ML20211P416

Proposed Tech Specs Replacing Pressurizer Max Water Inventory Requirement W/Pressurizer Max Indicated Level Requirement

ML20211P416
Person / Time
Site:	Millstone
Issue date:	10/15/1997
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20211P405	List: B16625, Application for Amend to License NPF-49,replacing Pressurizer Max Water Inventory Requirement W/Pressurizer Max Indicated Level Requirement ML20211P416
References
NUDOCS 9710200189
Download: ML20211P416 (18)

Text

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j 1-Docket No. 50-423 B16625 Millstone Nuclear Power Station Unit No. 3 Proposed Revision to Technical Specification Pressurizer Level MCR 3-31-97)

Marked Uo Paoes October 1997 9710200189 971015 DR ADOCK 0500 3

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i U.S. Nucl:ar Regulttory Commission B16625%ttachment 2\\Page1 i

MARKUP OF PROPOSED REVISION 7

Refer to the attached markup of the proposed revision to the Technical Specifications.

The - attached markup reflects the currently issued version of the Technical s

l Specifications listed below. Pending Technical Specification revisions or Technical Specification revisions issued subsequent to this submittal are not reflected in the enclosed markup.

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The following Technical Specification changes are included in the attached markup.

i The Technical Specification Limiting Condition for Operation (LCO) is reformatted e

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. based on the NUREG 1431 Rev i standard.

Limit (a) establishes Pressurizer Water level of 89% of instrument span as the new limil This value and wording is consistent with the numerical value for the Reactor Trip setpoint for pressurizer high water level in existing TS table 2.2-1. Limit (b) maintains the existing TS wording L

and is consistent with the unchanged portions of the LCO and Surveillance Requirements. Surveillance Requirement 4.4.3.1 is changed to reflect level instead 2

j of volume.

i 3/4.4.3 The wording is changed reflecting the new operability condition to ensure that a j

steam bubble exists.

Bases 3/4.4.3 i

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i REACTOR C00UWF SYSTEM 3/4.4.3 PRESSURI7ER k

LINITING CONDITION FOR OPERATION 4

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3.4.3 The ressurizer shall OPERA 5i.5 with a Ter volume of ns than or equal to li.

by em(1656 cubic fe, and at 'seast groups of pr urizer heaters ergency p each having a e acity of at le t 175 kW.

APPLICABILITY: MODES I, 2, and 3.

ACIl21:

With only one group of pressurizer heaters supplied by emergency power a.

OPERABLE, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the ne.it 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in '

HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, b.

With the pressurizer otherwise inoperable, be in at least I!OT STANDBY with the Reactor Trip System breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the fo11owing.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS Idel y

-a, 4.4.3.1 The pressurizer water =L shall be dete'rmined to be wiEhin its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.3.2 The capacity of each of the above required groups of pressurizer heaters supplied by emergency power shall be verified by energizing the heaters and ineasuring ci.rcuit current at least once each refueling interval.

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MILLSTONE - l#11T 3 3/4 4-11 Amendment No. 100 4347 7

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' Insert 3.4.3 :

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- 3.4.3 ' - The pressurizer shall be OPERABLE with:

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- Pressurizer water level 5 89% of instrument span; and r

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j-At least two groups of pressurizer heaters supplied by emergency power, each having a capacity of at least 175 kW, i

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e REACTOR COOLANT SYSTEM I(

JAN 81 1986 SASES 3/4.4.2 SAFETY VALVES p'ressurized above its safety Limit of 2750 psia.The pressurizer Code safety Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve Setpoint.

condition which could occur during shutdown. relief capacity of a single safety v The In the event that no safety valves are OPERABLE, an operating RNR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

Cold Overpressure Protection System provides a diverse means of protectionIn addition, the against RCS overpressurization at low temperatures.

During operation, all pressurizer Code safety valves sust be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2750 psia.

combined relief capacity of all of these valves is greater than the maximum The surge rate resulting from a complete loss of-load. assuming no Reactor trip until the first Reactor Trip System Trip Setpoint is reached (i.e., no credit is taken for a direct Reactor trip on the loss o'-load) and also assuming no cperation of the power operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASMe Boller and Pressure Code.

3/4. 4. 3 PRESSURIZER

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"00 h r~t e hc,dralk:11y = lid :yttr$ The requirement that a minimum

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number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant Systes pressure and establish natural circulation.

3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.

Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Code safety valves.

