Text

Request for Amendment to Facility Operating License DPR-28, Technical Specification Proposed Change No. 250 Re Scarm & Isolation Valve Closure Functions of Main Steam Line Radiation Monitors
	ML020810028
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	03/19/2002
From:	Balduzzi M; Vermont Yankee
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	BVY 02-18
	Download: ML020810028 (57)
	v • d • e

VERMONT YANKEE NUCLEAR POWER CORPORATION 185 OLD FERRY ROAD, PO BOX 7002, BRATTLEBORO, VT 05302-7002 (802) 257-5271 March 19, 2002 BVY 02-18 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555

Subject:

Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271)

Technical Specification Proposed Change No. 250 Scram and Isolation Valve Closure Functions of the Main Steam Line Radiation Monitors Pursuant to 10CFR50.90, Vermont Yankee (VY) hereby proposes to amend its Facility Operating License, DPR-28, by incorporating the attached proposed change into the VY Technical Specifications. This proposed change eliminates the reactor scram and main steam isolation valve closure requirements associated with the main steam line radiation monitors (MSLRMs) and modifies other requirements related to MSLRM trip functions.

Attachment I to this letter contains supporting information and the safety assessment of the proposed change. Attachment 2 contains the determination of no significant hazards consideration. Attachment 3 provides the marked-up version of the current Technical Specification pages. Attachment 4 is the retyped Technical Specification pages.

VY has reviewed the proposed Technical Specification change in accordance with 10CFR50.92 and concludes that the proposed change does not involve a significant hazards consideration.

VY has also determined that the proposed change satisfies the criteria for a categorical exclusion in accordance with IOCFR51.22(cX9) and does not require an environmental review.

Therefore, pursuant to 10CFR51.22(b), no environmental impact statement or environmental assessment needs to be prepared for this change.

Upon acceptance of this proposed change by the NRC, VY requests that a license amendment be issued no later than 180 days from the date of this letter for implementation within 60 days of its effective date. Issuance of an amendment by this date will support plant changes that are scheduled to be implemented prior to the next refueling outage.

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VERMONT YANKEE NUCLEAR POWER CORPORATION BVY 02-18 /Page 2 If you have any questions on this transmittal, please contact Mr. Gautam Sen at (802) 258-4111.

Sincerely, VERMONT YANKEE NUCLEAR POWER CORPORATION Michael A. Balduzzi Senior Vice President and Chief Nuclear Officer STATE OF VERMONT

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)ss WINDHAM COUNTY

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Then personally appeared before me, Michael A. Balduzzi, who, being duly sworn,,did stana tha;.he isSenidr Vice President and Chief Nuclear Officer of Vermont Yankee Nuclear Power C tio' th

l*I iy authorized to execute and file the foregoing document in the name and on the behalf o e

ont'*'

ie Power Corporation, and that the statements therein are true to the best of his knowledge

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Sally A. Sandstrum, Notary Public My Commission Expires February 10, 2003 Attachments cc:

USNRC Region I Administrator USNRC Resident Inspector - VYNPS USNRC Project Manager - VYNPS Vermont Department of Public Service

Docket No. 50-271 BVY 02-18 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 250 Scram and Isolation Valve Closure Functions of the Main Steam Line Radiation Monitors Supporting Information and Safety Assessment of Proposed Change

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BVY 02-18/Attachment I /Page I INTRODUCTION Description of Change The proposed changes to the Vermont Yankee Nuclear Power Station (VYNPS) Technical Specifications (TSs) eliminates the automatic reactor scram and main steam isolation valve (MSIV) closure functions of the main steam line radiation monitors (MSLRMs) and modifies other requirements related to MSLRM trip functions. Conforming changes are also being made to the Bases for these TSs. The proposed changes include the following TSs:

Table 3.1.1 Reactor Protection System (Scram) Instrument Requirements Table 4.1.1 Scram Instrumentation and Logic Systems Functional Tests - Minimum Functional Test Frequencies for Safety Instrumentation, Logic Systems and Control Circuits Table 4.1.2 Scram Instrument Calibration - Minimum Calibration Frequencies for Reactor Protection Instrument Channels Specification 4.2.F Mechanical Vacuum Pump Isolation Table 3.2.2 Primary Containment Isolation Instrumentation Table 4.2.2 Minimum Test and Calibration Frequencies - Primary Containment Isolation Table__4.7.2

_Instrumentation Table 4.7.2 Primary Containment Isolation Valves The proposed changes are consistent with the methodology of the NRC-approved BWR Owners Group Licensing Topical Report (LTR), NEDO-31400Al (NRC safety evaluation dated May 15, 1991).

Elimination of the trip functions identified in the LTR will result in a reduced potential for inadvertent reactor shutdowns and plant transients caused by spurious MSLRM actuation signals. This change also offers increased plant operational flexibility and reliability without compromising plant safety.

In addition, based on plant-specific analysis, all Group 1 isolation valve trip signals originating from the MSLRM are being eliminated. Other trip functions associated with the MSLRMs, including the alarm function and signals that trip and isolate the mechanical vacuum pump (MVP) are unchanged.

Surveillance requirements for the MSLRM are also unchanged.

The following Table I describes the specific TS changes and provides the basis for each change.

Purpose of the Change The most significant operational impact caused by existing MSLRM trip functions is the unnecessary scram and isolation of the reactor vessel. This action isolates the primary heat sink, imposes a significant transient on the vessel and results in safety-related system actuations. Subjecting the primary coolant pressure boundary to unnecessary vessel isolations diminishes plant reliability, complicates scram recovery and is adverse to plant safety. Eliminating the main steam line isolation, other Group 1 valve isolations, and scram functions from the MSLRM will help avoid undue vessel isolations during plant evolutions. Certain extenuating circumstances, such as instrument failures, chemistry excursions, and radiation monitor maintenance errors also have the potential for causing unnecessary MSLRM trips. The elimination of these trips would reduce the risk of the high radiation trip occurring from these operational situations, thereby increasing plant safety.

'NEDO-31400A, "Safety Evaluation for Eliminating the Boiling Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor," October 1992.

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BVY 02-18 / Attachment 1 /Page 2 In addition, as discussed in the referenced LTR, removal of trip functions permits continued use of the augmented offgas (AOG) system to process radioactive fission products following a design basis control rod drop accident (CRDA). Under these conditions, the plant operator maintains control over the release pathway and may, because of longer holdup times and offgas treatment, reduce levels of radioactivity released to the environs.

BACKGROUND NEDO-31400A This proposed change is based on the analysis provided in NEDO-31400A. That study demonstrated that removal of the MSLRM scram and isolation features would not exacerbate the consequences of the previously evaluated CRDA. General Electric prepared NEDO-31400A at the request of the BWR Owners Group to provide a bounding safety analysis that forms the basis for the elimination of the MSIV closure and scram functions of the MSLRMs.

In NEDO-31400A, a reevaluation of the role of the MSLRM in the CRDA analysis confirmed that removal of the MSLRM scram/isolation features would not compromise CRDA consequences.

NEDO-31400A has been evaluated for applicability to the Vermont Yankee Nuclear Power Station, and together with plant-specific analyses, support the Safety Assessment that follows.

The NEDO-31400A analysis considers the offsite dose consequences for two bounding accident sequences:

1)

A CRDA where the source term is not reduced, even though the MSIVs close, and the radionuclides enter the condenser at atmospheric pressure to leak at a rate of 1% volume/day.

2)

A CRDA where the MSIVs do not close and the radioactivity is processed through the AOG and released via the main stack.

The evaluation demonstrated that the radiological consequences of these accident sequences would remain below a fraction (25 percent) of the 10CFR100 limits without the scram and MSW closure functions of the MSLRMs. The resulting offsite doses for the first case were 4.3 rem thyroid, and 0.31 rem whole body. For the second case, the doses were comparable assuming AOG hold up times of l5days for Xenon and 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months for Krypton. The thyroid dose was 0 rem and the whole body dose was 0.55 rem. The LTR concluded that the reactor pressure vessel isolation function and scram function of the MSLRM are not required to ensure compliance with I0CFRl00. Additionally, it demonstrates that use of the AOG is an effective and acceptable method of controlling accident source terms.

The LTR also evaluated the potential effect on occupational exposure in the event of a sudden release of radioactive material from the fuel and concluded that the elimination of the scram/isolation features would have no adverse effect.

In a NRC Safety Evaluation dated May 15, 1991, and later incorporated into NEDO-31400A, the NRC staff concluded that removal of the MSLRM trips that automatically shut down the reactor and close the MSIVs was acceptable for referencing in plant-specific license amendment requests, provided that certain conditions were met. VYNPS's compliance with those conditions is discussed in the following section entitled, "Safety Assessment."

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BVY 02-18 / Attachment 1 / Page 3 VYNPS Design Features The VY MSLRMs consists of four redundant radiation detectors located on the outside of the main steam lines and external to the primary containment. The MSLRMs are designed to provide an early indication of gross fuel cladding failures. The original design basis of these monitors was to mitigate the release of radioactivity from failed fuel by providing a scram signal to terminate the initiating event and a MSIV closure signal to assure containment of the release. When a significant increase in the main steam line radiation level is detected, trip signals generated by the monitors are used to initiate a reactor scram and isolation of radioactive material released from the fuel.

For VY, the MSLRMs currently provide one of several possible signals (set at < 3 times background radiation at rated thermal power) to close "Group 1" primary containment isolation valves. At this same trip setting, the MSLRMs provide signals to trip the MVP, close its suction valve, and scram the reactor.

In addition, the MSLRMs provide an alarm (set at a nominal 1.5 times normal background radiation at rated thermal power) to alert the plant operators to off-normal conditions and take actions prescribed in plant procedures. The MSLRMs also provide trip signals to certain non-safety-related equipment.

The MVP is used to establish initial condenser vacuum or maintain partial vacuum when steam pressure is not adequate to operate the steam jet air ejector units. When employed during startup, the MVP takes suction on the main condenser and discharges air and the non-condensible gases to the vent stack via a discharge pipe. The MVP and its suction valve are designed to automatically de-energize and close, respectively, when radiation levels in the main steam lines exceed the MSLRM high setpoint.

VYNPS Accident Analysis The only design basis accident for which either the MSIV closure function or scram function of the MSLRMs is credited is the CRDA. The VYNPS UFSAR provides analyses of the CRDA in sections 14.6.2 and 14.9.2.4. The latter section analysis uses a TID-14844 source term and is essentially the same as the accident sequences of NEDO-31400A. In these scenarios no credit is taken for automatic reactor shutdown from the MSLRMs, and automatic isolation of the MSIVs is assumed to provide no benefit.

However, the MVP is assumed to trip and isolate. In conformance with NEDO-31400A, VY has re analyzed the CRDA, eliminating closing of all Group I isolation valves from a MSLRM actuation signal.

Comparison to Standard Technical Specifications Standard Technical Specifications (NUREG-1433) do not contain requirements for either the MSIV closure function or the reactor scram function of the MSLRMs. Requirements for the MVP are not included in the base specifications, but are handled on a supplemental, case-by-case basis.

Updated Final Safety Analysis Report (UFSAR)

The following VYNPS UFSAR sections provide additional background information:

1.6.2.20 - Main Steam Line Radiation Monitoring System 0

7.12.1 - Main Steam Line Radiation Monitoring System 7.12.2 - Off-Gas Radiation Monitoring System 9.4 - Gaseous Radwaste System 14.6.2 - Control Rod Drop Accident 14.9.2.4 - Control Rod Drop Accident S* II I

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BVY 02-18/ Attachment 1 /Page 4 SAFETY ASSESSMENT The following Table 1 provides a detailed description of each change, including the basis for the change and a safety assessment as necessary to supplement the analysis provided by NEDO-31400A.

The Change Numbers in the table's left-hand column correspond to the boxed (o) annotation numbers in, "Marked-Up Version of the Current Technical Specifications."

The basis for elimination of the reactor scram function and MSIV closure function of the MSLRMs is the analysis contained in NEDO-31400A; therefore, all the supporting bases contained in that LTR are not repeated here. The analysis of NEDO-31400A is applicable to and bounds VYNPS analysis of the CRDA. VY has performed a detailed comparison of the VYNPS CRDA analysis with the assumptions and conditions made in the bounding generic analysis.

That evaluation is summarized in Table 2.

