B16595, Application for Amend to License DPR-65,incorporating Various Compliance Issues Identified in TS

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Application for Amend to License DPR-65,incorporating Various Compliance Issues Identified in TS
ML20216C219
Person / Time
Site: Millstone Dominion icon.png
Issue date: 09/02/1997
From: Bowling M
NORTHEAST NUCLEAR ENERGY CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20216C223 List:
References
B16595, NUDOCS 9709080247
Download: ML20216C219 (16)
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Text

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%c Nottinut Utihist. $ 3.ieni SEP 2 1997 Docket No. 50-336 B16595 Re: 10CFR50.90 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 l

Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Compliance issues I

Introduction l Pursuant to 10CFR50.90, 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 will correct various compliance issues identified with Technical Specifications.

The proposed changes affect Technical Specification Sections Index 3/4.6.2, Definition 1.8, 4.1.1.3, 4.4.6.2, 4.5.2, 4.6.1.1, 3.6.2.1, 4.6.2.1, 3.6.2.2, 4.6.2.2, 5.5.1, B 3/4.5.2 and 3/4.5.3, B 3/4.6.1.1, B 3/4.6.2.1, and B 3/4.6.2.2.

The proposed change to Technical Specification Index 3/4.6.2 is on the same page (Vil) as Index 3/4.6.5, which has been proposed to be changed in a separate letter dated April 10,1997'. This previous submittal addressed Enclosure Building Integrity.

Technical Specification Definition 1.8, Section 4.6.1.1, and B 3/4.6.1.1 were alsc \\

proposed to be changed in a separate letter dated May 20,1997, which addressed containment isolation valves. Technical Specification Section B 3/4.5.2 and 3/4.5.3 was M. L. Bowling letter to the NRC, " Millstone Nuclear Power Station, Unit No. 2 Proposed /

Revision to Technical Specifications Enclosure Building," dated April 10,1997. gQ M. L. Bowling letter to the NRC, " Millstone Nuclear Power Station, Unit No. 2 Proposed \J Revision to Technical Specifications Containment Isolation Valves (TAC No. M94623),"

dated May 20,1997.

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1 I U. S. Nuclear Regul: tory Commission B16595/Page 2 also proposed to be changed in a separate [[letter::B15335, Application for Amend to License DPR-65,extending AOT from 48 H to 7 Days for ECCS Train Declared Inoperable as Result of Inoperable LPSI Subsystem|letter dated November 3,1995]],8 which addressed the allowed outage time of an Emergency Core Cooling subsystem. The proposed changes contained in this letter do not assume approval of any of the previously submitted changes.

Attachment 1 provides a discussion of the proposed changes and the Safety Assessment. 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 Technical J

l Specifications.

Environmental Considerations l

NNECO has reviewed the proposed license amendment request against ', ,e criteria of 10CFR51.22 for environmental considerations. The proposed changes modify several Technical Specifications. These changes do not significantly increase the type and amounts of effluents that may be released offsite, in addition, this License 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 because of the proposed changes to the allowed outage times. However, 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 Assessment 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.

3 E. A. DeBarba letter to the NRC,

  • Millstone Nuclear Power Station, Unit No. 2 Proposed Technical Specification Revision Emergency Core Cooling Subsystem Allowed Outage Time Extension," dated November 3,1995.

1 i U. S. Nucirr Regu:: tory Commir, tion B16595/Page 3 Schedule We request issuance at your earliest convenience, with the amendment to be imp!emonted within 30 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.

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

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY h

Martin L. Bowling, Jr.

w -11 Millstone Unit No.very 2 Officer

- Rec (o/

Sworn to and subscribed before me this A/ dayof $2ce/wM/e .1997 mk IN$-

Notary Pubic /

My Commission expires

/a d//d /

Attachments (5) cc: H. J. Miller , Region I Administrator D. G. Mcdonald, Jr., NRC Project Manager, Millstone Unit No. 2 D. P. Beaulieu, Senior Resident inspector, Millstone Unit No. 2 W. D. Travers, PhD, Director, Special Projects W. D. Lanning, Director, Millstone Assessment Team Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection

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Docket No. 50-336 B16595 l

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Attachment 1 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Compliance issues Discussion of Proposed Changes September 1997 o

i r U. S. Nuclear Regul: tory Commission B16595/ Attachment 1/Page 1 Proposed Revision to Technical Specifications Compliance issues Discussion of Proposed Changes jnLrgduction Northeast Nuclear Energy Company (NNECO) is proposing to change Technical Specification Definition 1.8, Technical Specification 3.1.1.3, Technical Specification 3.4.6.2, Technical Specification 3.5.2, Technical Specification 3.6.1.1, Technical Specification 3.6.2.1, Technical Specification 3 6.2.2, and Technical Specification 5.5.1. If a comparable Technical Specification exists, the proposed changes are consistent with the new, improved Standard Technical Specifications (STS) for Combustion Engineering plants (NUREG-1432).

Each proposed change is discussed in the next section. 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.

l Descrintion of Proposed Chances l

l The proposed changes are described below.

1. The Containment Spray (CS) System and the Containment Air Recirculation (CAR) and Cooling System provide sufficient heat removal capability to limit containment pressure and temperature to within design limits following a design basis loss of coolant accident (LOCA) or a main steam line break (MSLB). The CS System consists of two trains and the CAR and Cooling System consists of four CAR coolers. Two trains of CS and four CAR coolers provide approximately 200% of the required heat removal capacity Loss of one CS train or one or two CAR coolers does not significantly degrade overall system capacity. The loss of one CS train and two CAR coolers, or four CAR coolers is equivalent to the loss of an entire train. The remaining equipment can provide sufficient heat removal capacity. (The loss of two CS trains is not acceptable because of the need for CS to mitigate a main steam break inside containment. CS is more effective at removing the energy contained in superheated steam.)

The current Technical Specifications 3.6.2.1, ' Containment Spray System,' and 3.6.2.2, " Containment Air Recirculation System' do not address the loss of one CS train and two CAR coolers. This can occur if the associated Reactor Building Closed Cooling Water (RBCCW) or Service Water (SW) train becomes inoperable. Since the current Technical Specifications do not address this situation, entry into LCO 3.0.3 would be required. This has occurred at Millstone I

i  :

U. S. Nuclear Regul: tory Commission B16595/ Attachment 1/Page 2 Unit No. 2 as reported in Licensee Event Report (LER) 97-022-00'. Therefore, the current Technical Specifications need to be modified to address this situation.

t Index Page Vil will be modified to combine Technical Specifications 3.6.2.1,

' Containment Spray System,' and 3.6.2.2, " Containment Air Recirculation System" into one Technical Specification 3.6.2.1, ' Containment Spray and Cooling Systems."

Technical Specifications 3.6.2.1, ' Containment Spray System," and 3.6.2.2,

' Containment Air Recirculation System' will be combined into one Technical Specification 3.6.2.1, ' Containment Spray and Cooling Systems." All requirements of the original Technical Specifications will be retained. However, the new Technical Specification will refer to containment cooling (CC) trains instead of individual CAR cooling units. Each CC train will consist of two CAR cooling units.

The proposed LCO 3.6.2.1, ' Containment Spray and Cooling Systems," will require two CS trains and two CC trains (instead of four CAR cooling units, as i discussed above) to be operable. The additional detail in the original LCO 3.6.2.1 will be relocated to the associated Bases. There is no technical difference between the original LCOs and the proposed LCO.

The proposed applicability remains the same. However, the footnote referenced by the "" has been reworded. The proposed wording will not change any technical aspect of the footnote.

( The proposed changes to the original Technical Specification Action Statements (TSASs) are summarized in the following table.

The proposed changes to the TSAS account for the approximate 200% capacity when two CS trains and two CC trains are operable. Loss of one CS or CC train does not significantly degrade overall system capacity. An allowed outage time of 7 days is acceptable. The loss of one CS train and one CC train (two CAR cooling units), or two CC trains (four CAR cooling units) is equivalent to the loss of an entire tr.ah. (The loss of two CS trains is not acceptable because of the need for CS to mitigate a main steam break inside containment.) An allowed outage time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is acceptable and consistent with the allowed outage time for the loss of an emergency core cooling train, service water train, or reactor building closed cooling water train. The loss of any combinations not specifically addressed, would have required entry into LCO 3.0.3. The proposed change provides specific directions to enter LCO 3.0.3.

