ML20210D243

Application for Amend to License DPR-28,revising & Clarifying Operability & SRs of High Pressure Core Cooling Systems

ML20210D243
Person / Time
Site:	Vermont Yankee
Issue date:	07/20/1999
From:	Wanczyk R VERMONT YANKEE NUCLEAR POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20210D247	List: ML20210D243 ML20210D248
References
BVY-99-86, NUDOCS 9907270170
Download: ML20210D243 (15)

Text

V VERMONT YANKEE y

NUCLEAR POWER CORPORATION 185 Old Ferry Road, Brattleboro, VT 05301-7002 (802) 257 5271 July 20,1999 )

BVY 99-86 i

U.S. Nuclear Regulatory Commission ATfN: Document Control Desk Washington, D.C. 20555 1

Subject:

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

Technical Specification Proposed Change No. 219 Hieh Pressure Core Cooline Systems Pursuant to 10CFR50.90, Vermont Yankee (VY) hereby proposes to amend Facility Operating License DPR-28 by incorporating the attached proposed change into the VY Technical Specifications. The proposed change revises and clarifies operability and surveillance requirements of high pressure core cooling systems. Some of the changes are considered to be more restrictive, and others less restrictive.

Additional restrictions are being included in current specifications to incorporate existing practices that provide a specific time limit for meeting certain surveillance requirements. Also, some overly restrictive operability requirements are being relaxed for these systems.

Attachment I to this letter contains supporting infonnation and the safety assessment of the proposed change. Attachment 2 contains the determination of no significant hazards consideration. Attachment 3 provides a marked-up version of the current Technical Specification pages with the changes noted.

Attachment 4 is the retyped Technical Specification pages incorporating the changes.

VY has reviewed the proposed Technical Specification change in accordance with 10CFR50.92 and I 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 l 10CFR51.22(c)(9) and does not require an environmental review. Therefore, pursuant to 10CFR51.22(b),

I no environmental impact statement or environmental assessment needs to be prepared for this change.

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, hy,y 99-86 / Page 2 V*"V8 Y^*'t NM:ll^n l' owr.n Coiu onsinis if you have any questions on this transmittal, please contact Mr. Jeff Meyer at (802) 258-4105.

Sincerely, VEP T YANKEE NUCLEAR POWER CORPORATION Robert J. W cz k (/(/

Director of af y and Regulatory Affairs -

g SANf),

STATEOF VERMONT ) C l

)SS , NOTAR)

WINDHAM COUNTY )

FUBUC Then personally appeared before me, Robert J. Wanczyk, who, being duly sworn, did st at 1 is Directo f Safety and Regulatory Affairs of Vermont Yankee Nuclear Power Corporation, that he is du ri ute and file the foregoing document in the name and on the behalf of Vermont Yankee Nuclear P ryoration a$

that the statements therein are true to the best of his knowledge and belief. u00 -

I h8OL Sally A. sandstrum, Notary Public My Commission Expires February 10,2003 Attachments cc: USNRC Region 1 Administrator USNRC Resident Inspector- VYNPS USNRC Project Manager- VYNPS Vermont Department of Public Service L_ ..

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., _, . VERMONT YANKEE NUCLEAR POWER CORPORATION J Docket No. 50-271 BVY 99-86 l

Attachment i Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 219 High Pressure Core Cooling Systems Supponing Information and Safety Assessment of Proposed Change l

l BVY 99-86 / Attachm:nt i / Page 1

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INTRODUCTION This Proposed Change modifies current Technical Specifications and related Bases regarding operability requirements for the high pressure core cooling systems--High Pressure Coolant Injection (HPCI),

Reactor Core Isolation Cooling (RCIC) and Automatic Depressurization System (ADS)-and the safety and relief valves. These changes clarify, relax and add restrictions to the Limiting Conditions for Operation and Surveillance Requirements for these systems. Associated Technical Specifications Bases are also being changed.