Furthermore, VY has expanded its analysis of the CRDA to include elimination of the automatic closure of all Group 1 isolation valves on an actuation signal from the MSLRM. In addition to the eight MSIVs, the only other VY Group 1 isolation valves are the two main steam line drain valves and the two recirculation loop sample line valves. TS Table 4.7.2 identifies primary containment isolation valves by isolation group.

With the elimination of the Group 1 isolation valves closure function of the MSLRM, in addition to the MSIVs, the other Group 1 isolation valves that currently receive a high main steam line radiation closure trip are the two main steam line drain valves (designated as valves 2-74 and 2-77) and the two reactor water sample valves (also known as "recirculation loop sample line" valves and designated as valves 2-39 and 2-40). TS Table 4.7.2 identifies these valves as being in isolation Group 1, being normally closed, and staying closed upon receipt of a Group 1 initiation signal. Group 1 initiation signals are listed in Note 1 of TS Table 4.7.2.

Table 1 Change Current Technical Specifications Proposed Change 1

Current Technical Specifications (CTS) 3.1A Delete Trip Function #9 and associated requires plant operation in accordance with requirements from Table 3.1.1. Notes 7 and 8 Table 3.1.1, "Reactor Protection System of Table 3.1.1, which only pertain to this trip (Scram) Instrument Requirements."

Trip function, are also deleted.

Function

9 in Table 3.1.1 specifies requirements for a reactor scram initiated by a Insert the word "deleted" in place of Trip main steam line high radiation signal. Notes 7 Function #9 in Table 3.1.1 and in place of its and 8 of Table 3.1.1 are only applicable to this Notes 7 and 8.

trip function.

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BVY 02-18 / Attachment 1/Page 5 Table 1 (continued) change i

1Basis I Safety Assessment:

I The basis and safety assessment for elimination of the reactor scram trip function from main steam line radiation monitors are discussed in NEDO-31400A. Elimination of the scram trip function includes elimination of associated limiting conditions for operation (i.e., Trip Function

9 in Table 3.1.1).

Notes 7 and 8 of Table 3.1.1 have no applicability when Trip Function #9 is deleted from Table 3.1.1.

Note 7 does not impose any operational requirements and is only informational in explaining circuitry design. Therefore, Note 7 can also be deleted since is imposes no requirements and is only applicable to the trip function being eliminated.

Note 8 does impose operational requirements beyond the scope of the reactor trip function being eliminated in that it requires a MSLRM alarm set at a nominal 1.5 times normal background at rated power. This alarm requirement, however, is redundant to current TS Table 3.2.2, Note 9, which is unchanged.

Therefore, Note 8 can be deleted with no change in operability requirements, other than those associated with elimination of the scram function.

Change Current Technical Specifications Proposed Change 2

CTS 4.1.A requires that scram Remove Instrument Channel, "High Main instrumentation systems be functionally tested Steam Line Radiation" and associated in accordance with Table 4.1.1, "Scram requirements from Table 4.1.1.

The Table Instrumentation and Logic Systems Functional 4.1.1 Notes that are applicable to this Tests - Minimum Functional Test Frequencies Instrument Channel are also applicable to one for Safety Instrumentation, Logic Systems and or more of the other Instrument Channels Control Circuits."

Table 4.1.1 specifies an specified in this table, and are therefore instrument channel for "High Main Steam retained.

However, Note 2 is modified Line Radiation" with associated surveillance because of the elimination of the High Main requirements. Several Notes of Table 4.1.1 are Steam Line Radiation instrument channel of applicable to this instrument channel.

the reactor protection system.

Note 2 is changed to:

An instrument check shall be performed on reactor water level and reactor pressure instrumentation once per day.

This change to Note 2 of Table 4.1.1 deletes the existing phrase, "...and on streamline radiation monitors once per shift."

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BVY 02-18 /Attachment 1 /Page 6 Table 1 (continued) 2 Basis I Safety Assessment:

The basis and safety assessment for elimination of the reactor scram trip function from main steam line radiation monitors are discussed in NEDO-31400A. Elimination of the scram trip function includes elimination of associated surveillance requirements (i.e., the high main steam line radiation instrument channel in Table 4.1.1).

Note 2 is being revised because the high main steam line radiation trip function is being eliminated and all surveillance requirements (SRs) related to this function will no longer be applicable. Therefore, revising Note 2 to delete a SR for a function that is being eliminated is acceptable.

A similar SR for the primary containment isolation function of the MSLRM will remain in TS Table 4.2.2.

Change Current Technical Specifications Proposed Change 3

CTS 4.1.A requires that scram Delete Instrument Channel, "High Main Steam instrumentation systems be calibrated in Line Radiation" and associated requirements accordance with Table 4.1.2, "Scram from Table 4.1.2. Note 3 of Table 4.1.2 only Instrument Calibration - Minimum Calibration pertains to this instrument channel and is also Frequencies for Reactor Protection Instrument being eliminated from this table.

Channels." Table 4.1.2 specifies an instrument channel for "High Main Steam Line Insert the word "deleted" in place of Note 3 of Radiation" with associated surveillance Table 4.1.2.

requirements.

Note 3 of Table 4.1.2 is applicable to this instrument channel.

Basis / Safety Assessment:

The basis and safety assessment for elimination of the reactor scram trip function from main steam line radiation monitors are discussed in NEDO-31400A. Elimination of the scram trip function includes elimination of associated surveillance requirements (i.e., the high main steam line radiation instrument channel in Table 4.1.2).

Note 3 is being deleted because no SRs applicable to a scram function from high main steam line radiation will remain in Table 4.1.2.

SRs for the MSLRM trip function will remain in TS Table 4.2.2.

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BVY 02-18 / Attachment I / Page 7 Table 1 (continued)

Change Current Technical Specifications Proposed Change 4

CTS Surveillance Requirement 4.2.F, SR 4.2.F is restructured into two parts. The Mechanical Vacuum Pump," currently does existing SR 4.2.F will become SR 4.2.F.2 and not explicitly require periodic instrument a new requirement is added as SR 4.2.F.1.

check, functional testing and calibration of the Proposed SR 4.2.F.1 states:

high main steam line trip function of the MVP.

The High Main Steam Line Radiation trip function of the mechanical vacuum pump shall be functionally tested and calibrated in accordance with Table 4.2.2.

A Note 12 is added to the High Main Steam Line Radiation trip function in Table 4.2.2 to state:

This trip function is applicable to the mechanical vacuum pump; however, it is not applicable to the primary containment Group I isolation valves (i.e., main steam isolation, main steam line drain, and recirculation loop sample line isolation valves).

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BVY 02-18 / Attachment 1 / Page 8 Table 1 (continued)

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4 Basis / Safety Assessment:

SR 4.2.B requires that instrumentation and logic that provides primary containment isolation be functionally tested and calibrated in accordance with TS Table 4.2.2. Although the High Main Steam Line Radiation trip function identified in TS Table 4.2.2 is tested and calibrated as specified, with the elimination of this actuation signal for primary containment Group 1 isolation valves, this SR for the high main steam line radiation monitor might otherwise also be eliminated. However, these testing requirements are necessary to assure operability of the MVP trip logic when it is in service. Therefore, the surveillance requirement for the high main steam line radiation trip is retained in Table 4.2.2, but made applicable only to the MVP.

The restructuring of SR 4.2.F to add an explicit requirement for periodic functional test and calibration of the High Main Steam Line Radiation trip function of the MVP is acceptable because these requirements maintain operability of the MVP trip function. This change makes it explicit.

The High Main Steam Line Radiation trip function currently isolates Group 1 primary containment isolation valves (identified in TS Table 4.7.2). This change eliminates the Group 1 isolation valve closure function of the MSLRMs.

The addition of Note 12 to Table 4.2.2 clarifies the applicability of the High Main Steam Line Radiation trip function. New Note 12 provides design information to the operator that there are no longer any Group I isolation valve trips associated with the MSLRMs.

This Note is acceptable because it is informational and does not impose a change to operating requirements.

With the elimination of the MSLRM actuation signal to all Group 1 isolation valves, this change is acceptable since it clarifies that this trip function is applicable to the MVP and not Group 1 containment isolation valves. The basis and safety assessment for the elimination of the Group 1 closure trip function of the MSLRM are supported by NEDO-31400A and plant-specific analyses.

BVY 02-18 / Attachment I / Page 9 Table 1 (continued)

Change Current Technical Specifications Proposed Change II CTS 3.2.B requires that when primary containment integrity is

required, in accordance with Specification 3.7, the instrumentation that initiates primary containment isolation shall be operable in accordance with Table 3.2.2, "Primary Containment Isolation Instrumentation." Trip Function "High Main Steam Line Radiation" in Table 3.2.2 specifies requirements for a primary containment isolation initiated by high main steam line radiation.

For this trip function, the column, "Required ACTION When Minimum Conditions For Operation Are Not Satisfied (Note 2)," specifies Action "B" from Note 2.

Basis / Safety Assessment:

The Action specified under the column "Required ACTION When Minimum Conditions For Operation Are Not Satisfied (Note 2),"

for the High Main Steam Line Radiation trip function is changed from "B" to "B or C."

A new Action "C" is added to Note 2 of Table 3.2.2. New Action "C" requires:

The actions required by Specification 3.2.F.1 shall be taken immediately.

This change adds an alternative protective action to be taken in the event that minimum conditions for operation are not satisfied for the High Main Steam Line Radiation trip function.

Since this trip function will no longer apply to Group 1 containment isolation valves, the only remaining applicable TS equipment is the MVP. TS LCO 3.2.F addresses the operability requirements of the MVP. LCO 3.2.F.1 requires that whenever the MSIVs are open, the MVP shall be capable of being automatically isolated and secured by a signal of high radiation in the main steam line tunnel or shall be manually isolated and secured. Since this trip function is no longer applicable to the Group 1 isolation valves, manually isolating and securing the MVP is the appropriate action to close this potential release pathway.

New Action "C" is acceptable because immediately taking action to manually isolate a potential discharge pathway accomplishes the safety function provided by the High Main Steam Line Radiation instrumentation because the safety function of this instrumentation is to limit fission product release during and following postulated design basis events. Isolating the affected line completes the protective function of this instrumentation.

Therefore, this is an acceptable alternative to initiating an orderly load reduction and placing the reactor in "Hot Standby" (i.e.,

taking the actions prescribed by Action "B" of Note 2, Table 3.2.2).

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BVY 02-18 / Attachment I / Page 10 Table 1 (continued)

Change Current Technical Specifications Proposed Change I

CTS 3.2.B requires that when primary containment integrity is

required, in accordance with Specification 3.7, the instrumentation that initiates primary containment isolation shall be operable in accordance with Table 3.2.2, "Primary Containment Isolation Instrumentation." Trip Function "High Main Steam Line Radiation" in Table 3.2.2 specifies requirements for a primary containment isolation initiated by high main steam line radiation. Note 8 of Table 3.2.2 is applicable to this trip function.

Basis / Safety Assessment:

Note 8 of Table 3.2.2 is changed from:

Channel shared by the Reactor Protection and Primary Containment Isolation Systems.

To:

This trip function is applicable to the mechanical vacuum pump; however, it is not applicable to the primary containment Group I isolation valves (i.e., main steam isolation, main steam line drain, and recirculation loop sample line isolation valves).

This change involves the elimination of the MSLRM closure function for the MSIVs and other Group 1 isolation valves. With the elimination of the scram trip function and Group 1 isolation valve closure function of the MSLRM, current Note 8 is no longer correct, since the MSLRMs will no longer provide an input signal to the reactor protection system (RPS) or primary containment isolation systems. Therefore, current Note 8 is being deleted and replaced with a new Note 8 as specified above.

New Note 8 clarifies for the operator that the High Main Steam Line Radiation trip function exists for the MVP and that there is no Group I isolation valve closure trip associated with the MSLRM. This change is acceptable because it imposes no change to operating requirements, but merely adds clarification for the benefit of the operator in recognizing the design change modifying this circuitry.

The basis and safety assessment for the elimination of the MSIV closure and RPS trip functions of the MSLRM are discussed in NEDO-31400A. Plant-specific analysis of the CRDA also supports elimination of the MSLRM trip function for all Group 1 isolation valves.