J. A. Price letter to the U.S. Nuclear Regulatory Commission, Millstone Nuclear Power Station, Unit No. 2 Licensee Event Report (LER) 97 022 00, ' Technical Specification Violations,' dated July 09,1997.

i i U. S. Nuclear Regulatory Commission B16595/ Attachment 1/Page 3 inoperable Current TSAS Proposed improved TSAS Equipment TSAS NUREG 1432 (LCO 3.6.6B) 1 CS train 30 days 7 days 7 days TSAS 3.6.2.1.a 1 CAR 30 days 7 days - would 7 days - would inop TSAS 3.6.2.2.s inop a CC train a CC train 2 CARS 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 7 days 7 days TSAS 3.6.2.2.c 1 CAR and 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> - 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> - would 1CS TSAS 3.6.2.1.b would inop a inop a CC train

& 3.6.2.2.b CC train 2 CARS and 1 Not Covered 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CS (TS 3.0.3 shutdown) 3 or 4 CARS Not Covered 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> l (TS 3.0.3

! shutdown) 2 CS trains Not Covered Plant shutdown 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (TS 3.0.3 required - see shutdown) next item All other Not Covered Enter LCO Enter LCO 3.0.3 combinations (TS 3.0.3 3.0.3 immediately shutdown) immediately All surveillance requirements have been retained. However, Surveillance Requirements 4.6.2.1.a-d have been renumbered as 4.6.2.1.1.a-d and Surveillance Requirements 4.6.2.2.a-c have been renumbered as 4.6.2.1.2.a-c.

Surveillance Requirement 4.5.2.c.5 will be revised by changing "4.6.2.1.c" to

  • 4.6.2.1.1,c."

Minor wording changes have been made to replace " containment spray system" with ' containment spray train" throughout the proposed Technical Specification, The containment spray system is composed of two trains. The original Technical Specification 3.6.2.1 incorrectly used the word ' system" when referring to trains.

. i U. S. Nuclear Regul: tory Commission B16595/ Attachment 1/Page 4

2. An "" will be added to the definition of containment integrity and an associated footnote will be added to the bottom of the affected page. The footnote will l

explain how the requirement for an operable containment automatic isolation  ;

valve system is satisfied in Mode 4. When the plant is in Mode 4, the automatic containment isolation signals generated by low pressurizer pressure and high containment pressure are not required to be operable. Also, the low pressurizer pressure signal is blocked during the plant cooldown. This is acceptable because in Mode 4 automatic actuation of Engineered Safety Features Actuation System (ESFAS) functions is not required since adequate time is available for plant operators to evaluate plant conditions and respond by manually operating the ESF equipment. Because of the large number of valves actuated by a containment isolation actuation signal (CIAS), actuation is simplified by the use of the manual pushbuttons. Since the manual CIAS pushbuttons are requited to be operable in Mode 4, credit can be taken for remote manual operation to close the containment isolation valves. This change is consistent with NUREG-1432 (Analog Bases 3.3.4).

An "" will be added to Surveillance Requirement 4.6.1.1.a and an associated footnote will be added to the bottom of the affected page. This is the same footnote explained above.

3. An "" will be added to Surveillance Requirement 4.1.1.3 and an associated footnote will be added to the bottom of the affected page. The footnote will exempt performance of Surveillance Requirement 4.1.1.3 when the plant is in Modes 1 or 2 because at laast one reactor coolant pump will always be in operation (Technical Specification 3.4.1.1, ' Coolant Loops and Coolant Circulation Startup and power Operation") and the required dilution flow of 1000 gpm will always be met. This exception will not apply if operating in accordance with Special Test Exception 3.10.4, " Physics Tests," which relaxes the requirements of Technical Specification 3.4.1.1.
4. Surveillance Requirements 4.4.6.2.a and 4.4.6.2.b of Technical Specification 3.4.6.2, " Reactor Coolant System Leakage," will be deleted. Neither of these surveillances verify compliance with the leakage limits specified in LCO 3.4.6.2.

This LCO has specific leakage values that cannot be verified by monitoring the containment sump level or the containment atmosphere particulate radioactivity.

