ACTION STATEMENTS ACTION 1 -

With the number of channels OPERABLE one less than required by 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 I

l within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and/or open the protective system trip l

breakers.

1 ACTION 2 -

With the number of OPERABLE channels one less than the Total Number of Channels, operation may continue provided the following conditions are satisfied:

a.

The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The inoperable channel shall either be restored to OPERABLE status, or placed in the tripped condition, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, b.

Within I hour, all functional units receiving an input from the inoperable channel are also declared inoperable, and the appropriate actions are taken for the affected functional

units, c.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided one of the inoperable channels is placed in i

the tripped condition.

MILLSTONE - UNIT 2 3/4 3-4 Amendment No. 7, M U, NJ, Up, 0302

?

s TABLE 3.3-1 (Continued)

ACTION STATEMENTS ACTION 3 -

NOT USED l

ACTION 4 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, immediately verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, and at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter.

ACTION 5 -

With the number of OPERABLE channels one less than the Total Number of Channels, operation may continue provided the following conditions are satisfied:

l a.

The inoperable channel is placed in the bypassed condition within I hour. The inoperable channel shall be restored to OPERABLE status within 30 days or the plant shall be placed in MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

Within I hour, all functional units receiving an input from the inoperable channel are also declared inoperable, and the appropriate actions are taken for the affected functional units.

c.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided one of the inoperable chnnels is placed in the tripped condition, j

f i

MILLSTONE - UNIT 2 3/4 3-5 Amendment No.

0302

IC HND WAE 3

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TABLE 4.3-1 (Continued).

TABLE NOTATION With reactor trip breaker closed.

(1)

If not performed in previous 7 days.

(2)

Heat balance only, above 15% of RATED THERMAL POWER; adjust " Nuclear Power CaltLrate" potentiometers to make nuclear power signals agree with calorimetric calculation.

Cwing PHYSICS TESTS, these daily calibrations of nuclear power and AT power may be suspended provided these calibra-l tions are performed upon reaching each major test power plateau and prior to proceeding to the next major test power l

plateau.

(3) - Above 15% of RATED THERMAL POWER, recalibrates the excore detectors which monitor the AXIAL SHAPE INDEX by using the incore detectors or restrict THERMAL POWER during subsequent operations to 190% of the maximum allowed THERMAL POWER level with the existing Reactor Coolant Pump combination.

(4)

Above 15% of RATED THERMAL POWER, adjust "AT Pwr Calibrate" potentiometers to null " Nuclear Pwr - AT Pwr".

During PHYSICS TESTS, these daily calibrations of nuclear power and AT power may be suspended provided these calibrations are performed upon reaching each major test power plateau and prior to proceeding to the next major test power plateau.

(5)

Neutron detectors are excluded from the CHANNEL CALIBRATION.

l l

MILLSTONE - UNIT 2 3/4 3-9 Amendment No.

0302

l TABLE 3.3-3 (Continued)

TABLE NOTATION (a) Trip function may be bypassed when pressurizer pressure is < 1750 psia; bypass shall be automatically removed when pressurizer pressure is 11750 psia.

(b) An SIAS signal is first necessary to enable CSAS logic.

(c) Trip function may be bypassed below 600 psia; bypass shall be automatically removed at or above 600 psia.

(d) Deleted (e) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

ACTION STATEMENTS ACTION 1 -

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 COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

ACTION 2 -

With the number of OPERABLE channels one less than the Total Number of Channels, operation may continue provided the following conditions are satisfied:

a.

The inoperable channel is placed in either the bypassed or tripped condition within I hour. The inoperable channel shall either be restored to OPERABLE status, or placed in the tripped condition, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

b.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, all functional units receiving an input from the inoperable channel are also declared inoperable, and the appropriate actions are taken for the affected functional units.

c.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided one of the inoperable channels is placed in the tripped condition.

MILLSTONE - UNIT 2 3/4 3-16 Amendment No. J77, J77, J77, 0303

TABLE 3.3-3 (Continued)

ACTION 3 -

With less than the minimum channels OPERABLE the containment purge valves are to be maintained closed.

ACTION 4 -

With the number of OPERABLE channels one less than the Total Number of Channels and with the pressurizer pressure:

a.

< 1750 psia: immediately place the inoperable channel in the bypassed condition; restore the inoperable channel to OPERABLE status prior to increasing the pressurizer pressure above 1750 psia, j

b.

11750 psia, operation may continue with the inoperable channel in the bypassed condition, provided the following condition is satisfied:

d 1.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service 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 l per Specification 4.3.2.1.1 provided B0lli of the inoperable channels are placed in the bypassed condition.

4 l

MILLSTONE - UNIT 2 3/4 3-17 Amendment No. JJp, 777, Jpf, 0300 j

l N

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3/4.3 INSTRUMENTATION BASES 3/4.3.1 AND 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF) INSTRUMENTATION The OPERABILITY of the protective and ESF instrumentation systems and bypasses ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof exceeds its setpoint, 2) the specified coincidence logic is maintained,

3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available for protective and ESF purposes from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundance and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.