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BVY 02-18 / Attachment 1 / Page 11 Table 1 (continued)

Change Current Technical Specifications Proposed Change 7

CTS 4.7.D requires that certain primary Note 1 of Table 4.7.2 is revised to delete the containment isolation valves be tested in high main steam line radiation signal from the accordance with Table 4.7.2, "Primary list of signals that initiate closure of Group 1 Containment Isolation Valves," at least once isolation valves. Item 2 in the listing of Group per operating cycle. Table 4.7.2 identifies the 1 signals in Table 4.7.2 is changed to:

eight Main Steam Line Isolation valves (2 80A-D & 2-86A-D), two Main Steam Line

2. Deleted Drain (2-74, 2-77), and two Recirculation Loop Sample Line (2-39, 2-40) as Group 1 isolation valves and specifies associated surveillance requirements. Note 1 of Table 4.7.2 specifies six types of initiation signals applicable to Group 1 isolation valves. In this Note, Group 1, item 2 includes "high main steam line radiation."

Basis / Safety Assessment:

This change eliminates the closure function of the MSLRMs for the Group 1 isolation valves.

By deleting the MSLRM actuation signal, it is recognized that the Group 1 isolation valves no longer automatically close upon receipt of a main steam line high radiation trip signal.

The basis and safety assessment for the elimination of the Group I primary containment isolation valve closure trip function of the MSLRM are discussed in NEDO-31400A and plant specific analyses.

Change Current Technical Specifications Proposed Change 8

The TS Bases provide explanation and Associated changes to the TS Bases are being rationale for associated TS requirements, and made to conform to the changed TS and to add in some

cases, how they are to be clarity to existing requirements.

implemented.

Basis / Safety Assessment:

This proposed change revises the Bases of the VYNPS TSs to incorporate supporting information for the proposed TS changes. Bases changes are made for clarity purposes and to conform to the changes being made to the associated Specifications. Bases do not establish actual requirements, and as such do not change technical requirements of the TS. The Bases changes are therefore acceptable, since they administratively document the reasons and provide additional understanding for the associated TSs.

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BVY 02-18 / Attachment 1 / Page 12 Conformance to NRC Safety Evaluation Conditions The NRC issued a Safety Evaluation accepting NEDO-31400A for use as a reference in license amendment applications, provided the following three conditions are satisfied.

Condition 1:

The applicant demonstrates that the assumptions with regard to input values (including power per assembly, Chi/Q, and decay times) that are made in the generic analysis bound those for the plant.

Condition 2:

The applicant includes sufficient evidence (implemented or proposed operating procedures, or equivalent commitments) to provide reasonable assurance that increased significant levels of radioactivity in the main steam lines will be controlled expeditiously to limit both occupational doses and environmental releases.

Condition 3:

The applicant standardizes the MSLRM and offgas radiation monitor alarm setpoint at 1.5 times the nominal nitrogen-16 background dose rate at the monitor locations, and commits to promptly sample the reactor coolant to determine possible contamination levels in the plant reactor coolant and the need for additional corrective actions, if the MSLRM or offgas radiation monitors or both exceed their alarm setpoints.

The following discussion addresses how VYNPS satisfies each of the above three conditions.

NEDO-31400A vs. VYNPS-Specific Comparison Table An evaluation was performed by VY to compare the input parameters used in NEDO-31400A against VYNPS site-specific values. The evaluation demonstrates that VYNPS off-site doses for a CRDA are bounded by the values in NEDO-31400A. The results of this evaluation show that calculated VY off-site doses would be less than the off-site doses in the conservative analysis presented in NEDO-31400A. The comparison between the NEDO and the VYNPS site-specific results is summarized in Table 2 below.

Table 2 presents the critical input parameters that were used by General Electric in NEDO-31400A for the analyses of the CRDA. NEDO-31400A demonstrated that the Standard Review Plan (SRP) dose criteria could be met with the MSIV closure and reactor trip functions removed from the MSLRMs. The SRP dose criteria for this event are 6 rem whole body and 75 rem thyroid, or 25% of the IOCFR100 criteria2.

The CRDA was analyzed in two scenarios in the LTR. The first scenario assumed that the fission products released from the failed fuel and the reactor coolant were isolated in the condenser, and were subsequently leaked to the environment via a ground level release from the turbine building. This scenario is essentially the same as is presented in the VYNPS UFSAR (Section 14.9.2.4), and was intended to cover the case of a manual isolation of the MSIVs and the SJAE. The MVP was assumed to be automatically isolated by the MSLRMs. The second scenario assumed no MSIV or SJAE isolation 2 The SRP criteria are only being used as one acceptable method of evaluating risk. Use of this method is not a commitment to the SRP and does not incorporate SRP criteria or methodology into VYNPS' licensing basis.

BVY 02-18 / Attachment 1 / Page 13 and assumed that the fission products were released after appropriate holdup and processing in the AOG system. This scenario results in a release from the primary vent stack.

Table 2 compares the VYNPS site-specific input parameters to the NEDO-31400A values. In all cases the NEDO values are either bounding or equivalent to the VYNPS values.

The parameters presented in Table 2 are discussed below.

The power level of a single rod determines the fission product inventory of that rod. A peaking factor of 1.5 is applied to the average rod. The value used in NEDO-31400A was 0.12 MW/rod. The VYNPS UFSAR (Section 14.9.2) uses a power level of 1665 MWt for the analysis of the CRDA.

The power level per rod is 0.0754 MW for 8x8 assemblies and 0.0611 MW for 9x9 assemblies.

When multiplied by a peaking factor of 1.5 the results are values of 0.113 and 0.0917 MW/rod for 8x8 and 9x9 assemblies, respectively, which are within the NEDO values.

The number of fuel rods failed due to the control rod drop will determine the quantity of fission products released from the fuel gaps to the reactor coolant. NEDO-31400A assumed that 850 rods in 8x8 assemblies failed. The VYNPS UFSAR (Section 14.6.2) reports that 850 rods in 8x8 assemblies would fail or 1000 rods in 9x9 assemblies. When multiplied by the power level per rod above, the 8x8 assembly results in the highest equivalent MW for the failed rods (850 x 0.113 > 1000 x 0.0917).

The 8x8 assemblies result in 96.05 MW of equivalent fission product inventory in the failed rods.

This can be compared to the NEDO result if the 850 rods failed is multiplied by the 0.12 MW/rod.

This results in 102 MW of equivalent fission product inventory in the failed rods. The NEDO value is therefore bounding by a factor of 1.06.

"* The fraction of fuel melted due to the control rod drop will also add to the fission products released to the reactor coolant. NEDO-31400A assumed that 0.77% of the failed rods contained melted fuel (i.e.

the equivalent of 0.0077 x 850 = 6.54 rods melted). The VYNPS UFSAR reports no fuel rod melts for VYNPS. Since the NEDO assumed that 100% of the noble gases (i.e., 10 times the release from a failed rod) and 50% of the iodines (i.e., 5 times the release from a failed rod) would be released from melted fuel, the 6.54 melted rods would be equivalent to 65.4 failed rods for the noble gases and 32.7 failed rods for the iodines. NEDO-31400A bounds the VYNPS analysis in this regard by a factor of 1.08 [i.e. (850 + 65.4) / 850] for whole body doses from noble gases, and by a factor of 1.04 [i.e. (850

+ 32.7) / 850] for thyroid doses from iodines.

The atmospheric dispersion parameter (X/Q) used in NEDO-31400A for Scenario I was 2.5 x 10-3 s/m 3. This was based on a ground level release from the turbine building with the receptor at the exclusion area boundary (EAB). The VYNPS value for a turbine building release is 1.7 x 10-3 s/m 3 and was determined based on actual site meteorology. The NEDO value is bounding by a factor of approximately 1.5.

The holdup time of the off-gas system will determine the fraction of the noble gas fission products that enter the condenser that are subsequently released from the primary vent stack in Scenario 2.

NEDO-31400A determined that 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of holdup for krypton and 6 days for xenon gases would result in a dose that met the 6 rem whole body criterion (3.5 rem from Kr and 1 rem from Xe for a total of 4.5 rem). The VY design holdup times are 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for krypton and 16.6 days for xenon based on 6 charcoal beds in operation and a 30 scfm condenser air inleakage rate. These holdup times result in a total whole body dose of 0.35 rem (0.1 rem from Kr and 0.25 rem from Xe) when evaluated based on the same input parameters as used in NEDO-31400A. The NEDO doses due to

. I I I

I

BVY 02-18 / Attachment 1 / Page 14 off-gas system holdup are bounding by a factor of 12.8. No credit was taken for the fact that VY normally operates with seven charcoal beds in series and at condenser inleakage rates which usually range below 20 scfm. These two factors will extend the holdup in the VYNPS system by a factor of 1.75 for both krypton and xenon. There is no thyroid dose to consider from Scenario 2 since all of the iodine would be trapped in the charcoal delay beds.

The atmospheric dispersion factor used in the NEDO report for Scenario 2 was 3 x 10'4 s/m 3. This was based on a stack release with the receptor at the EAB. The VYNPS value determined from site specific meteorological data is 2.0 x 104 s/rn3. The NEDO value is bounding by a factor of 1.S.

Combining the factors above, VYNPS is conservatively bounded by the NEDO input assumptions by approximately a factor of 1.7 for both CRDA scenarios.

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BVY 02-18 / Attachment 1 / Page 15 TABLE 2 CONTROL ROD DROP ACCIDENT COMPARISON OF KEY INPUT PARAMETERS NEDO-31400A VS. VERMONT YANKEE-SPECIFIC Input Parameter NEDO-31400A Value Vermont Yankee Value Power Level (MW/rod) 0.12 (includes peaking 8x8 fuel = 0.113' factor of 1.5) 9x9 fuel = 0.0917' (values include peaking factor of 1.5)

Number of Rods Failed 850 (8 x 8 fuel) 850 (8 x 8 fuel) 1000 (9 x 9 fuel)b Melted Fuel Fraction 0.0077 (based on failed 0

rods)

I Scenario I - MSIV Closure Dispersion Parameter, XIQ 25 3

1.7x 10 3 (s/m3) 2.5 x 10 17x0.

Scenario 2 - No MSIV Closure Off gas System Holdup 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for Krc 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for Kr' Time 6 days for Xec 16.6 days for Xed Dispersion Parameter, X/Q 3.0 x 1 2.0 x 104e (s/m 3)

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a. Based on a power level of 1665 MWt, 368 assemblies in the core, and 60 fuel rods in 8x8 assemblies and 74 fuel rods in 9x9 assemblies.

b.

1000 rods of 9 x 9 fuel would have a slightly lower fission product inventory than 850 rods of 8 x 8 fuel.

c. These holdup times resulted in a whole body dose of less than 6 rem. A Kr holdup time of 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months was required to produce a whole body dose equivalent to Scenario 1.

d. Based on a condenser air inleakage rate of 30 scfm and 6 charcoal beds in operation.

e. Value based on fumigation condition existing.

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BVY 02-18 / Attachment I / Page 16 CRDA - Recirculation Loop Sample Lines The elimination of the automatic closure of Group I isolation valves would add a potential release pathway from a CRDA that was not included in the LTR. This pathway includes release from the recirculation loop sample lines in the reactor building. These sample lines contain cold reactor coolant which is released to a sample sink contained in a ventilation hood. The hood is exhausted via the reactor building ventilation system to the primary vent stack. The potential doses from this pathway were analyzed assuming no credit for manual isolation of the sample lines or for iodine removal by the standby gas treatment system. The resulting doses at the EAB were 3.0 rem to the thyroid and 0.015 rem to the whole body. Even when added to the doses which result from Scenario I or 2 in Table 2, these results are well below the SRP criteria of 75 rem thyroid and 6 rem whole body.

Occupational Doses and Environmental Releases The NRC Safety Evaluation for NEDO-31400A included a condition (identified as Condition #2 above) to limit both occupational doses and environmental releases. VY's response to this condition follows:

Occupational Doses VYNPS has design features, programs and procedures in place that will provide reasonable assurance that significantly increased levels of radioactivity in the main steam lines will be controlled expeditiously to limit occupational doses.

During reactor power operations, radioactive materials in the main steam lines are routed through the main condenser and SJAEs to the AOG system.

The recombiners and charcoal delay tanks of this system are in a separate shielded building that is designed to keep occupational doses ALARA (FSAR Section 9.4). In addition, important equipment is located in shielded compartments for isolation and maintenance.