Background

Current Technical Specifications 3.5.E.1 and 3.5.G.1 require that the HPCI and RCIC systems, respectively, he operable whenever reactor pressure is greater than 150 psig and irradiated fuel is in the reactor vessel. Similarly, Specification 3.5.F.1 requires that the ADS be operable whenever reactor pressure is greater than 100 psig and irradiated fuel is in the reactor vessel. If certain operability conditions are not met for these systems, remedial actions are to be taken within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to reduce reactor pressure.

Specification 3.5.E.1 adds the stipulation that HPCI be operable prior to reactor startup from a cold condition although the system cannot function as intended until adequate reactor steam conditions are achieved.

Specifications 3.5.E.1,3.5.F.1, and 3.5.G.1 use terminology of" reactor pressure," instead of the more accurate term, " reactor steam pressure." Sufficient reactor steam pressure is necessary before these systems can be capable of performing their intended safety functions.

Current Specifications 3.6.D.1 and 3.6.D.2 establish operability requirements when reactor coolant pressure is greater than 120 psig for safety and relief valves and, if those requirements cannot be met, require action to reduce reactor coolant pressure to below 120 psig. 1 Collectively, the changes described herein modify operability and testing requirements contained in Specifications 3.5.E,3.5.F,3.5.G,4.5.E,4.5.G, and 3.6.D as well as related Bases.

HPCI System The HPCI system is provided to assure that the reactor core is adequately cooled in the event of a small break in the nuclear system-the postulated loss of coolant accident (LOCA) which does not result in rapid depressurization of the reactor vessel. The IIPCI system is designed to operate until the reactor vessel is depressurized to allow operation of low pressure cooling systems. Failure of the HPCI system in the case of a small break LOCA could require operatios of the ADS to reduce vessel pressure in order to permit Core Spray (CS) or Low Pressure Coolant Injection (LPCI) operation. In the case of a very small LOCA and the unavailability of HPCI, ADS might not be required if the break size is within the capability of the RCIC system.-

The HPCI system consists of a steam turbine driving a flow-controlled pump with associated system piping, controls, and instrumentation. At VY it is designed to deliver 4,250 gpm to the reactor at reactor steam pressures ranging from 150 to 1,120 psig. The lower value bounds the reactor pressure at which 1

the LPCI system (258 psid between reactor and drywell) and the CS system (230 psid) begin to inject water and provides for sufficient coverage to allow those systems to supply adequate core cooling. (See  !

- UFSAR Table 6.5-3.) This ensures that the HPCI system can provide core cooling over a range of break

I VERMONT YANKEE NUCLEAR POWER CORPORATION BVY 99-85 / Attechment 1/ Pag; 2 sizes to overlap with the low pressure emergency core cooling systems to meet 10CFR50.46 requirements.

Two sources of water are available to the HPCI system (i.e., the condensate storage tank and suppression pool). Initially, water is taken from the condensate storage tank (CST), which has a dedicated reserve capacity of 75,000 gallons for the combined use of the liPCI and RCIC systems.

RCIC System RCIC provida raakeup water (at least 400 gpm) to the reactor vessel following a reactor vessel isolation event when normal sources of feedwater are not available. RCIC was not originally designed or intended to provide a safety function; however, its role has been modified as reflected in the UFSAR and Technical Specifications. While not designed or credited as an emergency core cooling system, RCIC functions {

similarly to the liPCI system, although on a much smaller scale.

The RCIC system provides makeup water to the reactor vessel whenever the reactor is isolated. Like IIPCI, the RCIC system uses a steam-driven turbine-pump unit for its motive power and operates automatically with sufficient coolant flow to maintain adequate reactor vessel water level.

Automatic Depressurization System The Automatic Depressurization System (ADS) serves as a backup to HPCI in the event of a small break in the reactor coolant pressure boundary. Upon receipt of initiation signals to depressurize the reactor, ADS automatically actuates to relieve reactor steam pressure to permit operation of low pressure cooling systems. ADS consists of four relief valves together with instrumentation and controls. The two main steam safety valves provide additional over-pressure protection, but are not part of the ADS.

In addition to the automatic initiation of ADS, the reactor can also be depressurized by remote manual operation of one or more cf the relief valves.

