B17492, Application for Amend to License DPR-65,modifying TSs 3.3.1.1 & 3.3.2.1 to Restrict Time That Reactor Protection or ESF Actuation Channel Can Be in Bypass Position to 48 H, from Indefinite Period of Time

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Application for Amend to License DPR-65,modifying TSs 3.3.1.1 & 3.3.2.1 to Restrict Time That Reactor Protection or ESF Actuation Channel Can Be in Bypass Position to 48 H, from Indefinite Period of Time
ML20195D396
Person / Time
Site: Millstone Dominion icon.png
Issue date: 11/10/1998
From: Bowling M
NORTHEAST NUCLEAR ENERGY CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20195D402 List:
References
B17492, NUDOCS 9811180031
Download: ML20195D396 (23)
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R pe Ferry Rd. (Route 156), Waterford, CT 06385 Nht:

. Nuclear Energy Minstone Nuclear Power station Northeast Nuclear Erwrgy Company P.O. Box 128 Waterford, CT 06385-0128 (860) 447-1791 l Fax (860) 444-4277

'the Northeast Utilities System Docket No. 50-336 B17492 Re: 10CFR50.90 U. S. Nuclear Regulatory Commission N0y 101998

Attention:' Document Control Desk

[ Washington, DC 20555

Millstane Nuclear Power Station, Unit No. 2 7 Proposed Revision to Technical Specifications p Reactor Protective and Engineered Safety Feature Actuation System Instrumentation b Introductio,n i . . . .

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, Pursuant to 10CFR50.90, Northeast Nuclear Energy Company (NNECO) hereby j proposes to amend Operating License DPR-65 by incorporating the attached proposed j changes into the Technical Specifications of Millstone Unit No. 2. The proposed  ;

j changes modify Technical Specifications 3.3.1.1, " Reactor Protective Instrumentation" l and 3.3.2.1, " Engineered Safety Feature Actuation System instrumentation" to restrict the time a reactor protection or engineered safety feature actuation channel can be in '

I 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.

e 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 E 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 4 requirements, add license amendment numbers to Technical Specification g Page 3/4 3-9, correct a reference to a surveillance requirement, and correct errors

[ contained on Technical Specification Page 2-4.

g Most of these proposed changes were originally submitted to the NRC by NNECO in a letter dated May 14,' 1998.* NNECO'has determined that the previous submittal needs to be modified with respect to the proposed required action for the failure of a pressunzer high pressure reactor protection channel. Since this modification will affect

- the Safety Summary and Significant Hazards Consideration, NNECO is withdrawing the submittal dated May 14,1998, and replacing it with this License Amendment Request.

  • M. L. Bowling, Jr. letter to the NRC, " Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protective and Engineered Safety Feature g

. Actuation System Instrumentation," dated May 14,1998.

9811180031'981110

~PDR -ADOCK 05000336i hq P PM g

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U. S. Nuclur Regulatory Commission l 817492/Pcg3 2 In addition, NNECO is including a non-technical change to add license amendment numbers to one page of the Millstone Unit No. 2 Technical Specifications.

Attachment 1 provides a discussion of the proposed changes and the Safety Summary.

Attachment 2 provides the Significant Hazards Consideration. Attachment 3 provides the marked-up version of the appropriate pages of the current Technical Specifications.

Attachment 4 provides the retyped pages of the Technica: Specifications. Attachment 5 provides the history for Technical Specification Page 2-4, from License Amendment

. No. 52 to License Amendment 199, and Page 2-6, from License Amendment No. 52 to License Amendment 61.

Environmental Considerations NNECO has reviewed the proposed License Amendment Request against the criteria of 10CFR51.22 for environmental considerations. The proposed changes will restrict the time a reactor protection or engineered safety feature actuation channel can be in the bypass position, modify reactor protection channel surveillance requirements, and correct various minor Technical Specification errors. These changes do not significantly increase the type and amounts of effluents that may be released off site. In addition, this amendment request will not significantly increase individual or cumulative occupational radiation exposures. Therefore, NNECO has determined the proposed changes will not have a significant effect on the quality of the human environment.