Each PORV'has a remotely operated block valve to provide a positive snutoff ecpability should a relief valve become inoperable.

Requiring the PORVs to be OPERABLE ensures that the capability for depressurization during safety grade cold shutdown is set.

MILLSTONE - UNIT 3 8 3/4 4-2 o (ILp

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Insert B.3.4.3 4

The requirement for the pressurizer to be OPERABLE, with a level less than or equal to 89%, ensures that a steam bubble exists. The 89% level preserves the steam space for pressure control. The 89% level has been established to ensure the capabi ity to establish and maintain pressure control for steady state i

operation and to minimize the consequences of potential overpressure transients. He 89% pressurizer level does not establish an initial condition for an accident or transient analysis. Automatic and procedural controls adequately maintain pressurizer level to assure the validity of the FSAR Chapter 15 accident analyses. -

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. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance requires that during steady state or: ration, pressurizer level is maintained below the nominal upper limit to provide a minimum space for a steam bubble. He -

4 Su:veillance is performed by observing the indicated level. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval has been shown_ by operating practice to be sufficient to regularly assess level for any deviation and to ensure that a steam bubble exists in the pressurizer. Alarms are also available for early detection of abnormal level indications.

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x Docket No. 50-423 gi@251 Millstone Nuclear Power Station Unit No. 3 Proposed Revision to Technical Specification Pressurizer Level (PTSCR 3-31-97)

Retvoad Paae5 October 1997 o

i U.S. Nucl;;r R:gulatory Commission B16625%ttachment 3\\Page1 -

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' RETYPE OF PROPOSED REVISION i

Refer to the attached retype of the proposed revision to the Technical Specifications.

The attached retype reflects the currently issued version of the Technical Specifications. Pending Technical Specification revisions or Technical Specification revisions issued subsequent to this submittal are not reflected in the enclosed retype.

4 The enciosad retype should be checked for continuity with Technical Specifications prior to issuance.

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i REACTOR COOLANT SYSTEN 3/4.4.3 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.?

The pressurizer shall be OPERABLE with:

a.

Pressurizer water level 189% of instrument span; and b.

At least two groups of pressurizer heaters supplied by emergency power, each having a capacity of at least 175 kW.

APfLICABILITY: MODES 1, 2, and 3.

ACTION:

a.

With only one group of pressurizer heaters supplied by emergency power OPERABLE, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With the pressurizer otherwise inoperable, be in at least HOT STAN0BY with the Reactor Trip System breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT l

SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b SURVEILLANCE REQUIRENENTS 4.4.3.1 The pressurizer water level shall be determined to be within its l

limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.3.2 The capacity of each of the above required groups of pressurizer heaters supplied by emergency power shall be verified by energizing the heaters and measuring circuit current at least once each refueling interval.

NILLSTONE - UNIT 3 3/4 4-11 Amendment No. Jpp oss4

f-REACTOR COOLANT SYSTEN BASES 3/4.4.2 SAFETY VALVES i-The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia.

Each safety valve is designed l

to relieve 420,000 lbs per hour of saturated steam at the valve Setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure 1 -

condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operatir,g RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

In addition, the Cold Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.

During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2750 psia.

The combined relief capacity of all of.these valves is greater than the maximum surge rate resulting from a cocplete loss-of-load assuming no Reactor trip until the first Reactor Trip System Trip Setpoint is reached (i.e., no credit is taken for a direct Reactor trip on the loss-of-load) and also assuming no 4

operation of the power-operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI l

of the ASME Boiler and Pressure Code.

.1 3/4.4.3 PRESSURIZER l'

The requirement for the pressurizer to be OPERABLE, with a level less than j

or equal-to 89%, ensures that a steam bubble exists. The 89% level preserves the steam space for pressure control.

The 89% level has been established to ensure the capability to establish and maintain pressure control for steady state operation and to minimize the consequences of potential overpressure transients. The 89% pressurizer level does not establish an initial condition for an accident or transient analysis. Automatic and procedural controls e

adequately maintain pressurizer level to assure the validity of the FSAR Chapter 15 accident analyses.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance requires that during steady state

-operation, pressurizer level is maintained below the nominal upper limit to i

provide a minimum space for a steam bubble. The Surveillance is performed by observing the indicated level. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval has been shown by operating i

practice to be sufficient to regularly assess level for any deviation and.to -

ensure that a steam bubble exists in the pressurizer. Alarms are also available for early detection of abnormal level indications.