The VYNPS programmatic and procedural controls include implementing procedures PP 7401 "Fuel Reliability Program," PP-7500 "Radiation Protection Program," OP-3140 "Alarm Response," ON-3152 "Off Gas High Radiation," and OT-3112, "Main Steam Line High Radiation."

These (and other procedures as necessary) will be revised as appropriate to incorporate specific considerations needed to change isolation of the main steam lines from an automatic action to a manual action for a condition involving high main steam line radiation. These actions will limit potential exposures to radioactive materials.

The recirculation loop sampling lines flow into a sample sink enclosed in a ventilation hood. Gases released from the liquid are collected by the ventilation hood and are exhausted to the primary vent stack. It is likely that any significant increase in gaseous releases will be detected by the reactor building ventilation radiation monitoring system and the reactor building area airborne monitoring system before concentrations in the reactor building become an occupational dose issue. In addition, area monitors in the reactor building will detect significant airborne concentrations of radioactivity and provide warning to occupants in that building.

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BVY 02-18/Attachment l/Page 17 Environmental Releases VYNPS has installed and currently operates an AOG system for processing the fission product gases from the condenser via the SJAE (FSAR Section 9.4). The AOG contains approximately 90,000 lbs. of charcoal distributed in seven equivalent tanks. The design basis holdup is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for krypton and 16.6 days for xenon during normal operation.

This design basis is based on the use of six of the seven charcoal tanks and a condenser air inleakage flow rate of 30 scfm. The actual holdup is normally greater than these design values since all seven tanks are normally used and the air inleakage rate usually averages less than 20 scfm (charcoal holdup time is directly proportional to the air flow rate). In addition, the charcoal tanks are configured such that either of the first two tanks may be valved out of service without impacting the operation of the system. Use of the AOG system provides operators with greater flexibility and extended time periods over which decisions may be made regarding the release of fission gases to the environment.

During normal operation, the VYNPS TSs limit the off-gas rate into the AOG to 160,000 uCi/s (30 minute decay equivalent). Procedures PP-7401 "Fuel Reliability Program," and ON-3152 "Off Gas High Radiation" require that actions be taken to limit off-gas rates to one-half of this value, or 80,000 uCi/s. The SJAE monitor, which is located between the condenser and the AOG system, measures these off-gas rates. These rates would be well below any applicable limits of 10CFR20 by the time the gas has passed through the AOG.

The AOG post-treatment (noble gas),monitors are located downstream of the charcoal delay tanks and monitor the noble gas levels between the tanks and the final section of delay pipe before release to the main stack. It takes the gas over an hour to pass through this final section of delay pipe. These monitors automatically isolate the off-gas line to the stack within 30 minutes if the off-gas activity exceeds the isolation setpoint. This setpoint is determined by the ODCM based on an instantaneous dose rate of 500 mrem/y (0.057 mrem/h) to the whole body or 3000 mrem/y to the skin of an off-site receptor.

The recirculation loop sampling lines flow into a sampling sink in the reactor building.

The liquid released to the sink is below 2120 and there will be no flashing of steam.

Iodine and particulates from the recirculation loop sampling lines will remain with the liquid, which flows from the sink to a contained sump in the reactor building. The ventilation exhaust hood surrounding the sink will capture any noble gas or iodine that is released into the air. Any significant quantities of airborne radioactivity will be detected by the reactor building ventilation radiation monitoring system, the reactor building area airborne monitoring system, and the primary vent stack radiation monitoring system. The reactor building ventilation radiation monitoring system will automatically isolate normal reactor building exhaust if the trip level is reached, and will initiate operation of the standby gas treatment system (SBGT). The SBGT contains high efficiency particulate and charcoal filters, which will remove particulate and iodine radioactivity before release to the environment.

The combination of off-gas processing systems, process radiation monitors, and plant procedures provides significant flexibility in controlling environmental releases from VYNPS.

These features will continue to provide adequate control of environmental releases. Manual isolation of the main steam lines will remain as one possible operator action to limit releases. Existing procedures related to gaseous waste processing will be I

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BVY 02-18 / Attachment I / Page 18 reviewed and revised as necessary to support the change from automatic to manual isolation of the main steam lines.

Radiation Monitor Setpoint and Additional Corrective Actions The NRC Safety Evaluation for NEDO-31400A included a condition (identified as Condition #3 above) to standardize MSLRM and off-gas radiation monitor alarm settings at 1.5 times background; and to commit to promptly sample the reactor coolant to determine activity levels in reactor coolant and determine the need for additional corrective actions, if the MSLRMs or off-gas monitors (or both) exceed their alarm septoints. VY's response to this condition follows:

The MSLRM alarm setpoint is currently a nominal 1.5 times the normal, full power background radiation level from N-16 (OP-4617, "Calculation of Chemistry Controlled Setpoints'). The setpoint will remain at that level as required by Note 9 of TS Table 3.2.2. VYNPS procedures (OT-3112 "Main Steam Line High Radiation") presently require that a reactor coolant sample for 1-131 be obtained if this alarm setpoint is reached.

The SJAE pre-treatment high radiation alarm is currently set to satisfy VYNPS TS 3.8.K, and the ODCM Section 8.2.2, which limit the gross radioactivity release rate from the SJAE to 160,000 uCi/s (after 30 minutes of decay). As stated in the Bases of the TS, "Restricting the gross radioactivity release rate of gases from the main condenser SJAE provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of IOCFR100 in the event this effluent is inadvertently discharged directly to the environment without treatment."

Additionally, this setpoint serves to limit buildup of fission product activity within station systems which would result if high fuel leakage were to be permitted over extended periods.

The SJAE pre-treatment radiation monitor is much more sensitive than the MSLRMs due to the decay of the short half-life N-16 (-7 seconds) and other activation gases. The total delay time from the MSLRMs to the SJAE pre-treatmen.t off-gas monitors is approximately 3 minutes (OP-2613 "Sampling and Analysis of Off Gas System"). This delay allows almost all of the N-16 and most of the other activation gases to decay before reaching the SJAE pre-treatment off-gas monitors. Therefore, the background radiation level at the off-gas monitors is very low, and setting these monitors to alarm at 1.5 times the background dose rate is not reasonable and would lead to numerous alarms at the very low off-gas rates currently experienced during normal operations.

VYNPS procedures currently require that an off-gas sample and a reactor coolant sample be taken and analyzed within four hours whenever the off-gas rate shows an increase of 5,000 uCi/s or 25% whichever is greater (OP-2150 "Advanced Off Gas System," OP-2613 "Sampling and Analysis of Off Gas System,"

ON-3152 "Off Gas High Radiation," PP-7410 "Fuel Reliability Program"). The VYNPS TSs also require an off-gas sample at the same off-gas rate increase and within the same four hour time limit (TS 4.8.K.2).

The SJAE off-gas monitor has an ERFIS alarm setpoint that supports this sampling requirement (OP 2613 "Sampling and Analysis of Off Gas System").

A SJAE pre-treatment off-gas monitor alarm setpoint based on an off-gas release rate increase of 5,000 uCi/s is very conservative. It is a factor of 32 below the TS LCO and is much more limiting than the MSLRM alarm setpoint of a nominal 1.5 times normal background at rated thermal power. The normal background reading on the SJAE off-gas monitors is in the range of 6-9 mR/h. This is based on operation with very low off-gas rates of approximately 200-300 uCi/s. Based on actual data, the monitor reading for an off-gas rate of 5,000 uCi/s would be approximately 70 mR/h, or about a factor of 10 higher than the background value. The actual setpoint value will depend on the radionuclide mix and the off-gas flow rate and is determined by a plant procedure (OP-2613 "Sampling and Analysis of Off Gas System"). An

BVY 02-18/Attachment 1 /Page 19 off-gas monitor that is set to alarm at a 5,000 uCi/s or 25% increase, whichever is greater, will ensure that even minor fuel performance changes are readily detected. Therefore, VY proposes to maintain its SJAE off-gas monitor alarm setting at a 5,000 uCi/s or 25% increase, whichever is greater. This is acceptable because these settings will ensure that even minor fuel performance changes are readily detected.

Conclusion/Summary In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) such activities will be conducted in compliance with the Commission's regulations; and (3) the issuance of the requested license amendment will not be inimical to the common defense and security or to the health and safety of the public.

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Docket No. 50-271 BVY 02-18 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 250 Scram and Isolation Valve Closure Functions of the Main Steam Line Radiation Monitors Determination of No Significant Hazards Consideration 1

1 1 ;I I

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BVY 02-18 / Attachment 2 / Page 1 Description of amendment request:

The proposed change removes requirements from the Technical Specifications (TSs) regarding main steam line radiation monitor (MSLRM) trips that automatically scram the reactor and close Group I isolation valves. Associated surveillance requirements would also be removed from TSs. In addition, conforming changes are being made to the Bases for these TSs.

Justification for removal of the MSLRM trip function is described in Licensing Topical Report (LTR)

NEDO-31400A, "Safety Evaluation For Eliminating the Boiling Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor." The subject LTR was accepted for referencing by the NRC staff in a Safety Evaluation dated May 15, 1991.

Plant-specific analysis of the design basis control rod drop accident (CRDA) supports elimination of all Group 1 isolation signals from the MSLRM.

Basis for No Significant Hazards Determination:

Pursuant to IOCFR50.92, Vermont Yankee has reviewed the proposed change and concludes that the change does not involve a significant hazards consideration since the proposed change satisfies the criteria in 1 OCFR50.92(c). These criteria require that the operation of the facility in accordance with the proposed amendment will not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The discussion below addresses each of these criteria and demonstrates that the proposed amendment does not constitute a significant hazard.

The proposed change does not involve a significant hazards consideration because the changes would not:

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

The proposed change involves the removal of the existing scram function and Group 1 isolation valve closure functions of the MSLRM. The purpose of the MSLRM reactor scram and Group 1 isolation signals is to mitigate the radiological consequences of gross fuel failures.

These functions do not serve as initiators for any of the accidents evaluated in the Updated Final Safety Analysis Report (UFSAR).

The MSLRM scram function is not credited in the UFSAR, and the Group 1 isolation trip function of the MSLRMs was only assumed in one design basis event-the control rod drop accident. Because these functions are not initiators of accidents, their removal does not increase the probability of occurrence of previously evaluated accidents.

Using the revised methodology of NRC-accepted NEDO-31400A and plant-specific analysis, it can be demonstrated that the consequences of a previously evaluated accident are not significantly increased by the proposed changes. Also, because there is no accident analysis that relies on the high radiation scram of the reactor protection system, its removal has no impact on the consequences of accidents previously evaluated. The results of the Vermont Yankee Nuclear Power Station's design basis control rod drop analysis are bounded by the NEDO-31400A results, which are well within the exposure guideline values in 10CFR Part 100.

Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated t

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BVY 02-18 / Attachment 2 / Page 2

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

The proposed changes to the plant involve limited changes to protective circuitry, but do not involve any plant hardware changes that could introduce any new failure modes. The changes will not affect non-MSLRM scram and isolation functions. In addition, the MSLRMs will remain active for other trip/isolation functions, and these monitors will still alarm in the control room to alert operators to off-normal conditions. The direct impact on the plant is that these particular trip functions (i.e., Group 1 closure and reactor scram) will no longer actuate. In demonstrating acceptable radiological consequences, the reconstituted design basis control rod drop accident analysis (performed in accordance with the methodology of NEDO-31400A) does not rely upon those trip functions being eliminated. The proposed changes were evaluated specifically for VYNPS and are enveloped by the NEDO-31400A analysis.

Therefore, the removal of the Group I isolation valve closure and scram functions of the MSLRMs does not create the possibility of a new or different kind of accident than those previously evaluated.

3) Involve a significant reduction in a margin of safety.

The proposed changes do not significantly increase calculated off-site dose consequences.

Furthermore, the changes will improve the overall reliability of the plant when compared to the existing system, since the proposed changes will reduce the chances of an unnecessary plant transient occurring as a result of an inadvertent MSLRM scram or MSIV closure.