Conclusions The proposed changes were evaluated utilizing the criteria of 10CFR50.59 and were determined to involve an unreviewed safety question. One of the proposed changes will allow two pressurizer high pressure reactor protection channels to be removed from service (ons channel in the tripped conditicr. 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. In addition, another proposed change will require a failed pressurizer pressure channel to be placed in the tripped condition if not restored to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Both of these changes willincrease the probability of occurrence of a previously evaluated accident. However, these changes will not result in a significant increase in the probability of occurrence of a previously evaluated accident. Therefore, we have concluded that the proposed changes are safe.

The proposed changes do not involve a significant impact on public health and safety (see the Safety Summary provided in Attachment 1) and do not involve a Significant Hazards Consideration pursuant to the provisions of 10CFR50.92 (see the Significant Hazards Consideration provided in Attachment 2).

Plant Operations Review Committee and Nuclear Safety Assessment Board The Plant Operations Review Committee and Nuclear Safety Assessment Board have reviewed and concurred with the determinations.

U. S. Nucl::rr Regulatory Commission B17492/Page 3 r Schedule We request issuance at your earliest convenience, with the amendment to be implemented within 60 days of issuance.

State Notification In accordance with 10CFR50.91(b), a copy of this License Amendment Request is being provided to the State of Connecticut.

There are no regulatory commitments contained within this letter.

If you should have any questions on the above, please contact Mr. Ravi Joshi at (800) 440-2080.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY M. L. Bowling, Jr. t/

Recovery Officer - Technical Services Sworn to and subscribed before me this / day of .1998 25u Votary Public MY COMMISSION EXPlRES My Commission expires JUNE 30. 2009 Attachments (5) cc: See Page 4

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U. S. Nucler R:gulatory Commission B17492/Page 4 cc: H. J. Miller, Region I Administrator D. G. Mcdonald, Jr., NRC Senior Project Manager, Millstone Unit No. 2

. S. Dembek, NRC Project Manager, Millstone Unit No.1 D. P.. Beaulieu, Senior Resident inspector, Millstone Unit No. 2 '

W. M. Dean, Director, Millstone Project Directorate W. D. Lanning, Director, Millstone inspections

~ J. P. Durr, Chief, inspections Branch, Millstone inspections E. V. Imbro, Director, Millstone ICAVP inspections i

Director _

Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 7F Ilm Street H ,rtford, CT 06106-5127 v-

r-1 Docket No. 50-336 i B17492 K

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Attachment 1

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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 Discussion of Proposed Changes I

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November 1998

U. S. Nuclur Reguistory Commission B17492/ Attachment 1/Page 1 Proposed Revision to Technical Specifications Reactor Protective and Engineered Safety Feature Actuation System Instrumentation i

Discussion of Proposed Changes introduction Northeast Nuclear Energy Company (NNECO) hereby proposes to amend Operating i 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 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.

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, add license amendment numbers to Technical Specification Page 3/4 3-9, correct a reference to a surveillance requirement, and correct errors l 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 uses four redLNant channels to monitor each of these plant parameters. Each l channel is physically and electrically independent from the other three. If any two of l 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 l l accident has occurred. The ESFAS uses four redundant channels to monitor each of i

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, the required plant system (s) is(are) actuated to mitigate the consequences of the accident. (This discussion does l not apply to the ESFAS automatic closure of the containment purge isolation valves on high radiation which uses a 1 of 4 actuation logic.)

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U. S Nclur Regulatory Commission B174S2/Attachmnnt 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 protective action is initiated.

For CE plants of the same vintage as Millstone Unit No. 2, 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.

The NRC required, in a letter dated March 31,1982,W that this vintage of CE plants, including Milletone 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 failure mechanism. In a letter dated June 25,1984, Millstonegiant is not Unit No. susceptible 2 provided to this 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 in response to the NRC request pursuant to 10CFR50.54(f) dated April 16, 1997,W and the Independent Corrective Action Verification Program order dated August 14, 1996,* 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, and for all engineered safety features (ESF) actuation channels except containment purge valve isolation and containment sump recirculation.

