ML20247J806

TS Proposed Change 204 to License DPR-28,reducing Normal Operating Supression Pool Water Temp Limit & Adding Time Restriction for Higher Temp Allowed During Surveillances to Add Heat to Suppression Pool.Rept Re Pool Temp,Encl

ML20247J806
Person / Time
Site:	Vermont Yankee
Issue date:	05/08/1998
From:	Reid D VERMONT YANKEE NUCLEAR POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20247J809	List: ML20247J806 ML20247J812 ML20247J834
References
BVY-98-69, NUDOCS 9805210414
Download: ML20247J806 (17)

Text

E VERMONT YANKEE NUCLEAR POWER CORPORATION 185 Old Ferry Road, Brattleboro, VT 05301-7002 (802) 257 5271 May 8,1998 BVY 98-69 U. S. Nuclear Regulatory Commission Attention: 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. 204 Suppression Pool Water Temperature Pursuant to 10CFR50.90, Vermont Yankee Nuclear Power Corporation (VY) hereby proposes to amend its Facility Operating License, DPR-28, by incorporating the attached proposed changes into the Technical Specifications of Vermont Yankee Nuclear Power Station.

While performing an FSAR cycle update, it was determined that elements of License Amendment #88 had not been properly inc6rporated into the FSAR. Specifically, the current containment response analyses assumes an initial suppression pool temperature of 90 F, while the Technical Specifications allow unlimited plant operations with the torus temperature up to 100 F. Administrative limits have been established prohibiting normal plant operation with suppression pool temperature greater than 90 F. Accordingly, this proposed change would reduce the normal operating suppression pool water temperature limit (from 100 F to 90 F) and would add a time restriction for the higher temperature allowed during surveillance that add heat to the suppression pool.

Attachment 1 of this letter provides supporting information and the safety assessment of the proposed change. Attachment 2 is the determination of no significant hazards considerations.

Attachment 3 provides the marked-up version of the current Technical Speci9 cations.

Attachment 4 is the retyped Technical Specification pages. Attachment 5 provides a summary report of the peak suppression pool temperature analyses for large break loss of coolant accidents.

/

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 reviewed the proposed change against the criteria of 10CFR51.22 for environmental considerations and concludes that the proposed change will not increase the  ;

types and amounts of effluents that may be released offsite. Thus, VY concludes that the [4)[

j proposed change is eligible for categorical exclusion from the requirements for an environmental impact statement in accordance with 10CFR51.22(c)(9).

The Plant Operations Review Committee and the Nuclear Safety Audit and Review Committee have reviewed the proposed Technical Specification change and concur with the above 9905210414 980500 PDR ADOCK 05000271' P PDR

VERMONT YANKEE NUCLEAR POWER CORPOR ATION U.S. Nucle *r Reguttory Commission BVY 98-69 \ Pags 2 determinations. Pursuant to 10CFR50.91(b)(1), we have provided a copy of this proposed change and the associated analysis regarding a no significant hazards consideration to the appropriate State of Vermont representative.

If you have any questions on this transmittal, please contact Mr. Thomas B. Silko at (802) 258-4146.

Sincerely, VERMONT YANKEE NUCLEAR POWER CORPORATION

(.LL e~~ .

Donald A. Reid ,9 p sat /pg Senior Vice President, Operations i STATE OF VERMONT ) NOTAR)

)ss l WINDHAM COUNTY ) PUBUC Then personally appeared before me, Donald A. Reid, who, being d Jly sworn, did state t Vice President, Operations of Vermont Yankee Nuclear Power Corporation, that he is duly a g'q'o- -

execute and file the foregoing document in the name and on the behalf of Vermont Yankee Nuclear Power Corporation, and that the statements therein are true to the best of his knowledge and belief. l l

J4 h. L Salfy A. Sa'ndstrum, Notary Public l My Commission Expires February 10,1999 l Attachments cc: USNRC Region 1 Administrator l USNRC Resident inspector-VYNPS USNRC Project Manager - VYNPS Vermont Department of Public Service L

VERMONT YANKEE NUCLEAR POWER CORPORATION l

Docket No. 50-271 l BVY 98-69  !

i i

1 l

Attachment 1 Vermont Yankee Nuclear Power Station.