These instruments provide indication of leakage, but cannot provide a value of leakage with the required accuracy to ensure compliance with the LCO. The operability and surveillance requirements for these instruments is addressed by Technical Specification 3.4.6.1, " Leakage Detection Systems." Surveillance Requirement 4.4.6.2.c will become 4.4.6.2 as a result of this change. This change is consistent with NUREG 1432 (LCOs 3.4.13 and 3.4.15).

. i U. S. Nucl:ar Regul: tory Commission B16S95/ Attachment 1/Page 5

5. 1 Surveillance Requirement 4.5.2.e will be revised.

The current wording, *will open to the correct position,' implies it is necessary to verify the valve actually opens to the correct position. An acceptable method should be to manually openfurther.

open the valve to the required position, and then verify that the valve will not method. However, the current wording does not allow this additional The proposed change will allow the use of alternate methods, as specified in the Bases, while ensuring the intent of the surveillance requirement is met. This change is consistent with NUREG-0212 and NUREG-1432.

6.

Technical Specification 5.5.1 will be modified by removing the word " original."

The original design provisions contained in FSAR Section 6.3 have been changed. Therefore, it is not correct to refer to the original design provisions.

Safety Assessment The proposed changes will:

1.

Combine Technical < Specifications 3.6.2.1 and 3.6.2.2 into one specification; reduce the allowed outage time for one inoperable CS train or one inoperable CAR cooler from 30 days to 7 days; increase the allowed outage time for two inoperable CAR coolers from 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to 7 days; add an allowed outage time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (instead of entering Technical Specification 3.0.3) for one inoperable CS train and two inoperable CAR coolers, or three or four inoperable CAR coolers; provide specific guidance when it is necessary to enter Technical Specification changes. 3.0.3; and expand the associated Bases to discuss these propose I

2.

Modify the definition of contairm ent integrity, modify Technical Specification 3.6.1.1, " Containment Integrity

  • c ad expand the associated Bases to explain why automatic containment isolation valves are operable in Mode 4.

3.

Provide an exception to Surveillance P*quirement 4.1.1.3 when the plant is in Modes 1 and 2.

4.

Delete Survelliance Requirements 4.4.6.2.a and 4.4.6.2.b.

5.

Modify Surveillance Requirement 4.5.2.e to allow the use of alternate methods and expand the associated Bases to discuss these alternate methods.

6.

Modify Technical Specification 5.5.1 by removing the word " original."

7.

Make editorial changes to terminology and item numbering.

Combining Technical Specifications 3.6.2.1, ' Containment Spray System," and 3.6

" Containment Air Recirculation System," into one specification does not reduce the

. i U. S. Nucirr R:gul: tory Commission B16595/ Attachment 1/Page 6 operability or surveillance requiruments for either of these systems because all operability and surveillance requirements have been retained. Reducing the allowed outage time for one inoperable CS train or one inoperable CAR cooler from 30 days to 7 days will enhance overall system reliability by placing a higher priority on restoring the inoperable equipment. Increasing the allowed outage time for two inoperable CAR l coolers from 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to 7 days will not adversely affect overall system reliability because a 7 day allowable outage time still places a high priority on restoring the l inoperable equipment, in addition,7 days better reflects the amount of containment cooling, approximately 25%, that is inoperable. (25% is approximately equivalent to one CG train. Currently Technical Specification 3.6.2.1 allows 30 days to restore a CS train. This proposed change will reduce this to 7 days.) Adding an allowed outage time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for one inoperable CS train and two inoperable CAR coolers, or three or four inoperable CAR coolers reflects the amount of containment cooling, approximately 50%, that is inoperable. This is equivalent to the loss of one entire train, and should have an equivalent allowed outage time. (Technical Specification 3.7.3.1,

' Reactor Building Closed Cooling Water System,' and 3.7.4.1, ' Service Water System,"

allow 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to restore an inoperable train.) Also, this configuration should not require entering Technical Specification 3.0.3. Finally, providing specific guidance when it is necessary to enter Technical Specification 3.0.3 does not change any current i requirement. This just states the required action, instead of relying on interpretation by l

licensed operators. These proposed changes do not reduce any operability requirements cr change any surveillance requirements. The proposed changes do modify allowed outage times. However, the changes will improve system reliability by providing allowed outage times that more accurately reflect the amount of system

+

capability that is inoperable, and the changes will be consistent and based on the inoperable system capability. Therefore, the CS and CAR Systems will continue to function as designed to mitigate design basis accidents, in addition, these proposed changes are consistent with NUREG-1432.