Proposed Channes (The proposed changes described below are indicated on marked-up pages of the current Technical Specifications in Attachment 3. Each change number is located in brackets [] on the marked-up pages.)

Proposed Change #1:

In current Technical Specifications 3.5.E.1, 3.5.F.1, and 3.5.G.1, insert the word " steam" in each Specification such that those requirements now specify:

3.5.E.1: Except as specified in Specification 3.5.E.2, whenever irradiated fuel is in the reactor vessel and reactor steam pressure is greater than...

3.5.F.1: Except as specilled in Specification 3.5.F.2 below, the entire Automatie Depressurization Relief System shall be operable at any time the reactor steam pressure is above...

3.5.G.1: Except as specified in Specification 3.5.G.2 below, the RCIC system shall be operable whenever the reactor steam pressure is greater than...

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• BVY 99-86 / Attachment 1/ Page 3 Proposed Change #2:

In current Specification 3.5.E.1, delete the ending of the first sentence such that the sentence, in cor. junction with Change #1 above, is now:

Except as specified in Specification 3.5.E.2, whenever irradiated fuel is in the reactor vessel and reactor steam pressure is greater than 150 psig:

Proposed Change #3:

Replace current Technicsl Specification 4.5.E.1. with the following:

1. Testing

a. A simulated automatic actuation test of the HPCI System shall be performed during each refueling outage,

b. Operability testing of the pump and valves simli be in accordance with Specificatian 4.6.E.

c. Upon reactor startup, HPCI operability testing shall be performed as required by Specification 4.6.E within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding 150 psig reactor steam pressure.

d. The IIPCI System shall deliver at least 4250 gpm at normal reactor operating pressure when recirculating to the Condensate Storage Tank.

1 Proposed Change #4:

Insert an additional condition in Specification 3.5.E.3 (i.e., reference to new Specification 4.5.E.1.c) such that, together with Proposed Change #5 below, this Specification is now:

If the requirements of either Specification 3.5.E or Specl/kanon 4.5.E.1.c cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to 5150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Proposed Change #5:

Change the pressure specified in Specification 3.5.E.3 from "5120 psig" to "5150 psig," such that together with Proposed Change #4 above, this Specification is now:

If the requirements of either Specification 3.5.E or Specification 4.5.E.1.c cannot be met. an orderly shutdown shall be initiated and the reactor pressure shall be reduced to s 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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VERMONT YANKEE NUCLEAR POWER CORPORATION BVY 99-86 / Attachment 1/ Pcge 4 Proposed Change #6:

Change the pressure specified in Specification 3.5.F.1 from "100 psig" to "150 psig," such that together with Proposed Change #1 above, this Specification is now:

Except as specified in Specification 3.5.F.2 below, the entire Automatic Depressurizatior. Relief System shall be operable at any time the reactor steam pressure is above 150 psig and irradiated fuelis in the reactor vessel.

Proposed Change #7:

Change the pressure specified in Specification 3.5.F.2 from "100 psig" to "150 psig," such that this Specification is now:

From and after the date that one of the four relief valves of the Automatic Depressurization Subsystem are made or found to be inoperable due to malfunction of the electrical portion of the valve when the reactor is pressurized above 150 psig with irradiated fuel in the reactor vessel, continued reactor operation is permissible only during the succeeding seven days unless such a valve is sooner made operable, provided that during such seven days both the remaining Automatic Relief System valves and the IIPCI System are operable.

Proposed Change #8:

Charge the pressure specified in Specification 3.5.F.3 from "5100 psig" to "5150 psig" such that this Specification is now:

If the requirements of Specnication 3.5.F cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to 5150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Proposed Change #9:

Change the reactor coolant pressure specified in Specifications 3.6.D.1 and 3.6.D.2 from "120 psig" to "150 psig" such that these Specificat cns are now:

D. Safety and Reh . Valves

1. During reacior power operating conditions and whenever the reactor coolant pressure is greater than 150 psig and temperature greater than 350 F, both safety valves and at least three of the four relief valves shall be operable.