Proposed Technical Specification Change Suppression Pool Water Temperature i Supporting Ir. formation and Safety Assessment of Proposed Change May 1998

l VERMONT YANKEE NUCLEAR POWER CORPORATION U.S. Nucisar Regulatory Commission BVY 98-69 / Attachm:nt 1/ Page 1 l Vermont Yankee Nuclear Power Station l Proposed Technical Specification Change l Suppression Pool Water Temperature l Supporting Information and Safety Assessment of Proposed Change i

INTRODUCTION l The proposed license amendment involves changes to the existing Technical Specifications for l Vermont Yankee. The proposed change will reduce the maximum allowed water temperature f

for the suppression pool during normal operation from 100 F to 90 F, and impose a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time restriction on the amount of time the suppression pool temperature can be above the l normal operating limit when surveillance are being done that add heat to the suppression pool.

l l

The proposed change will, in effect, revoke the changes made by License Amendment No. 88, issued June 6,1985'. The proposed changes will thus make the Technical Specifications consistent with the suppression pool temperature limits approved by License Amendment No.16, dated October 8,1975 2.

l These changes are consistent with administrative restrictions implemented as a result of a self j assessment that identified the potential for a maximum post-LOCA suppression pool temperature higher than that shown in the FSAR.

l BACKGROUND l From initial operation until the issuance of Amendment 88, dated June 6,1985, the normal l l operating suppression pool temperature limit was 90 F. Amendment 88 approved a change to increase the limit to 100 F. Internal eva! cations later conducted by the Vermont Yankee l Service Water Task Force identified a discrepancy between the suppression pool temperatures j shown on FSAR Figure 14.6-7 and the text of FSAR Section 14.6.3.3.2, which referred to the change in initial suppression pool temperature. This discovery initiated Vermont Yankee's  ;

corrective action program, and culminated in an operability assessment that concluded that continued operation with administrative limits in place was appropriate.

A technical evaluation was initiated to determine whether to remove the administrative limit or generate a proposed change to the Technical Specifications to reduce the normal operating l suppression pool temperature limit. The technical evaluation was performed using assumptions  !

that are more consistent with current standards and more conservative than Vermont Yankee's licensing basis. In particular, the new analysis used the ANS 5.11979 standard for decay heat, l with uncertainties, and significantly more heat addition from the feedwater system than assumed in the original analysis. The new analysis also revised the model for the Residual l Heat Removal (RHR) heat exchanger to provide additional margin for potential fouling and tube plugging. The resulting technical evaluation was thus more conservative than the original licensing basis in that it assumed a higher amount of heat addition to the suppression pool and I

' Reference USNRC Letter to WNPC dated June 6,1985, NW 85-116, and as corrected by USNRC Letter to WNPC dated August 7,1985, NW 85-164.

2 Reference USNRC Letter to Yankee Atomic Electric Company, dated October 8,1975.

1 L__-__-__-________-______________

VERMON r Y ANKEE NUCLEAR POWER CORPORATION U.S. Nucl:ar Regulatory Commission BVY 98-69 / Attachment 1/ Page 2 l s lower heat removal rate via the RHR system operating in the suppression pool cooling mode.

The technical evaluation has been completed and provides the basis for the proposed change.

A summary of the technical evaluation is provided in the following safety assessment.

i DESCRIPTION OF PROPOSED CHANGE The following changes to the Technical Specifications (TS), shown in Attachment 3, are proposed:

1 TS 3.7.A.1.a. maximum water temperature durina normal operation: Change from 100oF w 90 F.

Justification: As summarized in the Safety Assessment, all safety analysis requirements are met assuming a normal operating limit of 90 F.