A reliability assessment of the elimination of the MSLRM scram function on reactivity control failure frequency and core damage frequency was performed in NEDO-31400A. The results of this analysis indicate a negligible increase in reactivity control failure frequency with the deletion of the MSLRM trip function. However, this increase is offset by the reduction in the transient initiating events (inadvertent scrams). This reduction in transient initiating events represents a reduction in core damage frequency and, thus, results in a net improvement to safety.

Removal of the Group 1 closure and scram functions of the MSLRMs does not significantly increase the consequences of any design basis accident, including CRDA. Other trip signals for the reactor protection system and primary containment isolation valves are unaffected. Therefore, there is no significant reduction in the margin of safety as a result of this Technical Specification change.

Therefore, this change does not involve a significant reduction in a margin of safety.

Conclusion On the basis of the above, VY has determined that operation of the facility in accordance with the proposed change does not involve a significant hazards consideration as defined in IOCFR50.92(c), in that it: (1) does not involve a significant increase in the probability or consequences of an accident previously evaluated; (2) does not create the possibility of a new or different kind of accident from any accident previously evaluated; and (3) does not involve a significant reduction in a margin of safety.

Docket No. 50-271 BVY 02-18 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 250 Scram and Isolation Valve Closure Functions of the Main Steam Line Radiation Monitors Marked-up Version of the Current Technical Specifications i!t l

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VYNPS I

REACTOR PROTECT Deleted Trip SettingE And Allowable Trip Function Deviations

9.

ain*asea in radiation (7) background at (RM-17-251(A-D) rated power(8)

ION TABLE 3.1.1 (COnt 'd)

SYSTEM (SCRAM)

INSTRUMENT REQUIREMENTS modes in Refuel (1)

Required ACTIONS When Minimum Number Minimum Which Functions Must Operating Conditions For be Operating Instrument Operation Channels Per Are Not Startup Run Trip System (2)

Satisfied (3)

A.

A or_.

10. Main steamline isolation valve closure (POS-2-80A-AB1 POS-2-86A-AI,Bl POS-2-80B-Al,B2 POS-2-86B-Al,B2 POS-2-80C-A2,Bl POS-2-86C-A2,B1 POS-2 -80D-A2,B2 POS-2-86D-A2,B2)

11.

Turbine control valve fast closure (PS- (37-40))

12.

Turbine stop valve closure (SVOS-S- (1-4))

-clot valve "closure (9)(10) 410% valve (10) closure Amendment No.

a4-+, 186 x

4 A or C x

X 2

2 A or D A or D 22 I

5

VYNPS TABLE 3.1.1 NOTES (Cont'd)

3.

When the requirements in the column NMinimum Number of Operating Instrument Channels Per Trip System" cannot be met for one system, that system shall be tripped.

If the requirements cannot be met for both trip systems, the appropriate ACTIONS listed below shall be taken:

a)

Initiate insertion of operable rods and complete insertion of all operable rods within four hours.

b)

Reduce power level to IRM range and place mode switch in the "Startup/Hot Standby' position within eight hours.

c) Reduce turbine load and close main steam line isolation valves within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

d)

Reduce reactor power to less than 30k of rated within a hours.

4.

IWO is percent rated two loop drive flow where 100& rated drive flow is that flow equivalent to 48 x 101 lbs/hr core flow.

AW is the difference between the two loop and single loop drive flow at the same core flow.

This difference must be accounted for during single loop operation.

AW -

0 for two recirculation loop operation.

5.

To be considered operable an APRM must have at least 2 LPRM inputs per level and at least a total of 13 LPRM inputs, except that channels A, C, D, and F may lose all LPRM inputs from the companion APRM Cabinet plus one additional LPRM input and still be considered operable.

6.

The top of the enriched fuel has been designated as 0 inches and provides common reference level for all vessel water level instrumentation.

fhnnel h

a'Systems.

_,_j**channel signals for the turbine control valve fast closure trip shall be De=*leted erived from the same event or events which cause the control valve fast

10.

Turbine stop valve closure and turbine control valve fast closure scram signals may be bypassed at c30% of reactor Rated Thermal Power.

IIri

11. Not used.

12.

While performing refuel interlok checks which require the mode switch to be in Startup, the reduced APR14 high flux scram need not be operable provided:

a.

The following trip functions are operable:

i.

Mode switch in shutdown,

2.

Manual scram,

3.

High flux IRK scram

4.

High flux v

Rl scram in nonconncudence,

5.

Scram discharge volume high water level, and1

b.

No more than two (2) control rods withdrawn.

The two (2) control rods that can be withdrawn cannot be face adjacent or diagonally adjacent.

Amendment No. 14, 2, OZ, "4, 69, 4g,

, 04, 14+, 1-7-,

4;&, 196 24

VYNPS TABLE 4.1.1 SCRAM INSTRUMENTATION AND LOGIC SYSTEMS FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENTATION, LOGIC SYSTEMS AND CONTROL CIRCUITS Instrument Channel Mode Switch in Shutdown Manual Scram IRM High Flux Inoperative Group(3)

A Functional Test(7)

Place Mode Switch in Shutdown A

Trip Channel and Alarm C

Trip Channel and Alarm(s)

C Trip Channel and Alarm APRM High Flux High Flux (Reduced)

Inoperative Flow Bias B

Trip Output Relays( 5)

B Trip Output Relays( )

B B

Trip Output Relays Trip Output Relaysis Minimum Prequency(4)

Each Refueling Outage Every 3 Months Before Each Startup & Weekly During Refueling(6)

Before Each Startup & Weekly During Refueling1 ')

Every 3 Months Before Each Startup & Weekly During Refueling( 6)

Every 3 Months Every 3 Months High Reactor Pressure High Drywell Pressure Low Reactor Water Level(2 1(8 )

High Water Level in Scram Discharge Volume igýh ain StAýmLine Radiation(2)

Main Steam Line Iso. Valve Closure Turbine Con. Valve Fast Closure Turbine Stop Valve Closure Scram Test Switch (SA-S2(A-D))

First Stage Turbine Pressure Permissive (PS-5-14(A-D))

Amendment No. 44, 24r,

,G,

G, +6",

186 B

Trip B

Trip Zi A

A A

Trip Trip Trip Trip Trip Trip Channel and Alarm(5 )

Every 3 Months Channel and Alarm(s)

Every 3 Months Channel and larm Every 3 Months Channel and Alarm Every 3 Months Channel and Alarm Every 3 Months Channel and Alarm Every 3 Months Channel and Alarm Once each week (9)

Channel and Alarm Every 6 Months 25

[E-

VYNPS TABLE 4.1.1 NOTES S1.

Not used

2.

An instrument check shall be performed on reactor water level and reactor pressure instrumentation once per dalsnd streamline ra t on monitor

3.

A description of the three groups is included in the basis of this Specification.

4.

Functional tests are not required when the systems are not required to be operable or are tripped.

If tests are missed, they shall be performed prior to returning the systems to an operable status.

5.

This instrumentation is exempted from the Instrument Functional Test Definition (1.G.).

This Instrument Functional Test will consist of injecting a simulated electrical signal into the measurement channels.

6.

Frequency need not exceed weekly.

7.

A functional test of the logic of each channel is performed as indicated.

This coupled with placing the mode switch in shutdown e~ach refueling outage constitutes a logic system functional test of the scram system.

8.

The water level in the reactor vessel will be perturbed and the corresponding level indicator changes will be monitored.

This test will be performed every month.

9.

The automatic scram contactors shall be exercised once every week by either using the RPS channel test switches or performing a functional test of any automatic scram function.

If the contactors are exercised using a functional test of a scram function, the weekly test using the RPS channel test switch is considered satisfied.

The automatic scram contactors shall also be exercised after maintenance on the contactors.

26 Amendment No. 0&, SO, 186 I I I

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VYNPS TABLE 4.1.2 SCRAM INSTRUMENT CALIBRATION MINIMUM CALIBRATION PRROt1RN(TIR Pfl

AtWfl

flrWTflN Y

IPT1MrWI' r'uatna2ta Instrument Channel High Flux APRM Output Signal Output Signal (Reduced) (7)

Flow Bias LPRM (LPRM ND-2-1-104 (80))

High Reactor Pressure Turbine Control Valve Fast Closure High Drywell Pressure High Water Level in Scram Discharge Volume Low Reactor Water Level Turbine Stop Valve Closure 1-3 gh Main St Line Radiation First Stage Turbine Pressure Permissive (PS-5-14(A-D))

Main Steam Line Isolation Valve Closure 1rou3 (5)

B B

B siS)

B A

B B

Calibration Standard(4)

Heat Balance Heat Balance Standard Pressure Source Using TIP System Standard Pressure Standard Pressure Standard Pressure Water Level B

Standard Pressure A

(6)

B Appropriate Rad Source(3)

A Pressure Source A

(6) and Voltage Source Source Source Minimum Frequency(2)

Once Every? Days Once Every 7 Days Refueling Outage Every 2,000 MWD/T average core exposure (8)

Once/Operating Cycle Every 3 Months Once/Operating Cycle Once/Operating Cycle Source Once/Operating Cycle Refueling Outage

=ion Refuelingt Every 6 Months and After Refueling Refueling Outage Amendment No.

-4, Q4, aa, &S, 64-,

46, 6", 186, 191 27 MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INQqvT"AvuT ew'AmwvT-0 I

VYNPS TABLE 4.1.2 NOTES

1.

A description of the three groups is included in the bases of this Specification.

2.

Calibration tests are not required when the systems are not required to be operable or are tripped.

If tests are missed, they shall be performed prior to returning the systems to an operable status.

3.

Acurrent suc ovd an ins rumt Channel alignm*z every 3 months.*

4.

Rsponse time is not part of the routine instrument chec'"k and calibration, but will be checked every operating cycle.

6.

Physical inspection and actuation.

7.

The IRM and SRM channels shall be determined to overlap during each startuý after entering the STARTUP/HOT STANDBY MODE and the IRM and APRM channels shall be determined to overlap during each controlled shutdown, if not performed within the previous 7 days.

8.

The specified frequency is met if the calibration is performed within 1.25 times the interval specified, as measured from the previous performance.

Amendment No.

", *6, 191 P

28

SVYKPS8 BASES:

3.1 (Cont'd) nrmal nitrogen nd oxygen radioactivity a an indication of *aking fuel.

A scram is in ated whenever such red tion level exceeds t ee times normal back und.

The purpose of th scram is to reduce he source of such radia on to the extent necess to prevent release f radioactive material to the turbine.

An ala is initiated wheneve the radiation level eeds 1.5 times normal b ground to alert the perator to possible seri a radioactivity spikes du to abnormal core beh ior.

The Augmented offA as (AOG) monitors provided urther assurance aga est the release of rioactive materialste to the te environs by isol aing the AOG stacku

/

The main steam line isolation valve closure scram is set to scram when the isolation valves are 10 percent closed from full open in 3-out-of-4 lines.

This scram anticipates the pressure and flux transient, which would occur when the valves close.

By scramming at this setting, the resultant transient is insignificant.

A reactor mode switch Is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.

The manual scram function in active in all modes, thus providing for manual means of rapidly inserting control rods during all modes of reactor operation.

The IRM system provides protection against short reactor periods and, in conjunction with the reduced APEM system provides protection against excessive power levels in the startup and intermediate power ranges.

A source range monitor (SRM) system is also provided to supply additional neutron level information during startup and can provide scram function with selected shorting links removed during refueling.

Thus, the IRM and the reduced APRM are normally required in the startup mode and may be required in the refuel mode.

During some refueling activities which require the mode switch In startupi it is allowable to disconnect the LPRMs to protect them from damage during under vessel work.

In lieu of the protection provided by the reduced APRM scram, both the IRM scram and the SRM scram in noncoincidence are used to provide neutron monitoring protection against excessive power levels.

In the power range, the normal APRM system provides required protection.

Thus, the IRM system and 15%

APRM scram are not required in the run mode.

If an unsafe failure is detected during surveillance testing, it is desirable to determine as soon as possible if other failures of a similar type have occurred and whether the particular function involved is still operable or capable of meeting the single failure criteria.

To meet the requirements of Table 3.1.1, it Is necessary that all instrument channels in one trip system be operable to permit testing in the other trip system.

Thus, when failures are detected in the first trip system tested, they would have to be repaired before testing of the other system could begin.

In the majority of cases, repairs or replacement can be accomplished quickly.

If repair or replacement cannot be completed in a reasonable time, operation could continue with one tripped system until the surveillance testing deadline.