All automatic reactor trip functions were evaluated to determine if the proposed change to the associated action statement (TE,ble 3.3-1 Action 2) could result in equipment actuations that would be detrimental to plant safety. The only automatic reactor trip function 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 pressuizer pressure condition, this function will also open both pressurizer power operated relief M

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

A 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.

W S. J. Cohins letter to N. S. Cams, NRC request pursuant to 10CFR50.54(f), dated April 16,1997.

W. T. Russell letter to T. C. Felgenbaum, " Confirmatory Order Establishing Independent Corrective i Action Verification Program (Effective immediately) - Millstone Nuclear Power Station, Units 1,2 and 3," dated August 14,1996.

, U. S. Nuctsar Regulatory Commission B17492/ Attachment 1/Page 3 i

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 since FSAR Section 14.6.1

ans'yzes 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 ,

j Emergency Operating Procedures provide guidance on how to address this siietion. l Therefore, there is not sufficient justification to exclude the high pressurizer pressure

. reactor trip function from the change to address isolation /separatior issues.
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i 1t should be noted that the following action statements in Table 3.3-3 will not change.

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! 1. The current action statement for the failure of a containment purge channel j

) (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 i ESF function is not susceptible to the potential failure mechanisms addressed by j this change.

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2. 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, i not trip, position. This is appropriate. If the failed channel is placed in the trip

) position, and a second channel fails, an inadvertent sump recirculation actuation

signal (SRAS) would be generated. If this were to occur during the initial phases of a loss of coolant accident (LOCA), a low pressure safety injection pump,

. required for LOCA mitigation, could be stopped by the SRAS. In addition, in the i event of a LOCA or main steam line break, this inadvertent SRAS could open the ,

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I 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 i .

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 inadvertent SRAS and Technical Specification action statements is discussed by Combustion Engineering i infobulletin No. 97-02.* The current action statement for the failure of a containment sump recirculation channel Technical Specification Amendment ) 179.(Table 3.3-3 Action 4) was appr W 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.

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U. S. Nucis r Regulttory Commission B17492/ Attachment 1/Page 4:

The proposed changes are consistent with the Calvert Cliffs model RPS and ESFAS Technical Specifications provided in Enclosure 3 of the NRC correspondence dated April 16, - 1981.* The use of the Calvert Cliffs' RPS and ESFAS Technical Specifications 'as a model for changes to the Millstone Unit No. 2 Technical 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.(*)

Description of Proposed Chanoes 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 changee 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 state.nents for Technical Specifications 3.3.1.1 and 3.3.2.1 to require a failed RPS 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 condition. These changes are consistent with the Calvert Cliffs RPS and ESFAS Technical Specifications model, as discussed in the Design Basis and 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 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 />, regardless of power level. (The current requirement is immediately < 5% power, and no time limit specified > 5% power.) The j 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 /> for testing or maintenance. 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 M ' 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.

U. S. Nucle:r R:gulatory Commission l B17492/Attrchmsnt 1/Pcg3 5 I hours, provided one of the inoperable channels is placed in the tripped condition.

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

However, the words in the Calvert Cliffs model "while performing tests and maintenance on that channel" will not be adoad. This does not change the intent to allow one channel in trip and one chant el in bypass for a finite amount of time, and is consistent with NUREG-1432. 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 Spscification 3.0.4 will not be added.

2. Action Statement 2 of Table 3.3-3 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 /> for testing or maintenance. 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 consisten'. 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 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 pressurizer pressure (current wording), and an exception to Technical Specification 3.0.4 will not be added.

3. 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 all other RFS trip functions required in

U. S. Nuclear Regulatory Commission B17492/ Attachment 1/Page 6 i l

Modes 1 and 2. Even though the loss of turbine load reactor trip function is not l 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 the l

other RPS trip functions. A channel calibration will also be required every 18 months 1 for the RPS turbine trip function, to be consistent with the other RPS trip functions. ,

These changes are consistent with NUREG-1432. I

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 the other RPS trip .

functions. l

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 I since the required action for an RPS turbine trip channel failure will be changed to Action Statement 2.
3. The channel calibration requirement in Table 4.3-1 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 l with most of the other RPS trip' functions.