2. TS 3.7.A.1.b. maximum water temperature durina any test operation which adds heat to the suppression pool: The 100 F limit will be retained with the addition of a restriction that the temperature shall not be above the normal operating limit for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Justification: The higher temperature limit is allowed during those periods when surveillance testing of systems that add heat to the suppression pool is required. The time limit was in the original TS, but was removed when the normal operating limit was increased to 100*F since it no longer had any significance when the normal and test limits were the same. This change reinstates the limit since the normal operating limit for suppression pool water temperature is being reduced from 100 F to 90 F.

3. TS 3.7. A.1.c. Change the suppression pool temperature at which power operation is allowed to resume from 100 F to 90 F.

Justification: Make consistent with the revised maximum water temperature limit during normal operation in TS 3.7.A.1.a.

SAFETY ASSESSMENT

1. Introduction This evaluation reviews the basis for the proposed decrease in the normal operating pool temperature limit from 100 F to 90 F.

2. Evaluation 2.1 Description of the Normal Operating Limit The proposed change to the TS will reduce the normal operating limit from 100 F to 90 F, which will return the limit to the value it was from the time the plant was first licensed until the issuance of License Amendment No. 88. The proposed change will also add the requirement that the suppression pool temperature shall not be above the normal operating limit for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during testing. Plant operating procedures require initiation of suppression pool cooling when the suppression pool temperature exceeds the normal operating limit, and an

VERMONT YANKEE NUCLEAR POWER CORPORATION U.S. Nuclear Regulatory Commission BW 98-69 / Attachment i / Page 3 LCO is entered when the suppression pool cooling mode of RHR is initiated while the reactor is at power.

2.2 Bases for the Normal Operating Limit The suppression pool temperature of 90 F that has been used as an input for the containment analyses establishes the basis for the normal operating limit. The containment analyses are done for LOCAs and events involving SRV discharges to the suppression pool. The analyses evaluate the capability of the containment to maintain its integrity as a fission product barrier, to function as a heat sink during reactor isolation, and to provide a source of cooling water for certain reactor core cooling systems under accident conditions.

2.2.1 LOCA Evaluations The suppression pool temperature prior to a LOCA may influence several parameters associated with the LOCA event, including: the containment pressure and temperature response (section 2.2.1.1); containment dynamic loads (section 2.2.1.2); and the performance of pumps taking suction from the suppression pool (section 2.2.1.3).

2.2.1.1 Containment Pressure and Temperature The pressurac and temperatures reached in the containment following a LOCA will be influenced by the initial suppression pool temperature. The original FSAR containment analysis assumed an initial temperature of 90 F. The containment analysis was revised in 1981 as part of the Mark I containment program. The Plant Unique Load Definition Report (PULDR)*

provides the results for the design basis accident (DBA), intermediate break accident (IBA) and small break accident (SBA).

The initial suppression pool temperatures assumed for the PULDR analyses were 70 F for the DBA. and 90'F for the IBA and SBA. At the time the analysis was done, Vermont Yankee's normal operating Technical Specification (TS) limit was 90*F. The DBA analysis was initiated from the lower initial suppression pool temperature, which represented an arithmetic mean of the operating range. The DBA analysis was carried out for 30 seconds following the break.

The IBA and SBA analyses were initiated from the higher TS suppression pool temperature limit. The basis for selecting the initial suppression pool temperature, as well as other parameters, was given in the generic Mark i Containment Program Load Definition Reportd .

The results of the PULDR analyses are summarized below:

Event initial Pool Peak DW Time of Peak Pool Temp.

Temp. Pressure @ Peak DWP DBA 70 F 42.2 psig - 3 sec Not Determined IBA 90*F 35.1 psig - 1100 sec 172 F SBA 90*F 26.6 psig - 1100 see 139'F 8

NEDO-24581, Rev.1,

• Mark l Containment Program, Plant Unique Load Definition, Vermont Yanke?

Generating Station," General Electric Co., dated April 1981.

NEDO-21888, Rev. 2, "Ma.-k l Containment Program Load Dehnition Report." General Electric Co.,

dated November 1981.