Amendment No.

Q4, 40, -&", M 01-S2 31

VYNPS

1. Whenever the main steam line isolation valves are open, the mechanical vacuum pump shall be capable of being automatically isolated and secured by a signal of high radiation in the main steam line tunnel or shall be manually isolated and secured.

2.

If Specification 3.2.F.1 is not met following a routine surveillance check, the reactor shall be in the cold shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

G.

Post-AccidentInstrumentation During reactor power operation, the instrumentation that displays information in the Control Room necessary for the operator to initiate and control the systems used during and following a postulated accident or abnormal operating condition shall be operable in accordance with Table 3.2.6.

During each operating cycle, automatic isolation and securing of the mechanical vacuum pump shall be verified while the reactor is shutdown.

G.

Post-AccidentInstrumentation The post-accident instrumentation shall be functionally tested and calibrated in accordance with Table 4.2.6.

Amendment No.

9, 164 35 I

I 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION I

D.

Off-Gas System Isolation io__n During reactor power

1. The High Main Steam Line c

operation, the instrumentation that Radiation trip function of the initiates isolation of mechanical vacuum pump off-gas system shall bt shall be functionally tested operable in accordance and calibrated in accordance with Table 3.2.4.

with Table 4.2.2.

E.

Control Rod Block Actuat.

on During reactor power jgic operation the instrumentatioa-ictionally that initiates control rod ated as block shall be operable Table 4.2.5.

in accordance with Table 3.2.5.

F.

Mechanical Vacuum Pump F. Mechanical Vacuum Pump Isolation Isolation I

I I

VYNPS TABLE 3.2.2 PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION Minimum Number of Operable Instrument Channels per Trip System Trip Function Trip Setting Required ACTION When Minimum Conditions For Operation Are Not Satisfied (Note 2) 2 (Notes 11,12) 2 of 4 in each of 2 channels (Notes 11,12) 2/steam line (Notes 11,12) 2 (Notes 1,11,12) 2 (Notes 6,11,12) 2 (Notes 11,12) 2 (Notes 11,12) 2 (Notes 11,12) 2 (Notes 10,11,12) 1 Low-Low Reactor Vessel Water Level (LT-2-3-57A/B(S2),

LT-2-3-58A/B(S2))

High Main Steam Line Area Temperature (TS-2-(121-124)(A-D))

High Main Steam Line Flow (DPT-2-(116-119)(A-D) (M))

Low Main Steam Line Pressure (PS-2-134(A-D))

High Main Steam Line Flow (DPT-2-116A, 117B, 118C,119D(S1))

Low Reactor Vessel Water Level (LT-2-3-5VA/B(M),

LT-2-3-58A/B(M))

High Main Steam Line Radiation (7)

(8)

(RM-17-251(A-D))

High Drywell Pressure Condenser Low Vacuum

>82.5" above the top of enriched fuel

<2120F

<140% of rated flow

>800 psig

<40% of rated flow Same as Reactor Protection System

<3 x background at rated power (9)

Same as Reactor Protection System

<12" Hg absolute Trip System Logic Amendment No.

94, 444, 186 A

B B

A A

A I

45

VYNPS TABLE 3.2.2 NOTES

1.

The main steam line low pressure need be available only in the "Run" mode.

2.

If the minimum number of operable instrument channels are not available for one trip system, that trip system shall be tripped.

If the minimum number of operable instrument channels are not available for both trip systems, the appropriate actions listed below shall be taken:

A.

Initiate an orderly shutdown and have reactor in the cold shutdown condition in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B.

Initiate an orderly load reduction and have reactor in "Hot Standby" within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

3.

Close isolation valves in systqm and comply with Specification 3.5.

C. The actions required by Specification logic the "Refuel,"

.7.

This s.gi...-

.ue mechanical vacuum pump su4 5

i line isolation valves.

ction

8.

Channel shar by the Reactor Prote on and Primar C ainment Isola S9.

An alarm setting of 1.5 times normal background at rated power shall be S~established to alert the operator to abnormal radiation levels in the Sprimary coolant.

10.

A key lock the bypass ofo.this trip function to

v.

acuum is greater than

-'-"nass valves are This ~ ~ ~

~

c trpfntinIripial t

h ecaia vacuum pump; however, It Is not applicable to the lve, is steam Isolation, main steam line drain, and recirculation Aons loop sample line isolation valves).

_C edbility.

12.

Whenevei by Specification 3.7.A.2, there shall be two operable or tripped trip systems for each Trip 6

Function, except as provided for below:

A. With one or more automatic functions with isolation capability not maintained restore isolation capability in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or take the ACTION required by Table 3.2.2.

B. With one or more channels inoperable, place the inoperable channels (s) in the tripped condition within:

1) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functions common to RPS instrumentation, and

2) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for trip functions not common to RPS instrumentation, or, initiate the ACTION required by Table 3.2.2.

Amendment No. -,

1--,

186 48

.11, 1 1 1

VYNPS TABLE 4.2.2 MIN1MUM TEST AND CALIBRATION FREQUECTRS PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION Trip Function Low-Low Reactor Vessel Water Level High Steam Line Area Temperature High Steam Line Flow Low Main Steam Line Pressure Low Reactor Vessel Water Level High Main Steam Line Radiation High Drywell Pressure Condenser Low Vacuum Trip System Logic Functional Test(8)

Every Three Months Every Three Months Every Three Months Every Three Months Every Three Months Every Three Months (Note 7)

Every Three Months Every Three Months Once/Operating Cycle (Note 2)

Calibration(S)

Once/Operating Cycle Each Refueling Outage Once/Operating Cycle Every Three Months Once/Operating Cycle Each Refueling Outage Once/Operating Cycle Every Three Months Once/Operating Cycle (Note 3)

Instrument Check Once Each Day Once Each Day Once Once Each Day Each Day Amendment No. 4, 60,

,6, +6, i*-3, 186 64

VYNPS TABLE 4.2 NOTES

1.

Not used.

2.

During each refueling outage, simulated automatic actuation which opens all pilot valves shall be performed such that each trip system logic can be verified independent of its redundant counterpart.

3.

Trip system logic calibration shall include only time delay relays and timers necessary for proper functioning of the trip system.

4.

This instrumentation is excepted from functional test definition.

The functional test will consist of injecting a simulated electrical signal into the measurement channel.

5.

Deleted.

6.

Functional tests, calibrations, and instrument checks are not required when these instruments are not required to be operable or are tripped.

Functional tests shall be performed before each startup with a required frequency not to exceed once per week.

Calibration shall be performed prior to or during each startup or controlled shutdown with a required frequency not to exceed once per week.

Instrument checks shall be performed at least once per day during those periods when instruments are required to be operable.

7.

This instrumentation is excepted from the functional test definitions and shall be calibrated using simulated electrical signals once every three months.

8.

Functional tests and calibrations are not required when systems are not required to be operable.

9.

The thermocouples associated with safety/relief valves and safety valve position, that may be used for back-up position indication, shall be verified to be operable every operating cycle.

10.

Separate functional tests are not required for this instrumentation.

The calibration and integrated ECCS tests which are performed once per operating cycle will adequately demonstrate proper equipment operation.

11.

Trip system logic functional tests will include verification of operation of all automatic initiation inhibit switches by monitoring relay contact movement.

Verification that the manual inhibit switches prevent opening all relief valves will be accomplished in conjunction with Section 4.5.F.I.

i4 main steam Isolation, main steam line drain, and Amendment No. 64', 0&~, ~4",

446, &4", 186 74

8 High radiation mo tors in the main stea ine tunnel have bee provided to detect gross fu failure resulting fro a control rod drop a ident.

This instrumentatio causes closure of Gro 1 valves, the only Ives required to close for is accident.

With th established setting f 3 times normal background d main steam line iso tion valve closure, ission product release i limited so that 10CFR1 limits are not exc eded for the control rod dro accident.

With an ala setting of 1.5 tims normal background, the o rator is alerted to pos ible gross fuel fail e or abnormal fission prod t releases from failed uel due to transienyreactor operation.

Pressure instrumentation is provided which trips when main steam line pressure drops below 800 psig.

A trip of this instrumentation results in closure of Group 1 isolation valves.

In the refuel, shutdown, and startup modes, this trip function is provided when main steam line flow exceeds 40%

of rated capacity.

This function is provided primarily to provide protection against a pressure regulator malfunction which would cause the Amendment No. 46, 6 t4, 46, 64, EVY 01-52

ion line and Main steam line radiation monitors are located In the main meest he steam tunnel downstream from the MSIVs and provide an early indication of gross fuel failures. This instrumentation Initiates alarms and tripsfisolations assumed In the analysis of the vel design basis control rod drop accident (CRDA). Based on

-)n of NEDO-31400A, "Safety Evaluation for Eliminating the Boiling

.eak Water Reactor Main Steam Line Isolation Valve Closure out Function and Scram Function of the Main Steam Line Radiation I Its Monitor," and plant-speclfc analysis of the CRDA, this M.

instrumentation Is no longer required to automatically cause I.

closure of the MSIVs and other Group I Isolation valves to lot ensure compliance with the radiation dose limits of 10CFR1OO.

The mechanical vacuum pump tripfisolation setting of< 3 times normal background at rated thermal power Is as low as

-y practicable without consideration of spurious trips from nitrogen-16 spikes, instrument Instabilities and other operational occurrences. With an alarm setting of a nominal stea 1.5 times normal background at rated thermal power, the ing control room operator Is alerted to off-normal conditions and to in valves determine the need for corrective actions in accordance with

/n the main Splant procedures.

case Sj setting of d limiters and Ss such that fuel

.,an 12950F and release of radioau

_R10 Temperature monitoriziu

-on is provided in the main steam line tunnel to detect leaks in this area.

Trips are provided on this instrumentation and when exceeded cause closure of Group 1 isolation valves.

Its setting of ambient plus 95OF is low enough to detect leaks of the order of 5 to 10 gpm; thus, it is capable of covering the entire spectrum of breaks.

For large breaks, it is a backup to high steam flow instrumentation discussed above, and for small breaks, with the resultant small release of radioactivity, gives isolation before the limits of 10CFR10o are exceeded.

76

VYNPS TABLE 4.7.2 NOTES

1.

Isolation si ;nals are as follows:

Group 1:

The valves in Group 1 are closed upon any one of the following conditions:

1.

Low-low reactor water level Deleted

2.,

I

3.

ig main steam 1ne tlow

4.

High main steam line tunnel temperature

5.

Low main steam line pressure (run mode only)

6.

Condenser low vacuum Group 2:

The valves in Group 2 are closed upon any one of the following conditions*:

1.

Low reactor water level

2.

High drywell pressure Group 3:

The valves in Group 3 are closed upon any one of the following conditions:

1.

Low reactor water level

2.

High drywell pressure

3.

High/low radiation - reactor building ventilation exhaust plenum or refueling floor Group 4:

The valves in Group 4 are closed upon any one of the following conditions:

1.

Low reactor water level

2.

High drywell pressure

3.

High reactor pressure Group 5:

The valves in Group 5 are closed upon low reactor water level.

Group 6:

The valves in Group 6 are closed upon any signal representing a steam line break in the HPCI system's or RCIC system's respective steam line.

The signals indicating a steam line break for the respective steam line are as follows:

1.

2.

3.

4.

High steam line space temperature High steam line flow Low steam line pressure High temperature in the main steam line tunnel (30 minute delay for the HPCI and the RCIC)

2.

The closure time shall not be less than 3 seconds.

Valves V10-39A/B, V10-34A/B, Vl0-26A/B, V10-31A/B and V10-38A/B are closed upon either 1) low-low reactor water level and low reactor pressure or 2) high drywell pressure.

Amendment No. #, 194 I

162 I I

11, 1 1

1 1

Docket No. 50-271 BVY 02-18 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 250 Scram and Isolation Valve Closure Functions of the Main Steam Line Radiation Monitors Retyped Technical Specification Pages I III :

I I

BVY 02-18 / Attachment 4 / Page I Listing of Affected Technical Specifications Pages Replace the Vermont Yankee Nuclear Power Station Technical Specifications pages listed below with the revised pages included herein. The revised pages contain vertical lines in the margin indicating the areas of change.