2. If Specification 3.6.D.1 is not met, initiate an orderly shutdown and the reactor coolant pressure shall be below 150 psig and 350 F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Proposed Change #10:

Replace current Technical Specification 4.5.G.I. with the following:

VERMONT YANKEB NUCLEAR POWER CORPORATION i

• BVY 99-86 / Attachment 1/ Page 5 l

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1. Testing

a. A simulated automatic actuation test (testing valve operability) of the RCIC System shall be performed during each refueling outage,

b. Operability testing of the pump and valves shall be in accordance with Specification 4.6.E.

c. Upon reactor startup, RCIC operability testing c shall be performed as required by Specification 4.6.E within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding 150 psig reactor steam pressure.

d. The RCIC System shall deliver at least 400 gpm at normal reactor operating pressure when ,

recirculating to the Condensate Storage Tank.

Proposed Change #11:

Inse.t an additional condition in Specification 3.5.G.3 (i.e., reference to new Specification 4.5.G.I.c) such that, together with Proposed Change #12 below, this Specification is now:

If the requirements of either Specification 3.5.G or Specification 4.5.G.I.c cannot be met, an orderly shutdown shall be initiated and the reactor i pressure shall be reduced to 5150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l Proposed Change #12:

Change the pressure specified in Specification 3.5.G.3 from "5120 psig" to "5150 psig," such that together with Proposed Change #11 above, this Specification is now:

If the requirements of either Specification 3.5.G or Specification 4.5.G.I.c cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to 5150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Proposed Change #13:

Replace the Bases Section 3.5.F in its entirety with the following:

The Automatic Depressurization System (ADS) consists of the four safety-relief valves and serves as a backup to the High Pressure Coolant Injection System (IIPCI). ADS is designed to provide depressurization of the reactor coolant system during a small break loss-of-coolant accident if HPCI fails or is unable to maintaiu sufficient reactor water level. Since HPCI operability is required above 150 psig, ADS operability is also required above this pressure.

ADS operation reduces the reactor pressure to within the operating pressure range of the low pressure coolant injection and core spray

l VERMONT YANKEE NUCLEAR POWER CORPORATION

• BVY 99-86 / Attachment I / Page 6 systems, so that these systems can provide reactor coolant inventory makeup.

Replace the first sentence of Bases Section 4.5.D, E, and F with the following:

i HPCI system tasting demonstrates operational readiness of equipment and detects . gradations which may affect reliable operation. Testing is conducted during each reactor startup if maintenance that affects operability was performed on the HPCI system. Periodic testing is also performed in accordance with Specification 4.6.E and the inservice testing program.

Sufficient ste..m flow must be available prior to HPCI testing to avoid laducing an operational transient when steam is diverted to the RPCI system. Reactor startup is allowed prior to performing the required surveillance testing in order to achieve adequate steam pressure and flow. However, a 24-hour limitation is imposed for performing operability testing once reactor steam pressure exceeds 150 psig.

The short duration before full functional testing is performed is considered acceptabic.

'And, replace Bases Section 4.5.G with the following:

The frequency and conditions for testing of the RCIC system are the same as for the IIPCI system. Testing is conducted in accordance with Specification 4.6.E and provides assurance that the system will j function as intended.

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Reasons for Chknees Proposed Clunge #1:

Insertion of the word " steam"in Specifications 3.5.E.1,3.5.F.1, and 3.5.G.1 clarifies the applicability for operation Jthe HPCI, ADS, and RCIC systems, respectively. Since hydrostatic or leakage testing (in the absence of steam) could result in reactor pressure exceeding the specified pressures, literal interpretation of these Specifications under these conditions could be construed that operability of these systems is required when reactor steam is not present. However, the absence of reactor steam renders the systems incapable of operating as intended since adequate steam flow and pressure are unavailable. The HPCI and RCIC turbines require sufficient steam as the motive force for operation. Also, ADS operation is based on ths ability to relieve steam pressure through relief valves which require a minimum steam pressure to operate. Although the relief valves may be capable of passing two-phase flow, they are designed to relieve steam pressure. Hence, operation of these systems as designed is not achievable without adequate reactor steam, nor should operation be required in the absence of suflicient steam pressure. This change clarifies the current Specifications to permit hydrostatic and leakage testing when the plant is shut down and these systems are reridered inoperable in the absence of reactor steam.