I

! VERMONT Y ANKEE NUCLEAR POWER CORPORATION U.S. Nucl:ar R:gulatory Commission BVY 98-69 / Attachment 1/ Page 4 The pool temperature for the IBA at the time of peak Drywell pressure is higher than the peak pool temperature shown in FSAR Section 14.6.3.3.2 Containment Response, which shows a peak of 166 F (Table 14.6.4C).

l The difference in the calculated pool temperature is attributed to the difference in assumptions l on feedwater addition applied to the DBA in the FSAR and the IBA in the PULDR. The DBA assumed that feedwater pumps trip at the beginning of the accident (FSAR Section 14.6.3.3.1),

which is consistent with the assumption of a simultaneous loss of offsite power with the LOCA.

Since Vermont Yankee has electric feedwater pumps, the pumps will immediately coastdown l when the LOCA occurs because the feedwater pump motors will lose their power supply. On the other hand, the analyses performed for the Mark 1 load definition assumed, for the IBA and SBA, that feedwater would be manually tripped by the operators after 10 minutes5 .

The Mark I program analyses were not carried out far enough to determine the peak suppression pool temperature. Therefore, a new evaluation of suppression pool temperatures

, was done in order to assess the impact of continued feedwater addition. The evaluation also considered the effect of RHR heat exchanger tube plugging and fouling beyond that assumed in the FSAR. This change to the RHR heat exchanger modeling was based upon Service Water System evaluations done for Generic Letter 89-13. The design heat rate for the RHR heat exchanger given in the FSAR is 57.5 x 106 Btu /hr. The corresponding heat rate used for the evaluation was 52 x 106 Btu /hr. Finally, the decay heat was determined using the ANS 5.11979 standard, with 2-sigma uncertainty, and a 2% thermal power uncertainty, combined statistically.

The new suppression pool temperature evaluation, which considered large and small breaks, shows that the peak suppression pool temperature will not exceed 185 F'. The higher calculated peak pool temperature is mostly attributed to the more conservative treatments of feedwater addition following a LOCA (relative to suppression pool temperature), RHR heat exchanger performance, and the addition of energy to the suppression pool water by the RHR and Core Spray pumps.

The design temperature of the primary containment structure is 281 F. Therefore, a potential peak suppression pool temperature of 185 F (currently 182.6) does not violate any containment design limits. The effect of the higher temperature on other aspects of the safety analysis are discussed in the following sections.

2.2.1.2 Containment Dynamic Loads The LOCA containment dynamic loads occur as a result of the LOCA containment thermal-hydraulic response, which forces air (air is used generically and represents the initial mix of noncondensable gases in the containment), and, subsequently, steam to flow through the vents from the Drywell to the suppression pool. The major containment dynamic LOCA loads include pool swell, condensation oscillation (CO), and chugging. The pool swell loads result from the expulsion of air from the Drywell into the suppression pool immediately after the LOCA. The pool swell loads are controlled by the Drywell pressurization rate immediately following the LOCA. Suppression pool temperature has a negligible effect on these loads. The CO and 5

Section 2.2, NEDO-21888, Rev. 2,

• Mark l Containment Program, Load Definition Report," General Electric Co., dated November 1981.

• Current calculations show a peak temperature of 182.6 F (Attachment 5) assuming an initial suppression pool temperature of 90*F. It is noted that in general, the following discussions assume a peak suppression pool temperature of 185 F.

VERMONT YANKEE NUCLEAR POWER CORPORATION U.S. Nuclear Regulatory Commission BVY 98-69 / Attachment 1/ Page 5 chugging loads result from the condensation of steam at the vent exit into the suppression pool.

These loads are influenced by the vent steam mass flux, air content in the vent flow and suppression pool temperature.

Steam condensation during a blowdown through th a Drywell vent system is expected to occur

, when the suppression pool temperature is equal to or less than 170*F7 . Although the peak suppression pool temperature is predicted to be above 170 F, the peak occurs well after the blowdown phase of a LOCA, and the suppression pool temperature is less than 170 F during the blowdown phase. Decreasing the initial normal operating suppression pool temperature limit from 100 F to 90 F willincrease the margin for steam condensation during a blowdown.