Remove Insert 22 22 24 24 25 25 26 26 27 27 28 28 31 31 35 35 45 45 48 48 48a 64 64 74 74 76 76 162 162 I

I

VYNPS TABLE 3.1.1 (Cont'd)

REACTOR PROTECTION SYSTEM (SCRAM)

INSTRUMENT REQUIREMENTS Trip Function Trip Settings And Allowable Deviations Modes in Which Functions M be Operating Refuel (1)

Startup Minimum Number lust Operating Instrument Channels Per Run Trip System (2)

Required ACTIONS When Minimum Conditions For Operation Are Not Satisfied (3)

9.

Deleted

10.

Main steamline isolation valve closure (POS-2-80A-A1,B1 POS-2-86A-Al,BI POS-2-80B-Al,B2 POS-2-86B-Al,B2 POS-2-80C-A2,Bl POS-2-86C-A2,B1 POS-2-80D-A2,B2 POS-2-86D-A2,B2)

11.

Turbine control valve fast closure (PS- (37-40))

12.

Turbine stop valve closure (SVOS (1-4))

<10% valve closure (9) (10)

<10% valve (10) closure Amendment No. +", a62 x

x x

4 2

2 A or C A or D A or D 22

VYNPS TABLE 3.1.1 NOTES (Cont'd)

3.

When the requirements in the column "Minimum Number of Operating Instrument Channels Per Trip System" cannot be met for one system, that system shall be tripped.

If the requirements cannot be met for both trip systems, the appropriate ACTIONS listed below shall be taken:

a)

Initiate insertion of operable rods and complete insertion of all operable rods within four hours.

b)

Reduce power level to IRM range and place mode switch in the "Startup/Hot Standby" position within eight hours.

c)

Reduce turbine load and close main steam line isolation valves within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

d)

Reduce reactor power to less than 30% of rated within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

4.

"W" is percent rated two loop drive flow where 100% rated drive flow is that flow equivalent to 48 x 106 lbs/hr core flow.

AW is the difference between the two loop and single loop drive flow at the same core flow.

This difference must be accounted for during single loop operation.

AW = 0 for two recirculation loop operation.

5.

To be considered operable an APRM must have at least 2 LPRM inputs per level and at least a total of 13 LPRM inputs, except that channels A, C, D, and F may lose all LPRM inputs from the companion APRM Cabinet plus one additional LPRM input and still be considered operable.

6.

The top of the enriched fuel has been designated as 0 inches and provides common reference level for all vessel water level instrumentation.

7.

Deleted.

8.

Deleted.

9.

Channel signals for the turbine control valve fast closure trip shall be derived from the same event or events which cause the control valve fast closure.

10. Turbine stop valve closure and turbine control valve fast closure scram signals may be bypassed at <30% of reactor Rated Thermal Power.

11.

Not used.

12.

While performing refuel interlock checks which require the mode switch to be in Startup, the reduced APRM high flux scram need not be operable provided:

a.

The following trip functions are operable:

1.

Mode switch in shutdown,

2.

Manual scram,

3.

High flux IRM scram

4.

High flux SRM scram in noncoincidence,

5.

Scram discharge volume high water level, and;

b.

No more than two (2) control rods withdrawn.

The two (2) control rods that can be withdrawn cannot be face adjacent or diagonally adjacent.

Amendment No.

-*, 2-*, 4, 64., 6, 9,

94, 4-",

a-!74, a4, 6 24 I

I

VYNPS TABLE 4.1.1 SCRAM INSTRUMENTATION AND LOGIC SYSTEMS FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENTATION, LOGIC SYSTEMS AND CONTROL CIRCUITS Instrument Channel Group (3 Functional Test(7)

Minimum Frequency(4)

Mode Switch in Shutdown A

Place Mode Switch in Shutdown Each Refueling Outage Trip Channel and Alarm Trip Channel and Alarm(s)

Trip Channel and Alarm APRM High Flux High Flux (Reduced)

Inoperative Flow Bias High Reactor Pressure High Drywell Pressure Low Reactor Water Level(2)(8)

High Water Level in Scram Discharge Volume Main Steam Line Iso. Valve Closure Turbine Con. Valve Fast Closure Turbine Stop Valve Closure Scram Test Switch (SA-S2(A-D))

First Stage Turbine Pressure Permissive (PS-5-14(A-D))

B B

A A

A Trip Output Relays(5)

Trip Output Relays(5 )

Trip Output Relays Trip Output Relays(s)

Trip Trip Trip Trip Trip Trip Trip Trip Trip Channel Channel Channel Channel Channel Channel Channel Channel Channel and and and and and and and and and Alarm(5)

Alarm(5)

Alarm(s)

Alarm Alarm Alarm Alarm Alarm Every 3 Months Before Each Startup & Weekly During Refueling( 6)

Before Each Startup & Weekly During Refueling( 6)

Every 3 Months Before Each Startup & Weekly During Refueling(6)

Every 3 Months Every 3 Months Every Every Every Every 3

3 3

3 Months Months Months Months Every 3 Months Every 3 Months Every 3 Months Once each week (9)

Every 6 Months Amendment No.

44, Q+,

&&, -7, 6, +,

a42 Manual Scram IRM High Flux Inoperative A

C C

I 25

VYNPS I

Amendment No.

.1-,

&&, 142 TABLE 4.1.1 NOTES

1.

Not used

2.

An instrument check shall be performed on reactor water level and reactor pressure instrumentation once per day.

3.

A description of the three groups is included in the basis of this Specification.

4.

Functional tests are not required when the systems are not required to be operable or are tripped.

If tests are missed, they shall be performed prior to returning the systems to an operable status.

5.

This instrumentation is exempted from the Instrument Functional Test Definition (1.G.).

This Instrument Functional Test will consist of injecting a simulated electrical signal into the measurement channels.

6.

Frequency need not exceed weekly.

7.

A functional test of the logic of each channel is performed as indicated.

This coupled with placing the mode switch in shutdown each refueling outage constitutes a logic system functional test of the scram system.

8.

The water level in the reactor vessel will be perturbed and the corresponding level indicator changes will be monitored.

This test will be performed every month.

9.

The automatic scram contactors shall be exercised once every week by either using the RPS channel test switches or performing a functional test of any automatic scram function.

If the contactors are exercised using a functional test of a scram function, the weekly test using the RPS channel test switch is considered satisfied.

The automatic scram contactors shall also be exercised after maintenance on the contactors.

26

VYNPS TABLE 4.1.2 SCRAM INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREOUJENCIES FOR REACTOR PROTECT!ON !MSTR!NRNT C*HANNRL.S Instrument Channel High Flux APRM Output Signal Output Signal (Reduced)

(7)

Flow Bias LPRM (LPRM ND-2-I-104(80))

High Reactor Pressure Turbine Control Valve Fast Closure High Drywell Pressure High Water Level in Scram Discharge Volume Low Reactor Water Level Turbine Stop Valve Closure First Stage Turbine Pressure Permissive (PS-5-14(A-D))

Main Steam Line Isolation Valve Closure Group(1)

B B

B B(5)

B A

B B

B A

A A

Calibration Standard(4)

Heat Balance Heat Balance Standard Pressure Source Using TIP System Standard Pressure Standard Pressure Standard Pressure Water Level and Voltage Source Source Source Standard Pressure Source (6)

Pressure Source (6)

Minimum Frequency(2)

Once Every 7 Days Once Every 7 Days Refueling Outage Every 2,000 MWD/T average core exposure (8)

Once/Operating Cycle Every 3 Months Once/Operating Cycle Once/Operating Cycle Once/Operating Cycle Refueling Outage Every 6 Months and After Refueling Refueling Outage Amendment No.

Q-1, 4,

46, 4.64, 4-",

4-27 MIIU CALIRATIN..RQUEN.ES.OR.RAC..

PRTCTO INSTRUMENT...

CHNNL I

27

VYNPS TABLE 4.1.2 NOTES

1.

A description of the three groups is included in the bases of this Specification.

2.

Calibration tests are not required when the systems are not required to be operable or are tripped.

If tests are missed, they shall be performed prior to returning the systems to an operable status.

3.

Deleted.

4.

Response time is not part of the routine instrument check and calibration, but will be checked every operating cycle.

5.

Does not provide scram function.

6.

Physical inspection and actuation.

7.

The IRM and SRM channels shall be determined to overlap during each startup after entering the STARTUP/HOT STANDBY MODE and the IRM and APRM channels shall be determined to overlap during each controlled shutdown, if not performed within the previous 7 days.

8.

The specified frequency is met if the calibration is performed within 1.25 times the interval specified, as measured from the previous performance.

Amendment No. a-2, 486, 4412 28

VYNPS BASES:

3.1 (Cont'd)

The main steam line isolation valve closure scram is set to scram when the isolation valves are 10 percent closed from full open in 3-out-of-4 lines.

This scram anticipates the pressure and flux transient, which would occur when the valves close.

By scramming at this setting, the resultant transient is insignificant.

The main steam line isolation valve closure scram is set to scram when the isolation valves are 10 percent closed from full open in 3-out-of-4 lines.

This scram anticipates the pressure and flux transient, which would occur when the valves close.

By scramming at this setting, the resultant transient is insignificant.

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.

The manual scram function is active in all modes, thus providing for manual means of rapidly inserting control rods during all modes of reactor operation.

The IRM system provides protection against short reactor periods and, in conjunction with the reduced APRM system provides protection against excessive power levels in the startup and intermediate power ranges.

A source range monitor (SRM) system is also provided to supply additional neutron level information during startup and can provide scram function with selected shorting links removed during refueling.

Thus, the IRM and the reduced APRM are normally required in the startup mode and may be required in the refuel mode.

During some refueling activities which require the mode switch in startup; it is allowable to disconnect the LPRMs to protect them from damage during under vessel work.

In lieu of the protection provided by the reduced APRM scram, both the IRM scram and the SRM scram in noncoincidence are used to provide neutron monitoring protection against excessive power levels.

In the power range, the normal APRM system provides required protection.

Thus, the IRM system and 15%

APRM scram are not required in the run mode.

If an unsafe failure is detected during surveillance testing, it is desirable to determine as soon as possible if other failures of a similar type have occurred and whether the particular function involved is still operable or capable of meeting the single failure criteria.

To meet the requirements of Table 3.1.1, it is necessary that all instrument channels in one trip system be operable to permit testing in the other trip system.

Thus, when failures are detected in the first trip system tested, they would have to be repaired before testing of the other system could begin.

In the majority of cases, repairs or replacement can be accomplished quickly.

If repair or replacement cannot be completed in a reasonable time, operation could continue with one tripped system until the surveillance testing deadline.

Amendment No. a-2,

4, a-,

BW 062 31

VYNPS 3.2 LIMITING CONDITIONS FOR OPERATION D.

Off-Gas System Isolation During reactor power operation, the instrumentation that initiates isolation of the off-gas system shall be operable in accordance with Table 3.2.4.

E.

Control Rod Block Actuation During reactor power operation the instrumentation that initiates control rod block shall be operable in accordance with Table 3.2.5.

F.

Mechanical Vacuum Pump Isolation

1.

Whenever the main steam line isolation valves are open, the mechanical vacuum pump shall be capable of being automatically isolated and secured by a signal of high radiation in the main steam line tunnel or shall be manually isolated and secured.

2.

If Specification 3.2.F.1 is not met following a routine surveillance check, the reactor shall be in the cold shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

G.

Post-AccidentInstrumentation During reactor power operation, the instrumentation that displays information in the Control Room necessary for the operator to initiate and control the systems used during and following a postulated accident or abnormal operating condition shall be operable in accordance with Table 3.2.6.

4.2 SURVEILLANCE REQUIREMENTS D.

Off-Gas System Isolation Instrumentation and logic systems shall be functionally tested and calibrated as indicated in Table 4.2.4.

E.

Control Rod Block Actuation Instrumentation and logic systems shall be functionally tested and calibrated as indicated in Table 4.2.5.

F.

Mechanical Vacuum Pump Isolation

1.

The High Main Steam Line Radiation trip function of the mechanical vacuum pump shall be functionally tested and calibrated in accordance with Table 4.2.2.

2.

During each operating cycle, automatic isolation and securing of the mechanical vacuum pump shall be verified while the reactor is shutdown.

G.

Post-AccidentInstrumentation The post-accident instrumentation shall be functionally tested and calibrated in accordance with Table 4.2.6.