Proposed Change #2:

Deletion of the phrase,"...and prior to reactor startup from a cold condition" from Specification 3.5.E.1 removes the requirements that (a) HPCI shall be operable prior to reactor startup and (b) the condensate storage tank shall contain at least 75,000 gallons of condensate water prior to reactor startup.

, VERMONT YANKEE NUCLEAR POWER CORPORATION BVY 99 86 / Attachment I / Page 7 l B

HPCI operation is not required below 150 psig to meet its current licensing basis or satisfy emergency l' core cooling needs. Below this pressure, HPCI becomes incapable of meeting its intended safety function since insufficient steam pressure is available to properly operate the system. Also, at low pressure (270

psig--see Technical Specifications Table 3.2.2) the HPCI steam supply line is automatically isolated for protectise purposes.

The VY CST has-a reserve volume of 75,000 gallons for HPCI and RCIC purposes only. The requirement that the CST contain 75,000 gallons of condensate water is only applicable to HPCI and RCIC systems operation; and, since operation of HPCI and RCIC is only required when reactor steam pressure is greater than 150 psig, there is no need to impose such a requirement " prior to reactor startup from a cold condition."

The HPCI system design basis excludes operation below 150 psig. This change is acceptable since the VY accident analysis demonstrates that the low pressure emergency core cooling systems are adequate to maintain core cooling to reactor pressures'in excess of 150 psig. 'Iherefore, the current Technical Specification requiring HPCI operability " prior to reactor startup from a cold condition" is not necessary and is inconsistent with the licensing and design bases of this system, since it is physically not possible to operate HPCI without adequate steam. Since steam does not exist in a cold condition, a literal interpretation of this requirement cannot be met.

Proposed Changes #3 and #10:

Current Specifications 4.5.E.1 and 4.5.G.1 are restructured and additional restrictions are added such that, when required by Specification 4.6.E, testing must be conducted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of exceeding reactor steam pressure of150 psig.

New Specifications 4.5.E.1.a, 4.5.E.1.b, and 4.5.E.1.d and 4.5.G.1.a, 4.5.G.I.b, and 4.5.G.I.d do not change the technical requirements of current Technical Specifications 4.5.E.1 and 4.5.G.1, respectively.

These changes are administrative in nature as they only represent reformatting changes.

i New Specifications 4.5.E.1.c and 4.5.G.I.c provide additional restrictions. The addition of requirements to conduct certain testing of the HPCI and RCIC systems within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding 150 psig reactor steam pressure reflects current practice. Since receipt of an operating license in 1972, VY has demonstrated operability of the HPCI and RCIC systems at "normel teactor operating pressure" as specified in current Specifications 4.5.E and 4.5.G. VY has historically conducted such testing soon after i achieving adequate steam conditions to permit testing without undue risk of an undesirable transient which might be induced through diversion of steam at low pressures. This change adds clarity and time limitations for performing such testing consistent with the current licensing basis. J l

Based on experience, the specified 24-hour time limitation for performance of testing has proven adequate for obtaining required reactor steam conditions while restricting the allowable time available for verifying operability. This change does not modify the existing surveillance requirements in either type or frequency. Since the systems are administratively operable prior to testing, the short duration (i.e.,24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) before full functional testing is performed is considered acceptable. The short time allowed to perform testing is acceptable based on the small probability of an event requiring a potentially inoperable component to function.

Current Licensing Basis of HPCI Operability Testina VY.has been testing the HPCI system in essentially the same manner since original plant licensing.