Modifying Technical Specifications to stata the requirement for an operable containment automatic isolation valve system is met by use of the manual containment isolation pushbuttons (in Mode 4) is acceptable because in Mode 4 automatic actuation of Engineered Safety Features Actuation System (ESFAS) functions is not required.

Adequate time is available for plant operators to evaluate plant conditions and respond by manually operating the ESF equipment. Because of the large number of valves actuated by a containment isolation actuation signal (CIAS), actuation is simplified by the use of the manual pushbuttons. Since the manual CIAS pushbuttons are required to be operable in Mode 4, credit can be taken for remote manual operation to close the

, containment isolation valves. This change does not reduce operability or surveillance requirements for any containment isolation valve or Engineered Safety Feature Actuation System (ESFAS) component. Therefore, the containment isolation valves and ESFAS components will continue to fuaction as designed to mitigate design basis accidents. This proposed change is consistent with the new, improved Standard Technical Specifications (STS) for Combustion Engineering plants (NUREG-1432).

. i U. S. Nuclear Regul: tory Commission B16595/ Attachment 1/Page 7 Millstone Unit No. 2 la currently required to verify sufficient flow through the core during a reduction in Reactor Coolant System (RCS) boron concentration during all modes of operation. The proposed change will provide an exception to Surveillance Requirement 4.1.1.3 when the plant is in Mode 1 or 2. It is not necessary to perform I Surveillance Requirement 4.1.1.3 because Millstone Unit No. 2 is required to have all four reactor coolant pumps (RCPs) in operation whenever the plant is in Mode 1 or 2 (Technical Specification 3.4.1.1). During normal Mode 1 or 2 operation, the loss of any RCP will result in the initiation of a reactor trip by the Reactor Protection System (RPS),

which will place the plant in Mode 3. It is not necessary to verify sufficient core flow during a reduction in RCS boron concentration in Mode 1 or 2 because all RCPs will be in operation. This change will not result in any new approach to plant operation, it simply removes the requirement to perform an unnecessary surveillance. This oxception is not applicable if the plant is operating in accordance with Technical Specification 3.10.4, "Special Test Exception - Physics Test."

Surveillance Requirements (SRs) 4.4.6.2.a and 4.4.6.2.b do not verify compliance with the leakage limits contained in Technical Specification 3.4.6.2, " Reactor Coolant System Leakage." The equipment covered by these 2 SRs, containment atmosphere particulate radioactivity monitors and containment sump inventory monitor, provide i

early indication that RCS leakage exists, but do not provide the specific information (amount of leakage) necessary to verify operation within the leakage limits.

l Performance of an RCS water inventory balance (currently SR 4.4.6.2.c) will be used to verify compilance with the leakage limits. Operability of the containment atmosphere particulate radioactivity monitors and containment sump inventory monitor is verified by SRs 4.4.6.1.a and 4.4.6.1.b. Therefore, this change does not reduce the operability requirements for any equipment used to monitor for RCS leakage. This change is l consistent with NUREG-1432.

Modifying Surveillance Requirement 4.5.2.e to allow the use of alternate methods and expanding the associated Bases to discuss these alternate methods does not reduce operability or surveillance requirements for any of the Emergency Core Cooling System (ECCS) throttle valves. Therefore, these ECCS throttle valves will continue to function as designed to mitigate design basis accidents.

Modifying Technical Specification 5.5.1 by removing the word " original' has no affect on how the plant is operated. This change will still require the ECCSs to be designed and maintained in accordance with the FSAR, however reference to original design is not appropriate since the' systems can be changed by using approved processes.

Therefore, the ECCSs will continue to function as designed to mitigate design basis accidents.

Expanding the Bases of the affected Technical Specifications to discuss the proposed changes, and making editorial changes to terminology and item numbering will have no affect on equipment operation. Therefore, all associated equipment will continue to function as before.

e i U. S. Nuclear Regulatory Commission B16595/ Attachment 1/Page 8 The proposed changes have no significant affect on how any of the associated systems or components function to mitigate the consequences of design basis accidents. Also, the proposed changes have no significant affect on any design basis accident previously evaluated. Therefore, there is no significant impact on the public health and safety.