Containment dynamic loads were defined by the application of the Mark i Load Definition Report methodology to Vermont Yankee specific plant design and operating conditions. The dynamic loads following a LOCA were defined, as previously mentioned, for a large break (DBA), intermediate break (IBA), and a small break (SBA). The initial suppression peo!

temperature assumed for the DBA was 70 F, and for the IBA and SBA was 90 F. Therets, the proposed change to reduce the normal operating limit from 100 F to 90 F will not invalidate the analyses performed in the Plant Unique Analysis Report (PUAR)* The potential increase in the long-term peak suppression pool temperature has no effect on the LOCA-related POAR loads.

2.2.1.3 ECCS Considerations Since the suppression pool is an emergency water source for Emergency Coro Cooling Systems (ECCS), the impact of changes to the suppression poo! temperature needs to be evaluated. This includes the direct effect of water temperature on the core ecoling capability and the indirect effect on the pump operability, such n NPSH requirements and pump seal integrity The discussion of the effects on ECCS equipment is limited to the low pressure systems that are normally aligned to take suction from the suppression pool, i.e. the RHR (Low Pressure Coolant injection mode) and Core Spray (CS) systems. The High Pressure Coolant injection (HPCI) is normally aligned to take suction from the Condensate Storage Tank (CST). The HPCI will automatically switch to the suppression pool on low water level in the CST. However, accident analyses and other events do not credit the switch-over to the suppression pool from the CST.

2.2.1.3.1 Core Cooling Capability Analyses performed in accordance with 10CFR50.46 demonstrate the core cooling capability of the ECCS. The analyses assume that the RHR and CS pumps, take suction from the suppression pool. The analyses are based on approved methods that are independent of the initial and transient suppression pool temperatures, therefore, the proposed change and the supporting technical evaluation have no adverse impact on the analysis results.

2.2.1.3.2 Pump NPSH Availability New ECCS suction strainers are being installed during the current refueling outage in order to satisfy the guidance of NRC Bulletin 96-03. ECCS pump NPSH calculations are being updated 7

Technical Specifications Bases, 3.7. A. _ Primary Containment Plant Uniaue Analysis Report for Ve,mont Yankee, Technical Report TR-5319-1, Rev. 2, Teledyne Engineering Services, dated November 30,1983.

VERMONT YANKEE NUCLEAR POWER CORPORATION U.S. Nucl:ar R:gulatory Commission BVY 98-69 / Attachment 1/ Page 6 as part of that effort. The specifications for the new strainers assure adequate NPSH margins at the higher suppression pool temperatures discussed herein.

The principal effect of an increase in peak suppression pool temperature is the reduction of NPSH margin for the low pressure ECCS pumps. Operator action is credited in throttling the ECCS pump flow rates after 10 minutes for the most limiting scenarios for suppression pool temperatures and suction strainer debris loads, assuming an atmospheric cont .1 ment pressure. Operator action after 10 minutes is consistent with Vermont Yankee's design basis and Emergency Operating Procedures. The proposed reduction in the normal operating suppression pool temperature limit from 100 F to 90 F will provide more time for operators to take actions, if required.

2.2.1.3.3 Impact on ECCS Pump SealIntegrity l A peak suppression pool temperature of 185 F is below the maximum acceptable temperatures for the RHR and CS pump mechanical seals. The specified process fluid temperature range for the CS pump seal is 32 F to 210 F. The RHR pump seal is capable of operating with a fluid temperature of 185*F for one week with negligible effect on seal life. Suppression pool temperature is calculated to be above 180oF for less than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, which is well within the capability of 185 F for one week.

I 2.2.2 Safety / Relief Valve (S/RV) Evaluations

! Steam discharged from a S/RV is routed into the suppression pool via the discharge line and ,

quencher. Prior to the S/RV actuation, the S/RV discharge line (SRVDL) above the water level in the suppression poolis filled with air. The sudden open;ng of the S/RV and the ensuing rapid steam discharge results in pressurization of the line and creates a large force which pushes the gas and water leg out of the discharge line through the quencher and into the suppression pool. l' The gas then forms bubbles which oscillate and impart loads to the submerged boundaries and structures in the suppression pool. This mechanism is known as S/RV air-clearing. After the S/RV air-clearing phase, steam is discharged into the suppression pool. The rapid condensation of the steam also causes a loading on the submerged structures and boundaries.