Following refueling outages which involve llPCI maintenance the HPCI system is conditionally returned u

, , VERMONT Y ANKEE NUCLEAR POWER CORPORATION BVY 99-86 / Attachment I / Page 8 to service until the system can be tested with sufficient steam. VY pursued clarity on these testing ,

requirements many years ago as evidenced by the plant's licensing history.

l As a result of VY's application dated July 15,1976 (WVY.76-89), NRC issued VY license Amendment No. 27 on August 2,1976, which revised the HPCI testing requirements in Specification 4.5.E.1 to: "The ,

' HPCI system shall deliver at least 4250 gpm at normal reactor operating pressure." In its discussion of 1 the reason for the change to Specification 4.5.E.1, VY clearly stated: "The flow rate test is done only at normal reactor operating pressure." NRC's associated Safety Evaluation supporting Amendment No. 27 concluded that the changes to the Technical Specifications and bases are consistent with current practice and reflect a surveillance and test frequency that will provide assurance of system reliability, and are therefore considered acceptable.

Current Licensing Basis of RCIC Operability Testing Specification 4.5.G.1 for the RCIC system specifies that the system delivers at least 400 gpm at " normal reactor operating pressure." This provision was contained in the original VY operating license.

Consequently, it was not necessary to change this Specification as was done in Amendment No. 27 for HPCI Specification 4.5.E.1.

Proposed Changes #4 and #11:

The addition of reference to new Specification 4.5.E.1.c to current Specification 3.5.E.3 and new Specification 4.5.G.1.c to current Specification 3.5.G.3 represents additional restrictions to assure that if required testing is not conducted, remedial actions shall be taken to reduce reactor pressure to less than 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. These additional requirements assure that successful testing of t'n e HPCI and RCIC systems is taken in the period specified, or reactor pressure is reduced to below the pressure at which these systems are required to be operable. Reducing reactor pressure enables operability oflow pressure cooling systems, should they be needed. Thus, the additional limitations assure safety.

Proposed Changes #5, #8, #9 and #12:

Current Specifications 3.5.E.3,3.5.F.3, and 3.5.G.3 specify corrective action to reduce reactor pressure in the event that other Specifications cannot be met. Since the HPCI, ADS, and RCIC systems need not be operable at reactor steam pressures < 150 psig, it is overly restrictive to require any greater reductions in reactor pressure. This change also adds consistency within the Technical Specifications.

~ Similarly, current Technical Specification 3.6.D.] needs not require that safety and relief valves be operable until reactor coolant pressure is greater than 150 psig, and it is adequate in Specification 3.6.D.2 to specify that reactor pressure only needs to be reduced to less than 150 psig. Since the steam relief function accomplished through the ADS logic is only applicable to ADS operability, these valves need to be operable at pressures corresponding to ADS required operability. In a low pressure situation, below 150 psig, low pressure core cooling systems are adequate to supply coolant without further depressurization.

Proposed Changes #6 and #7:

Current Specifications 3.5.F.1 and 3.5 F.2 provide operability requirements for the ADS when reactor pressure is above 100 psig and irradiated fuel is in the reactor vessel. These are overly restrictive operability requirements since ADS's safety function is to serve as a backup to the HPCI system and the HPCI system is only required to be operable above 150 psig. Also, since ADS operation is designed to reduce reactor pressure to within the operating pressure range of the LPCI and CS systems,150 psig is a

BVY 99-86 / Attachment I / Page 9 conservatively Ic.v reactor pressure for operation of these systems. As discussed above,230 psid between the reactor and drywell is sufficient for operation of the low pressure cooling sy::tems.

Proposed Change #13:

The changes to the Bases for Section 3.5.F and Sections 4.5.D, E, F and G prov de clarity and additional information on the bases for the operability requirements of ADS and the surveillance requirements of HPCI and RCIC, inc!uding when the testing is to be performed. This additional information explains why the testing is conducted as required by the subject Specifications.

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, , VERMui;r YANKEE NUCLEAR POWER CORPORATION

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l Docket No. 50-271 BVY 99-86 Attachment 2 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 219 High Pressure Core Cooling Systems Determination of No Significant Hazards Consideration l

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, , VERMONT YANKEE NUCLEAR POWER CORPORATION BVY 99-86 / Attochment 2 / Page 1 Pursuant to 10CFR50.92, VY has reviewed the proposed change and concludes that the change does not involw a significant hazards consideration since the proposed - change satisfies the criteria in 10CFR50.?2(c).