I

Docket No. 50-336 B16595 l

l Attachment 2 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Complianca lasues Sicnificant Hazards Consideration September 1997

.. o U. S. Nuclear Regulatory Commission B16595/ Attachment 2/Page 1 Proposed Revision to Technical Specifications Compliance issues Significant Hazards Consideration Sianificant 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 would not:

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

The proposed change to combine Technical Specifications 3.6.2.1 and 3.6.2.2 into one specification reduces the allowed outage time for one inoperable containment spray (CS) train or one inoperable containment air recirculation (CAR) cooler from 30 days to 7 days; increases the allowed outage tims for two inoperable CAR coolers from 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to 7 days; adds an allowed outage time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (instead of entering Technical Specification 3.0.3) for one inoperable CS train and two inoperable CAR coolers, or three or four inoperable CAR coolors; and provides specific guidance when it is necessary to enter Technical .

Specification 3.0.3 will not affect how these systems function to mitigate design basis accidents. Therefore, this change does not significantly increase the probability or consequences of an accident previously evaluated. l The proposed changes to modify the definition of containment integrity, modify the Technical Specification 3.6.1.1, ' Containment Integrity," and expand the Bases to explain why automatic containment isolation valves are operable in Mode 4 have no affect on any containment isolation valve or Engineered Safety Feature Actuation System (ESFAS) component. These components will still i f Tetion as designed to mitigate design basis accidents. Therefore, this change i not significantly increase the probability or consequences of an axident t jously evaluated.

The proposed change to provide an exception to Surveillance Requirement 4.1.1.3 when the plant is in Modes 1 and 2 will not result in any new approach to plant operation, it simply removes the requirement to perform an unnecessary surveillance. The minimum coolant flow through the core during a reduction in Reactor Coolant System (RCS) boron concentration will still be met. Therefore, this change does not significantly increase the probability or consequences of an accident previously evaluated.

U. S. Nuclear Regul tory Commission B16505/ Attachment 2/Page 2 The proposed change to delete Surveillance Requirements (SRs) 4.4.6.2.a and 4.4.6.2.b does not reduce the operability requirements for any equipment used to monitor RCS leakage. The equipment covered by these 2 SRs, containment atmosphere particulate radioactivity monitors and containment sump invento.y  :

monitor, provide early indication that RCS loabge exists, but do not provide the  !

specific information (amount of leakage) necessary to verify operation within the 4 leakage limits contained !n Technical Specification 3.4.6.2, " Reactor Coolant System Leakage."

Opvability .of the containment atmosphere particule's radioactivity monitors and containment sump inventory monitor is verified by SRs 4.4.6.1.s and 4.4.6.1.b. Therefore, this change does not significantly_ increase the probability or consequences of an accident previously evaluated. .i 1

The proposed change to Surveillance Requirement 4.5.2.s to allow the use of alternate methods does not reduce operability or surveillance requirements for d any of the Emergency Core Cooling System (ECCS) throttle valves. Therefore, d these ECCS throttle valves 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 evaluated.

The picpcsed change to Technical Specification 5.5.1 has no affect on how the ECCS operates. The ECCS 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 changes to add information to the Bases of the affected Technical Specifications, and make editorial changes to terminology and item numbering will have no affect on equipment operation. Therefore, all associated equipment 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 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 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 assumptions made in the safety analysis and licensing basis. The affected

U. S. Nuclear Reguttory Commission B16595/ Attachment 2/Page 3 components and systems will still function 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 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 requirement to check containment radiation and centainment sump level every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> has been eliminated.

However, this equipment is still required to be operable, and the surveillance requirements to verify operability have not been changed. Therefore, this equipment will be available to provide early indication of RCS leakage.

The effectiveness of Technical Specifications will be maintained since the i

changes will not alter the operation of any component or system. In addition,

! the changes are consistent 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 correct terminology and numbering are enveloped by example (i), a purely administrative change to Technical Specifications. All other changes proposed herein are not enveloped by a specific example.

As described above, this License Amendment Request does not 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 l concluded that the proposed changes do not involve an SHC. I 1

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