The S/RV steam condensation loads are much lower than the air-clearing loads. The following is an evaluation of the impact of suppression pool temperature on S/RV loads.

2.2.2.1 Steam Condensation Loads l The steam condensation S/RV loads occur folloiving air clearing. These result from the steady I

condensation of steam at the quencher exit holes. As part of the Mark i program, VY provided information in the PUAR on the performance characteristics of the tee-quencher and i

suppression pool temperatures in response to the most limiting case, a stuck open S/RV from

! power'. The analysis evaluated two cases, one for an initial suppression pool temperature of l 90 F and the other for an initial suppression pool temperature of 100 F. The PUAR noted that the results of the analysis satisfied the guidance of NUREG-0661, which established pool temperature limits for S/RV discharges "so that the ' threshold' temperature for severe vibrations will not be achieved during operational and upset modes;.e.g. a stuck-open SRV event."'

(

2.2.2.2 Air-Clearing Load on the SRV Discharge Line

• Section 8.0, Suppression Pool Temerature Evaluation, Technical Report TR-5319-1, Rev. 2, Teledyne Engineering Services, dated November 30,1983.

Section 3.10.7, Suppression Pool Temperature Limit, Safety Evaluation Report, Mark l Containment Long-Term Program, NUREG-0661, dated July 1980. l l

VERMONT YANKEE NUCLEAR POWER CORPORATION U.S. Nucbar Rcgulatory Commission BVY 98-69 / Attachment 1/ Page 7 When the S/RV opens, the SRVDL experiences a transient pressurization load and thrust loads due to the acceleration and expulsion of water in the submerged portion of the piping. These loads are controlled by the S/RV flow rate, pipe geometry and SRVDL quencher submergence.

The air clearing loads used in the PUAR are based on the guidance given in the Load Definition Report". The loads resulting from acceleration and expulsion of water in the submerged portion of the piping would be slightly affected by the temperature of the water due to its effect on density. Since the loads were defined for an initial water temperature of between 70 F and 90 F, the proposed change to reduce the normal water temperature from 100 F to 90 F will have no effect on air-clearing loads on the SRVDL.

2.2.2.3 Air-Clearing Load on the Pool Boundary .

The air-clearing pool boundary loads are sensitive to S/RV flow rate, initial gas mass in the SRVDL, submergence, and suppression pool temperature. S/RV loads increase slightly with higher pool temperatures. Since the loads were defined for an initial water temperature of between 70 F and 90 F, the proposed change to reduce the normal water temperature from 100 F to 90*F will have no effect on air-clearing loads on the pool boundary. Since the peak post-LCCA suppression pool temperature is reached several hours after the accident, the reactor is depressurized and S/RV openings are either not expected or not of concern from a load definition perspective. Therefore, the higher predicted post-LOCA peak suppression pool temperature will not have an adverse impact on air-clearing loads at the pool boundary.

2.2.2.3.1 Single-Valve Actuation Load The S/RV air-clearing loads for single valve actuation were evaluated in the Load Definition Report for a pool temperature of 120*F'2 This temperature was chosen because it is the Technical Specification limit at which the reactor vesse! must be depressurized at the normal cooldown rate. Since the proposed change to the normal operating limit will not affect the Technical Specification depressurization limit, the single valve actuation load will not be affected.

2.2.2.3.2 Multiple-Valve Actuation Load The pool temperature for multiple valve actuations not related to ADS is specified at the normal operating limit for the following reasons. The S/RV. loads for Mark I containments during the initial multiple-valve S/RV actuation produce higher S/RV loads than subsequent actuations.

The initial actuation of multiple valves can occur during an isolation event or the early part of an intermediate or small break accident. The initial pool temperature assumed for these events is the maximum normal operating pool temperature limit allowed by Technical Specifications.