The proposed changes modify the operability requirements for the high pressure cooling systems-High Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling (RCIC) and Automatic Depressurization System (ADS)--and the safety and relief valves, and adds a time limitation for conducting operability testing of HPCI and RCIC.

1. The operation of Vermont Yankee Nuclear Power Station in accordance with the proposed amendment will not involve a sinnificant increase in the orobability or conseauences of an accident previousiv evaluate.A The changes proposed add clarity, additional limitations, and relaxation to operability requirements and also reflect current surveillance practices. The proposed changes do not change the function nor needed range of operability pressures for the affected systems. The revisions ensure the applicability of operating requirements consistent with the design and operational bases of these systems.

The high pressure cooling systems (HPCI, RCIC and ADS) and the steam safety and relief valves do not initiate any accident considered in the Updated Final Safety Analysis Report. HPCI and ADS (with relief valves), as emergency core cooling systems, do function to mitigate accidents. Credit is not taken for RCIC in this regard. This change will not alter assumptions relative to the initiation or mitigation of any accident event.

The less restrictive changes proposed to not require operability of HPCI, ADS (and safety and relief valves) and RCIC at reactor steam pressures below 150 psig when irradiated fuel is in the reactor vessel and do not affect the probability of any accident previously evaluated. These changes furthermore do not  !

significantly increase the consequences of accidents previously evaluated since reliance on these systems )

is not assumed below 150 psig.

The addition of rcquired surveillance testing and completion times are intended to require a reduction in reactor pressure if HPCI and RCIC system operability requirements are not met. These additional Technical Specifications testing requirements and completion times are consistent with the current licensing basis and represent current practice.

The proposed changes do not involve accident initiators, do not change the configuration or method of operation of any equipment used to mitigate the consequences of an accident, and do not alter any conditions assumed in the plant accident analysis. Therefore, operation in accordance with the proposed changes would not involve a significant increase in the probability or consequences of an accident previously evaluated since there is no physical alteration of the plant configuration or relaxation of required setpoints or operating parameters.

2. The operation of Vermont Yankee Nuclear Power Station in accordance with the proposed amendment will not create the nossibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes will not modify the physical plant or the modes of plant operation. The changes do not involve the addition or modification of equipment nor do they alter the design or operation of plant systems. These changes to operability requirements do not create any new or different kind of accident since they do not involve any change in the physical configuration of the plant, nor relaxation of required

VERMONT YANKEE NUCLEAR POWER CORPORATION

'BVY 99-86 / Attachment 2 / Page 2 setpoints or operating parameters. Operation and design of the subject high pressure cooling systems (and the steam relief function) are not altered by the proposed changes.

The changes in operability requirements governing normal plant operation are consistent with the current safety analysis assumptions. These changes ensure adequate emergency core cooling system capability exists to mitigate the consequences of loss of coolant accidents without introducing new modes of

. operation.

Therefore, VY has determined that the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed changes are designed to clarify and add limitations to operation.

3. The operation of Vermont Yankee Nuclear Power Station in accordance with the proposed amendment will not involve a sienificant reduction in a margin of safety.

The proposed changes involve adding clarity, additional restrictions, and less restrictive requirements to current Technical Specifications without the safety bases. The added restrictions require an operability demonstration of IIPCI and RCIC within an acceptable period of time following plant startup when testing is required. The additional restrictions to perform timely testing of the llPCI and RCIC systems do not result in any reduction in margins of safety.

The less restrictive changes do not significantly reduce the margin of safety because current accident analyses demonstrate that low pressure cooling systems are adequate to maintain core cooling to pressures in excess of 150 psig without relying on the HPCI, ADS (or safety and relief valves), or RCIC systems.

The proposed changes do not alter the basis for any Technical Specification that is related to the establishment or maintenance of margins of safety. VY has determined that the proposed change does not involve a significant reduction in a margin of safety since operation of the plant remains consistent with the plant's design and operational bases.

Conclusion 1

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 10CFR50.92(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.