Neutron Flux Detectors The proposed changes will require a channel calibration of the wide range logarithmic 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, " Reactor Protective Instrumentation Surveillance Requirements." The statement "(5) -

Neutron detectors are excluded from the CHANNEL CAllBRATION." will be added to the table notation for Table 4.3-1. This note will exclude the neutron detectors from the channel calibration requirement 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, " Reactor Protection j Instrumentation Surveillance Requirements," for Wide Range Logarithmic l

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, ...o...- m_.. - . _ < _ - .n.

U. S. Nucl::gr Regulatory Commission B17492/ Attachment 1/Page 7-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 perform a channel check and channel function test, and will exclude the neutron detectors from the channel calibration requirement, as previously discussed. l 1

3. The channel calibration requirement in Table 4.36, " Remote Shutdown Instrumentation Surveillance Requirements," for Wide Range Logarithmic l Neutron Flux, will be changed from "N.A." to "R*." This change to add a requirement to perform a channel calibration is consistent with the current

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requirement to perform a channel check. The statement "* Neutron detectors i 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 requirement, as previously discussed. j Additional items The proposed additional changes will correct errors inadvertently introduced on Technical Specification Page 2-4 when replacement pages were submitted with license - i amendment requests. The proposed changes will also correct a reference to a surveillance requirement contained in an action statement.

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% >

power, provided the bypass is automatically removed at or above 5% power. f j 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 -

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Low, Functional Unit 3. Both of these 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.

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 March 2,1979.* A reference to Note 1 was also contained in a subsequent NNECO submittal to the NRC dated March 23,1979," ) which added the setpoint 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.""

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  • W. G. Counsil letter to U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power Station, Unit  !

No. 2, Proposed Revisions to Technical Specifications," dated March 2,1979.  ;

"* W. G. Counsil letter to U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power Station, Unit No. 2, Reactor Coolant Pump Speed Sensing System," dated March 23,1979.

M R. W. Reid letter to W. G. Counsil, License Amendment No. 52 to Facility Operating License DPR-65, dated May 12,1979.

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l U. S. Nuclur Regulatory Commission B17492/ Attachment 1/Page 8 l In a letter dated August 29,1980,0* NNECO proposed to revise the setpoint for the reactor coolant pump underspeed trip. As part of this cubmittal, NNECO 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 refersnce to Note 1 was not discussed in the NNECO submittal. It appears l'

that the reference to Note 1 was inadvertently removed when the wording was changed. The reference to Note 1 for Reactor Ccolant Pump Speed - Low, l Functional Unit 4, Page 2-4, was not contained in License Amendment No. 61 to Facility Operating License DPR45.08 l

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 I Specifications or the Safety Evaluation Report for License Amendment No. 61.

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

! 61.

2. The trip setpoint for Power Level - High, Four Reactor Coolant Pumps Operating, Functional Unit 2, in Table 2.2-1, " Reactor Protective Instrumentation

( Trip Setpoint Limits," will be modified by adding ", and a maximum of < 106.6%

l of RATED THERMAL POWER."

The proposed change will correct an error on Technical Specification Page 2-4.

The maximum power level - high trip setpoint, four reactor coolant pumps operating, was inadvertently omitted by NNECO in the license amendment 0

requests dated November 15,1988

  • and February 1,1989"*. These license amendment requests included an incorrect Technical Specification Page 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 UD W. G. Counsil letter to U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power Station, Unit l No. 2, Proposed Revisions to Technical Specifications," dated August 29,1980.

US

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

' DPR-65, dated October 6,1980.

US 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. 68360)," dated November 15,1988.

UM E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power Station, Unit l No. 2, Proposed Revision to Technical Specifications, Reduced Reactor Coolant System Flow Rate,"

i dated February 1,1989.