Air clearing loads are based on a model'8 which predicts a slight reduction in torus shell pressures from S/RV air clearing as pool temperature is decreased. Since the basis for the loads is the normal operating limit, and the normal operating limit at the time the loads were defined was 90 F, the proposed change to reduce the normal operating limit from 100 F to 90*F will have no effect on multiple-valve actuation loads.

" Section 5.2.1, S/RV Discharae Line Clearino Transient Loads, NEDO-21888, Rev. 2," Mark l Containment Program, Load Definition Report," General Electric Co., dated November 1981.

$2 Section 5.2 , NEDO-21888, Rev. 2. " Mark i Containment Program Load Definition Report," General Electric Co., dated November 1981.

'8 NEDO-21878,

• Mark l Containment Program Analytical Model for Computing Air Bubble and Boundary Pressures Resulting from an S/RV Discharge through a T-Quencher Device," dated January 1979.

VERMONT YANKEE NUCLEAR POWER CORPORATION U.S. Nucl:ar Regulatory Commission BVY 98-69 / Attachment 1/ Page 8 2.2.2.S.3 ADS Load The pool temperature for ADS actuation was specified to match the calculated pool temperature at the time of ADS during the IBA or SBA in the Mark I containment loads program. The pool temperature at the time of ADS was evaluated based on the initial pool temperature prior to IBA or SBA and considers the temperature increase due to energy discharge into the pool via S/RV actuations or break flow. The IBA and SBA evaluations were based on an initial pool temperature of 90 F. Therefore, the proposed change to reduce the normal operating temperature limit from 100 F to 90 F will not have an adverse effect on the ADS air-clearing loads.

2.2.3 Other Events Section 14 Station Safety Analysis of the FSAR was reviewed to determine what other events besides the LOCA were potentially affected by the proposed change to the normal operating temperature limit, or the higher calculated peak suppression pool temperature.

None of the other Design Basis Accidents discussed in Section 14.6 are affected. The Control Rod Drop Accident is concerned with the reactivity addition to the reactor core associated with j the " drop" of the highest worth control rod. The Main Steam Line Break is concerned with a j rupture of the Main Steam Lina outside the primary containment. Subsequent shutdown of the i reactor and long-term heat removal would involve the suppression pool, but these effects are bounded by the LOCA analysis, which results in all of the energy released from the reactor vessel to be transported, eventually, to the suppression pool. The Refueling Accident involves a mechanical failure of fuel handling equipment and the resulting drop of a fuel assembly. The suppression pool has no effect on this event.

FSAR Section 14.5 discusses the Abnormal Operational Transients (AOT). The foliowing AOTs were the events identif;ed that are potentially affected by the changes in the normal operating suppression pool temperature limit or the peak post-accident suppression pool temperature.

FSAR 14.5.4.2 -Inadvertent Opening of a Relief Valve or a Safety Valve The effect of the inadvertent opening of a relief valve on the suppression pool was previously discussed in Section 2.2.2. The discussion in FSAR 14.5.4.2 concerns the short-term (<60 seconds) effects of the event on reactor parameters. These parameters are affected by the flow rate of steam out of the reactor vessel, which is not affected by suppression pool temperature. Therefore, the proposed change to the normal operating limit has no effect on this event as described in the FSAR.

FSAR 14.5.4.4 - Loss of Auxiliary Power The FSAR discussion of this event concerns the relatively short-term (<60 seconds) effects of l the event on reactor parameters. These parameters are affected by the coastdown of electrically powered muipment, such as the feedwater. recirculation, and circulating water pumps. The effect cc m suppression pool is considered in the Station Blackout Analysis, discussed below in Section 2.2.31.

FSAR 14.5.7.1 - Loss of RHRSW Flow This event involves the loss of RHR Service Water while the RHR is in the Shutdown Cooling mode. If this event were to occur while the reactor head is installed, eventually the suppression l

l 4

VERMONT YANKEE NUCLEAR POWER CORPORATION U.S. Nuclear R:gulatory Commission BVY 98-69 / Attachment 1/ Page 9 pool could become the heat sink if the reactor pressurizes enough to actuate a relief valve. The resulting suppression pool temperature transient, which is not specifically analyzeo in the FSAR, would be mild because decay heat levels would be low. The proposed reduction in the normal operating temperature limit will provide more margin for energy absorption in the suppression pool for this event.