Amendment No. 4, 464 35

VYNPS TABLE 3.2.2 PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION Minimum Number of Operable Instrument Channels per Trip System Trip Function Trip Setting Required ACTION When Minimum Conditions For Operation Are Not Satisfied (Note 2) 2 (Notes 11,12) 2 of 4 in each of 2 channels (Notes 11,12) 2/steam line (Notes 11,12) 2 (Notes 1,11,12) 2 (Notes 6,11,12) 2 (Notes 11,12) 2 (Notes 11,12) 2 (Notes 11,12) 2 (Notes 10,11,12) 1 Low-Low Reactor Vessel Water Level (LT-2-3-57A/B(S2),

LT-2-3-58A/B(S2))

High Main Steam Line Area Temperature (TS-2-(121-124) (A-D))

High Main Steam Line Flow (DPT (116-119) (A-D) (M))

Low Main Steam Line Pressure (PS-2-134(A-D))

High Main Steam Line Flow (DPT-2-116A, 117B, 118C,119D(S1))

Low Reactor Vessel Water Level (LT-2-3-57A/B(M),

LT-2-3-S8A/B(M))

High Main Steam Line Radiation (7)

(8)

(RM-17-251(A-D))

High Drywell Pressure Condenser Low Vacuum

>82.5" above the top of enriched fuel

<2120 F

<140% of rated flow

>800 psig

<40% of rated flow Same as Reactor Protection System

<3 x background at rated power (9)

Same as Reactor Protection System

<12" Hg absolute Trip System Logic Amendment No.

9, &&,

64, 94, 0a4, -164 A

B B

A B or C A

A A

45 I

VYNPS TABLE 3.2.2 NOTES

1.

The main steam line low pressure need be available only in the "Run" mode.

2.

If the minimum number of operable instrument channels are not available for one trip system, that trip system shall be tripped.

If the minimum number of operable instrument channels are not available for both trip systems, the appropriate actions listed below shall be taken:

A.

Initiate an orderly shutdown and have reactor in the cold shutdown condition in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B.

Initiate an orderly load reduction and have reactor in "Hot Standby" within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

C.

The actions required by Specification 3.2.F.1 shall be taken immediately.

3.

Close isolation valves in system and comply with Specification 3.5.

4.

Deleted.

5.

One trip system arranged in a one-out-of-two twice logic.

6.

The main steam line high flow is available only in the "Refuel,"

"Shutdown," and "Startup" modes.

7.

This signal also automatically closes the mechanical vacuum pump suction line isolation valves.

8.

This trip function is applicable to the mechanical vacuum pump; however, it is not applicable to the primary containment Group 1 isolation valves (i.e., main steam isolation, main steam line drain, and recirculation loop sample line isolation valves).

9.

An alarm setting of 1.5 times normal background at rated power shall be established to alert the operator to abnormal radiation levels in the primary coolant.

10.

A key lock switch is provided to permit the bypass of this trip function to enable plant startup and shutdown when the condenser vacuum is greater than 12 inches Hg absolute provided that both turbine stop and bypass valves are closed.

11.

When a channel, and/or the affected primary containment isolation valve, is placed in an inoperable status solely for performance of required instrumentation surveillances, entry into associated Limiting Conditions for Operation and required ACTIONS may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Trip Function maintains isolation capability.

12.

Whenever Primary Containment integrity is required by Specification 3.7.A.2, there shall be two operable or tripped trip systems for each Trip Function, except as provided for below:

A.

With one or more automatic functions with isolation capability not maintained restore isolation capability in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or take the ACTION required by Table 3.2.2.

Amendment No. 9, 649,

-4, a,6 48

VYNPS TABLE 3.2.2 NOTES (Cont'd)

B.

With one or more channels inoperable, place the inoperable channels (s) in the tripped condition within:

1) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functions common to RPS instrumentation, and

2) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for trip functions not common to RPS instrumentation, or, initiate the ACTION required by Table 3.2.2.

13.

Whenever the High Pressure Cooling Injection System and Reactor Core Isolation Cooling System are required to be operable in accordance with Specification 3.5, the low steam supply pressure automatic isolation trip system shall be operable, except as provided below:

A.

With the automatic isolation trip function not maintained, restore isolation capability in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or take the ACTION required by Table 3.2.2.

B.

With one or more required channels inoperable, place the inoperable channel(s) in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or take the ACTION required by Table 3.2.2.

Amendment No. 28G 48a

VYNPS Trip Function Low-Low Reactor Vessel Water Level High Steam Line Area Temperature High Steam Line Flow Low Main Steam Line Pressure Low Reactor Vessel Water Level High Main Steam Line Radiation (Note 12)

High Drywell Pressure Condenser Low Vacuum Trip System Logic TABLE 4.2.2 MINIMUM TEST AND CALIBRATION FREQUENCIES PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION Functional Test(8)

Calibration(8)

Every Three Months Once/Operating Cycle Every Three Months Each Refueling Outage Every Three Months Once/Operating Cycle Every Three Months Every Three Months Every Three Months Once/Operating Cycle Every Three Months Each Refueling Outage (Note 7)

Every Three Months once/Operating Cycle Every Three Months Every Three Months Once/Operating Cycle Once/Operating Cycle (Note 2)

(Note 3)

Instrument Check Once Each Day Once Each Day Once Once Each Day Each Day Amendment No. 4, 68, 46, 06,

-144, a46 I*

64

VYNPS TABLE 4.2 NOTES

1.

Not used.

2.

During each refueling outage, simulated automatic actuation which opens all pilot valves shall be performed such that each trip system logic can be verified independent of its redundant counterpart.

3.

Trip system logic calibration shall include only time delay relays and timers necessary for proper functioning of the trip system.

4.

This instrumentation is excepted from functional test definition.

The functional test will consist of injecting a simulated electrical signal into the measurement channel.

5.

Deleted.

6.

Functional tests, calibrations, and instrument checks are not required when these instruments are not required to be operable or are tripped.

Functional tests shall be performed before each startup with a required frequency not to exceed once per week.

Calibration shall be performed prior to or during each startup or controlled shutdown with a required frequency not to exceed once per week.

Instrument checks shall be performed at least once per day during those periods when instruments are required to be operable.

7.

This instrumentation is excepted from the functional test definitions and shall be calibrated using simulated electrical signals once every three months.

8.

Functional tests and calibrations are not required when systems are not required to be operable.

9.

The thermocouples associated with safety/relief valves and safety valve position, that may be used for back-up position indication, shall be verified to be operable every operating cycle.

10.

Separate functional tests are not required for this instrumentation.

The calibration and integrated ECCS tests which are performed once per operating cycle will adequately demonstrate proper equipment operation.

11.

Trip system logic functional tests will include verification of operation of all automatic initiation inhibit switches by monitoring relay contact movement.

Verification that the manual inhibit switches prevent opening all relief valves will be accomplished in conjunction with Section 4.5.F.1.

12.

This trip function is applicable to the mechanical vacuum pump; however, it is not applicable to the primary containment Group 1 isolation valves (i.e., main steam isolation, main steam line drain, and recirculation loop sample line isolation valves).

Amendment No. 64, 4&, 4G&, 145, "&4, 4S 74

VYNPS BASES:

3.2 (Cont'd)

For the complete circumferential break of 28-inch recirculation line and with the trip setting given above, ECCS initiation and primary system isolation are initiated in time to meet the above criteria.

The instrumentation also covers the full range of spectrum breaks and meets the above criteria.

The high drywell pressure instrumentation is a backup to the water level instrumentation, and in addition to initiating ECCS, it causes isolation of Group 2, 3, and 4 isolation valves.

For the complete circumferential break discussed above, this instrumentation will initiate ECCS operation at about the same time as the low-low water level instrumentation, thus, the results given above are applicable here also.

Certain isolation valves including the TIP blocking valves, CAD inlet and outlet, drywell vent, purge and sump valves are isolated on high drywell pressure.

However, since high drywell pressure could occur as the result of non-safety-related causes, such as not venting the drywell during startup, complete system isolation is not desirable for these conditions and only certain valves are required to close.

The water level instrumentation initiates protection for the full spectrum of loss of coolant accidents and causes a trip of certain primary system isolation valves.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident.

In addition to monitoring steam flow, instrumentation is provided which causes a trip of Group 1 isolation valves.

The primary function of the instrumentation is to detect a break in the main steam line, thus only Group 1 valves are closed.

For the worst case accident, main steam line break outside the drywell, this trip setting of 140 percent of rated steam flow in conjunction with the flow limiters and main steam line valve closure limit the mass inventory loss such that fuel is not uncovered, cladding temperatures remain less than 12950F and release of radioactivity to the environs is well below 10CFR100.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in this area.

Trips are provided on this instrumentation and when exceeded cause closure of Group 1 isolation valves.

Its setting of ambient plus 95*F is low enough to detect leaks of the order of 5 to 10 gpm; thus, it is capable of covering the entire spectrum of breaks.

For large breaks, it is a backup to high steam flow instrumentation discussed above, and for small breaks, with the resultant small release of radioactivity, gives isolation before the limits of 10CFR100 are exceeded.

Main steam line radiation monitors are located in the main steam tunnel downstream from the MSIVs and provide an early indication of gross fuel failures.

This instrumentation initiates alarms and trips/isolations assumed in the analysis of the design basis control rod drop accident (CRDA).

Based on NEDO-31400A, *Safety Evaluation for Eliminating the Boiling Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor," and plant-specific analysis of the CRDA, this instrumentation is no longer required to automatically cause closure of the MSIVs and other Group 1 isolation valves to ensure compliance with the radiation dose limits of 10CFR100.

The mechanical vacuum pump trip/isolation setting of :3 times normal background at rated thermal power is as low as practicable without consideration of spurious trips from nitrogen-16 spikes, instrument instabilities and other operational occurrences. With an alarm setting of a nominal 1.5 times normal background at rated thermal power, the control room operator is alerted to off-normal conditions and to determine the need for corrective actions in accordance with plant procedures.

Pressure instrumentation is provided which trips when main steam line pressure drops below 800 psig.

A trip of this instrumentation results in closure of Group 1 isolation valves.

In the refuel, shutdown, and startup modes, this trip function is provided when main steam line flow exceeds 40%

of rated capacity.

This function is provided primarily to provide protection against a pressure regulator malfunction which would cause the Amendment No.

-S,

, ",64, 4*,,

B l

096 52 76

ý' I 1.11, 1 1 1

VYNPS TABLE 4.7.2 NOTES

1.

Isolation signals are as follows:

Group 1:

The valves in Group 1 are closed upon any one of the following conditions:

1.

Low-low reactor water level

2.

Deleted

3.

High main steam line flow

4.

High main steam line tunnel temperature

5.

Low main steam line pressure (run mode only)

6.

Condenser low vacuum Group 2:

The valves in Group 2 are closed upon any one of the following conditions*:

1.

Low reactor water level

2.

High drywell pressure Group 3:

The valves in Group 3 are closed upon any one of the following conditions:

1.

Low reactor water level

2.

High drywell pressure

3.

High/low radiation - reactor building ventilation exhaust plenum or refueling floor Group 4:

The valves in Group 4 are closed upon any one of the following conditions:

1.

Low reactor water level

2.

High drywell pressure

3.

High reactor pressure Group 5:

The valves in Group 5 are closed upon low reactor water level.

Group 6:

The valves in Group 6 are closed upon any signal representing a steam line break in the HPCI system's or RCIC system's respective steam line.

The signals indicating a steam line break for the respective steam line are as follows:

1.

High steam line space temperature

2.

High steam line flow

3.

Low steam line pressure

4.

High temperature in the main steam line tunnel (30 minute delay for the HPCI and the RCIC)

2.

The closure time shall not be less than 3 seconds.

Valves V10-39A/B, V10-34A/B, V10-26A/B, V10-31A/B and V10-38A/B are closed upon either 1) low-low reactor water level and low reactor pressure or 2) high drywell pressure.

Amendment No. 4, I94 162 I I I I

f!j I

I I

v t e Letter Sequence Request
TAC:MB6477, (Approved, Closed)
Initiation Request, Request Acceptance... Administration Questions Answers Results Approval Other:
MONTHYEAR2002-11-08 [Table View]

ML020810028

Text

Subject:

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