The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation.

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

J 0656 l

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i REACTOR COOLANT SYSTEM BASES 3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.

Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Code safety valves.

f Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.

Requiring the PORVs to be OPERABLE ensures that the capability for depressurization during safety grade cold shutdown is met.

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m-s MILLSTONE - UNIT 3 B 3/4 4-2a Amendment No.

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Docket No. 50-423 B16625 i

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Millstone Nuclear Power Station Unit No. 3 i

Proposed Revision to Technical Specification Pressurizer Level (PTSCR 3-31-97)

Packaround and Safety Assessment 3

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October 1997 t

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U.S. Nuclear Regulitory Commission B16625%ttachment 4\\Page1 I

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Background

The proposed change to Technical Specification 3/4.4.3, Pressurizer, replaces the pressunzer maximum water inventory requirement with a pressurizer maximum indicated level requiremont. The proposed change to Technical Specification Section 3.4.3 will:

Change the pressurizer requirement from:

" water volume of less than or equal to 92% (1656 cubic feet)"

to

" water level of less than or equal to 89% of instrument Nan" The associated surveillance, Surveillance 4.4.3.1, is changing to a level surveillance from a volume surveillance. This surveillance change is being done to nwke the surveillance consistent with the change to the requirement.

s The bases section is also being modified. The discussion regarding pressurizer levei is being replaced in its entirety. The current discussion and proposed discussion is shown below.

Current bases section wording:

"The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady-state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions. The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water volums also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system."

Proposed bases section wording:

"The requirament for the pressurizer to be OPERABLE, with a les el less thar. x equal to 89%, ensures that a steam bubble exists. The 89% leve, preserves the steam space for pressure control. The 89% level has been este..shed to ensure the mpahility to establish and maintain pressure control for steaa.y state operation and to minimize the consequences of potential overpressure transients. The 89%

pressurizer level does not establish an initial condition for an accident or transient analysis. Automatic and procedural controls adequately maintain pressurizer level to assure the validity of the FSAR Chapter 15 accident analyses."

"The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance requires that during steady state operation, pressurizer level is maintained below the nominal upper limit to provide a

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U.S. Nucl:ar Regulttory Commission q

. B16625\\ Attachment 4\\Page 2 j

minimum space for a steam bubble. The Surveillance is performed by observing the indicated level. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval has been shown by operating practice to be sufficient to regularly _ assess level for any deviation and to ensure that a steam bubble exists in the pressurizer, - Alarms are also available for early i

detection of abnormallevelindications."

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The proposed indicated water level requirement is more restrictive than the current volume requirement. This change is consistent with the Standard Technical Specifications (STS) '

for Westinghouse plants, NUREG 1431, Rev 1.

As in STS, the Millstone Unit No. 3

.l numerical value for the level requirement is equal to the trip setpoint for the pressurizer high water level reactor trip in Technical Specification Section 2.2.

i The design basis non-LOCA analyses. assume that pressurizer level is maintained at the i-

. programmed level of 28% at no load T increasing linearly to 61.5% at full power T The

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full power design basis LOCA analyses assume that pressurizer level is 62%. Those events that are analyzed to address pressurizer filling concerns are initiated assuming a higher initial pressurizer water level that accounts for 6% level uncertainty. The change does not impact normal plant operation. The change to the bases clarifies that the 89%

level requirement does not assure that the design basis assumptions are valid. Instead, as stated in the change to the bases, automatic and procedural controls maintain pressunzer

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level to assure the validity of the FSAR Chapter 15 accident analyses.

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SAFETY ASSESSMENT

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The change imposes a more restrictive. maximum pressurizer level requirement in Technicci Specification 3.4.3. The numerical value for the new requirement uses the

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l pressurizer high water level reactor trip setpoint from Technical Specification Section 2.2.

l This level requirement does not affect normal plant operation. Placing a more restrictive i

requirement on pressurizer level, which does not_ impact normal plant operation, can not j.

negatively affect the probability-of occurrence of a previously evaluated. malfunction..

Similarly, this change can neither change the consequences of a malfunction nor create the i

possibility of a malfunction of a different type, i

The change to the bases is consistent with the change to the specification and improves j

the description in the bases. Improving the bases, by describing better the relationship between the accident analyses and the pressurizer level requirements, can not affect the probability of occurrence of previously evaluated malfunctions.

Also, improving the

- description in the bases can neither change the consequences of a malfunction nor create the possibility of a malfunction of a different type.