US G. S. Vissing letter to E. J. Mroczka, " Issuance of Amendment (TAC NO. 68360)," dated March 20,

1989.

b

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l U. S. Nuclear Regulttory Commission B17492/ Attachment 1/Page 9 in License Amendment No. 61, to Facility Operating License DPR45,"U 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.

'icense 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 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 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."* 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 .

Attachment 5 contains the history for Technical Specification Page 2-4 from License Amendment No. 61 to License Amendment No.199.

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

US 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, *lssuance of Amendment (TAC NO. M94466)," dated July 2,1996.

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U. S. Nuclear Regulitory Commission l B17492/ Attachment 1/Page 10 This issue was discovered during a review of the Technical Specifications for the Millstone Unit No. 2 Improved Technical Specifications project. It was documented by Adverse Condition Report (ACR) M2-96-0210. A reportability -

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 5 106.6%. '

3. Technical Specification Page 3/4 3-9 was previousi revised by License Amendment No. 9" and License Amendment No.15.g" These amendment numbers will be added to the bottom of the page, e
4. The surveillance requirement (Specification 4.3.2.1) referenced in Action 4.b.1 of .

Table 3.3-3, " Engineered Safety Feature Actuation System Instrumentation,"

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

i Safety Summarv 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.

i 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. l

5. Exclude neutron detectors from the channel calibration requirement.

l 6. Add license amendment numbers to a page, correct a reference to a surveillance requirement and errors on Technical Specification Page 2-4, Table 2.2-1, i " Reactor Protective instrumentation Trip Setpoint Limits."

l 6 ,

! "* Letter from the NRC, Millstone Nuclear Power Station, Unit No. 2 License Amendment No. 9, dated l March 31,1976.

i US Letter from the NRC, Millstone Nuclear Power Station, Unit No. 2 License Amendment No.15, dated September 2,1976.

g$- ,--T y-- - - - - r q .g e , y - --es m,e -+,m- c,

U. S. Nuclecr Regulatory Commission B17492/ Attachment 1/Page 11 j 7. Expand the Bases of the affected Technical Specifications to explain how the Reactor Protection System (RPS) or Engineered Safety Feature Actuation System (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 ,

operability or surveillance requirements for any reactor protection or engineered safety l 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 Enclosure 3 of the NRC corresponh dated Aptil 16, 1981,g2 and with I NUREG-1432. ,

. 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 j condition. This change will continue to allow a second channel to be removed from l 3 service. However, the length of time this configuration is allowed will be increased from j 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 />. Tha 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 j 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.) j 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 requirements 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 detectora 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 r22)

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

U. S. Nuciser Regulatory Commission B17492/Attachm:nt 1/Page 12

, calorimetric calibration and the monthly calibration using the incore detectors.

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

Adding the license amendment numbers to Technical Specification Page 3/4 3-9 will not result in a technical change to the Millstone Unit No. 2 Technical Specifications.

Therefore, the RPS will continue to function as before.

Changing the surve!Ilance requirement (Specification 4.3.2.1) referenced in Action 4.b.1 of Table 3.3-3, " Engineered Safety Feature Actuation System instrumentation," 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 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 effect on how the plant is operated. Adding a reference to the reactor coolant pump low speed reactor trip function to a note that states this trip may be bypassed < 5% power, and that the bypass must be automatically removed > 5% will not affect this reactor 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 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, tLe RPS will continue to function as before.

Expanding the Bases of the affected 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 adverse 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 B17492 i

I Attachment 2 1

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

November 1998 l

I

l U.S. Nucl=r Regulttory Commission -

l B17492\ Attachment 2\Page 1 Proposed Revision to Technical Specifications j _ Reactor Protective and Engineered Safety l Feature Actuation System Instrumentation l Significant Hazards Consideration Sianificant Hazards Consideration in accordance with 10CFR50.92, NNECb 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 would not:

1

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

previously evaluated.