Action Statement 2 of Tables 3.3-1 and 3.3-3 requires an inoperable Reactor Protection System (RPS) or Engineered Safety Feature Actuation System (ESFAS) channel to be placed in the bypassed or tripped condition within I hour. The inoperable channel may remain in the bypassed condition for a maximum of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. While in *.he bypassed condition, the affected functional unit trip coincidence will be 2 aut of 3.

After 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the channel must either be declared OPERABLE, or placed in the tripped condition.

If the channel is placed in the tripped condition, the affected functional unit trip coincidence will become 1 out of 3.

One additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided one of the inoperable channels is placed in the tripped condition.

Action Statement 2 of Tables 3.3-1 and 3.3-3 is modified by a note stating the provisions of Specification 3.0.4 are not applicable.

This note was added to allow the changing of M0gES even though two channels are --

inoperable, one channel bypassed and one channel tripped. MODE changes in this configuration are allowed to permit maintenance and testing on one of the I

inoperable channels.

In this configuration, the affected RPS or ESFAS l

functional unit trip coincidence is 1 out of 2, and the probability of a l

common cause failure affecting both of the OPERABLE channels during the l

48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is remote.

Action Statement 5 of Table 3.3-1 will require a failed pressurizer high pressure reactor protection channel to be placed in the bypassed condition within one hour.

If the failed channel is not restored to OPERABLE status within 30 days, the plant will be shut down to MODE 3.

The 30 day restriction on plant operation with an inoperable pressurizer high pressure reactor protection channel is necessary since the failed channel is placed in the bypassed condition instead of the tripped condition because of the potential inadvertent opening of both pressurizer power operated relief valves (PORVs) l MILLSTONE - UNIT 2 8 3/4 3-1 Amendment No. JJ7, Jpp, Jpp, 0305

I 3/4.3 INSTRUMENTATION

~~

s BASES 3/4.3.1 AND 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF) INSTRUMENTATION (continued) if a second pressurizer high pressure reactor protection channel failed while the first.chani,el was in the tripped condition.

The surveillance requirements specified for these systems ensure that the overall. system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimus j

frequencies are sufficient to demonstrate this capability.

The surveillance testing verifies OPERABILITY of the Reactor Protection-System (RPS) by overlap testing of the four interconnected modules:

measurement channels, bistable trip units, RPS logic, and reactor trip circuit breakers. When testing the measurement channels or bistable trip units that provide an automatic reactor trip function, the associated RPS channel will be removed from service, declared inoperable, and Action Statement 2 of Technical Specification 3.3.1.1 entered. When testing the RPS logic (matrix testing),

the 'ndividual RPS channels will not be affected.

Each parameter within each RPS channel supplies three contacts to make up the 6 different logic ladders /

matrices (AB, AC, AD, BC, BD, and CD). During matrix testing, only one logic matrix is tested at a time. Since each RPS channel supplies 3 different' logic ladders, testing one ladder matrix at a time will not remove an RPS channel from the overall logic matrix. Therefore, matrix testing will not remove an RPS channel from service or make the RPS channel inoperable.

It is not necess'iry to enter an action statement while performing matrix testing. This also applies when testing the reactor trip circuit breakers since this test will not remove an RPS channel from service or make the RPS channel inoperable.

The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable. The Reactor Protective and Engineered Safety Feature response times are contained in the Millstone Unit No. 2 Technical Requirements Manual.

Changes to the Technical Requirements Manual require a 10CFR50.59 review as well as a review by the Plant Operations Review Committee.

The containment airborne radioactivity monitors (gaseous and particulate) are provided to initiate closure of the containment purge valves upon

' detection of high radioactivity levels in the containment. Closure of these valves prevents excessive amounts of radioactivity from being released to the environs in the event of an accident. The actuation logic for this function is 1 out of 4.

Action Statement 3 of Table 3.3-3 addresses inoperable containment. purge channels.

. MILLSTONE - UNIT 2 8 3/4 3-la Amendment No. 177, Jpp, J77, 0306

Docket No. 50-336 B16186 i

Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protective and Engineered Safety Feature Actuation System instrumentation History of the Millstone Unit No. 2 Technical Specifications for Page 2-4 from License Amendment No. 52 to License Amendment No.199 l

Page 2-5 from License Amendment No. 52 to License Amendment No. 61 1

l l

l May 1998

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5p4k ' r gy Oy M? - 82 Docket No. 50-336 B16186 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protective and Engineered Safety Feature Actuation System Instrumentation NNECO Commitments i May 1998 U. S. Nuclear Regulatory Commission B16186/ Attachment 6/Page 1 Proposed Revision to Technical Specifications Reactor Protective and Engineered Safety Feature Actuation System instrumentation List of Regulatory Commitments The following table identifies those actions committed to by NNECO in this document. Please notify the Manager - Regulatory Compliance at Millstone Unit No. 2 of any questions regarding this document, or any associated regulatory commitments. { Commitment Committed Date or Outage NONE N/A I