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j The Technical Specification maximum pressurizer inventory requirement is being reduced to the numerical value for the reactor trip on pressurizer high level from Technical i

Specification Section 2.2.

The bases change reflects that automatic and procedural j

controls maintain pressurizer level to provide assurance that the design bases analyses are i

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U.S. Nucl ar Regulttory Commission B16625%ttachment 4\\Page 3

, valid. This does not modify plant operation. Therefore, the change can not negatively impact the probability of occurruace of the previously evaluated accidents. The Technical Specification and bases change does not reflect any change to plant operations nor any change to accident mitigation strategy. Therefore, the changes can neither affect the consequences of previously evaluated accidents nor introduce the possibility of an accident of a different type.

The change imposes a tighter ret'.riction on pressurizer level. The design basis non-LOCA analyses use the current pressurizer programmed level and the LOCA analysis uses 62% -

level for full power. Those events that are analyzed to address pressurizer filling concerns are initiated assuming a higher initial pressurizer water level that accounts for 6% level uncertainty. The bases change makes it clear that the validity of the design basis analyses are maintained by the automatic and procedural controls and not the less than or equal to 89% level Technical Specification requirement. Therefore, the change does not reduce the margin of safety.

Based on the above, the proposed Technical Specification change is safe.

Docket No. 50-423-B16625 l

Millstone Nuclear Power Station Unit No. 3 Proposed Revision to Technical Specification Pressurizei Level

' (PTSCR 3-31-97)

Sianificant Hazards Consideration and Environmental Considerations October 1997

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U.S. Nuclear Regulclory Commission B16625\\Attachmunt 5\\Page 1 Sianificant Hazards Consideration NNECO has reviewed the proposed revision in accordance with 10CFR50.92 and has conc!uded that the revision does not involve a significant hazards consideration (SHC).

The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not satisfied. The proposed revision does not involve a SHC because the revision would not:

1.

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

The Technical Specification maximum pressurizer inventory requirement in Technical Specification 3.4.3 is being changed to use the numerical value for the Reactor Trip setpca on pressurizer high water level in Technical Specification Section 2.2. This changes the requirement from a volume to a level requirament, is consistent with the Improved Standard Technical Specifications for Westinghouse plants, and represents a more restrictive level requirement than the current technical specification. The bases change clarifies that the 89% level requirement only assures that there is a steam bubble in the pressurizer. Also, the bases change states that pressurizer level is maintained by automatic and procedural controls to provide assurance thH the design basis analyses are valid. These changes do not modify plant operr. bon.

Lowering the maximum level requirement so that it is numerically cons'atent with the reactor trip setpoint, while clarifying the bases of the requirement, ran not involve a significant increase in the probability or consequence of an c.ccident previously evaluated.

Therefore, the proposed revision does not involve a significant increase in the probability or consequence of an accident previously evaluated.

2.

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

There are no hardware modifications associated with the change. The change does not modify the way that the plant is operated. The change modifies neither accident mitigation nor system response post-accident.

Therefore, the proposed revision does 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 change places a lower maximum pressurizer level requirement for the pressurizer. The change imposes the numerical setpoint value for the reactor trip on pressuriner high water level as the restriction on the pressurizer level. The change to the bases clarifiec that the 89% level requirement only ensures the existence of a

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U.S. Nucl:ar Regul: tory Commission B18625%ttachment 5\\Page 2 steam bubble and not the validity of the design basis analyses. The design basis non-LOCA analyses use the current programmed pressurizer level and the LOCA analysis uses 62% level for full power. Those events that are analyzed to address l

pressurizer filling concems are initiated assuming a higher initial pressurizer water level that accounts for 6% level uncertainty The bases dange makes it clear that the pressurizer level required to assure the validity of the design basis analyses is -

maintained by the automatic and procedural controls and not the less tha, or equal l

to 89% levelin the requirement.

Therefore, the moposed revision does not involve a significant reduction in a margin of safety.

In conclusion, based on the information provided, it is determined that the proposed revision does not involve an SHC, Environmental Considerations NNECO has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. The proposed revision does not involve -

a SHC, does not significantly increase the type and amounts of offluents that may be-released offsite, nor significantly increase individual or cumulative occupational radiation exposures. Based on the feregoing, NNECO concludes that the proposed revision meets the criteria delineated in 10CFR51.22(c)(9) for categorical exclusion from the requirements for environmental review.

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