l 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 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 the design of the Reactor Protection System (RPS) or the Engineered Safety Feature Actuation l System (ESFAS) and does not affect how these systems operate. In addition, l' this will minimize the susceptibility of these systems to the remote possibility of fault propagation between channels. However, this proposed change will require an inoperable pressurizer high pressure reactor protection 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 />. With a pressurizer pressure channel in the tripped condition, the high failure of a second pressurizer pressure channel would initiate a reactor trip and open both pressurizer power operated relief valves (PORVs). Opening the pressurizer PORVs would result in an undesired loss of l primary coolant. Thus, this change will increase the probability of occurrence of a previously evaluated accident. Howerar, this would not place the plant in an

)

unanalyzed condition since FSAR Section 14.6.1 analyzes the inadvertent '

l opening of both PORVs, the release of reactor coolant can be terminated by closure of the PORV block valves frum the control room, and the Emergency Operating Procedures provide guidance on how to address this situation.

Therefore, this change does not significantly increase the probability or consequences of an accident previously evaluated. I 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 cystems operate.

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

i However, this change will also impact the probability of occurrence of a j

previously evaluated accident since it will allow a second pressurizer high

U.S. Nuclatr Regulttory Commission B17492\ Attachment 2\Page 2 pressure reactor protection channel to be placed in the tripped 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. The impact of this change is bounded by the proposed change to require an inoperable pressurizer high prvssure reactor protection channel 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 /> as previously discussed. Therefore, this change does 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 consequences of an accident previously evaluated.

The proposed change to add the license amendment numbers to Technical Specification Page 3/4 3-9 will not result in a technical change to the Millstone Unit No. 2 Technical Specifications. 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 proposed change to correct the surveillance requirement referenced in an 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.

i 1 - y- i+----- - & *e e

' U.S. Nucle:r Regulatory Commission l B17492\Attechmant 2\Pege 3 i

The proposed change to add a reference to the reactor coolant pump low speed j reactor trip function to a note that states this trip may be bypassed < 5% power, and that the bypass must be automatically removed t 5% will not affect this ,

l- reactor trip function. This bypass capability currently exists in the design of the '

L 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 addition of this note will not '

l result in any technical change to the Millstone Unit No. 2 RPS. The RPS will  ;

l continue to function as before. Therefore, this change does not significantly l

) increase the probability or consequences of an accident previously evaluated.

The proposed change to correct the power level high trip setpoint on Technical ,

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 listed l on Page 2-4 will agree with the setpoint previously approved by the NRC, and i currently used by the RPS. The change has no effect on the design of the RPS I 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 i significantly increase the probability or consequences of an accident previously j l

evaluated.

l The information added to the Bases of the affected Technical Specifications to l 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. The RPS and ESFAS will continue to function as desigr'ed to mitigate design basis accidents. Therefore, this change does not significantly increase the probability or consequences of an accident previously 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 y consequences of an accident previously evaluated.

1 l

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

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

assumptions made in the safety analysis and licensing basis. The RPS and the
ESFAS will still function as designed to mitigate design basis accidents.

4

U.S. Nuclarr Regulatory Commission B17492\Attechm:nt 2\ Pegs 4 Therefore, these changes do not create the p ssibility 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 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, most of the changes are consistent with the Calvert Cliffs RPS and ESFAS E Technical Specifications model provided in Enclosure 3 of the NRC co respondence dated April 16, 1981* and with the new, improved Standard Technical Specifications (STS) for Combustion Engineering plants (NUREG-1432).

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

The NRC has provided guidance concerning the application of standards in 10CFR50.92 by providing certain examples (March 6, 1986, 51 FR 7751) of amendments that are considered not likely to involve an SHC. The changes proposed herein to add the license amendment numbers to a Technical Specification page, to correct the error on Technical Specification Page 2-4 (Power Level - High and Reactor Coolant Pump Speed - Low), and to correct the surveillance requirement referenced in an action statement are enveloped by example (i), a purely administrative change to 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 lose of turbine load reactor trip function, and to require a channel calibratien 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 increase 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, 2

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

U.S. NuclIr Regulatory Commission B17492\ Attachment 2\Page 5 and does not result in a significant reduction in a margin of safety. Therefore, NNECO has corscluded that the proposed changes do not involve an SHC.

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