FSAR 14.5.9 - Loss of Habitability of the Main Control Room -

l This event arbitrarily postulates loss of habitability of the Control Room and discusses the abit;ty l to bring the reactor to a hot and cold shutdown condition by manipulating controls and equipment from outside the Control Room. Loss of off-site power is not considered to occur, nor are other accidents assumed to occur simultaneously or subsequently. Water level is assumed to be snaintained by the HPCI and RCIC systems, and suppression pool cooling by the RHR system. There are no specific evaluations of suppression pool temperature in the FSAR for this event, but because both trains of RHR can be assumed available, and because reactor vessel makeup is frorn the Condensate Storage Tank via HPCI and RCIC, the resulting suppression pool temperature transient will be relatively mild. The proposed decrease in the normal operating temperature limit will provide more margin for energy absorption by the suppression pool for this event.

2.2.3.1 Station Blackout Station blackout calculations of suppression pool temperature are based on assuming an initial suppression pool temperature of 100 F and an RHR heat exchanger ca city that corresponds to that used in the long-term post-LOCA suppression pool neatup < c .:ulation. The station blackout evaluation is not affected by assumptions regarding continu Js feedwater addition because, by definition, the event postulates that there is no electrical power available to operate the feedwater or condensate pumps. Therefore, the proposed change to reduce the normal operating limit from 100 F to 90 F will not have an adverse impact on the station blackout evaluation 2.2.3.2 Appendix R -

Appendix R calculations of suppression pool temperature response have been done assuming an initial pool temperature of at least 90 F. The RHR heat exchanger model assumptions are consistent with those used in the LOCA analysis, and the effects of feedwater addition are considered in some of the scenarios, when applicable. Therefore, the proposed change to reduce the normal operating limit from 100 F to 90oF.will not have an adverse impact on Appendix R calculations of suppression pool water temperatures.

2.3 Impact on Related Topics 2.3.1 EQ Temperature Limits Environmentally qualified (EO) electrical equipment located in the reactor building and relied on post LOCA was reviewed and determined to be qualified for a 185 F torus temperature over the first 20 days post-LOCA. Beyond the first 20 days, there would be no effect due to the initial pool temperature, since the pool temperature in the long term is governed by the Service Water l temperature rather than the initial pool temperature.

The long term post LOCA Drywell EQ temperature profile may also be slightly increased. EQ temperature profiles which conservatively bound any change to Drywell temperature due to i

VERMONT YANKEE NUCLE AR POWER CORPORATION U.S. Nucl:ar Regulatory Commission BVY 98-69 / Attachment 1/ Page 10 increased peak suppression pool temperatures have been evaluated. All EQ components in the Drywell have been reviewed and are qualified to the slightly increased temperature profiles predicted.

l 2.3.2 Instrumentation Accuracy The overall effect of a 185 F peak suppression pool temperature on instrument loop error is negligible. Since no critical functions other than post-accident monitoring are performed by these instruments, and the error is small, no adverse impact results.

I 2.3.3 ECCS Pipe Stress Torus attached piping analyses were assessed for impact due to a 185 F peak torus temperature. There are no adverse affects on any torus attached piping analyses.

3. Summary and Conclusions l The Technical Specification amendment request reduces the maximum allowed normal i operating suppression pool temperature from 100 F to 90 F. This is a conservative change j whicil is currently implemented administratively at VY under the Corrective Action Program. '

The safety parameters potentially effected by an increased torus temperature are related to l containment integrity and fuel clad integrity. Containment design temperature and pressure are not impacted and ECCS performance is not adversely affected. Therefore, this change does eot reduce the difference between a system failure point and accepted safety limit or reduce the l margin of safety, as defined in the basis for any Technical Specification. l l

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VERMONT YANKEE NUCLEAR POWER CORPORATION Docket No. 50-271 BVY 98-69 Attachment 2 Vermont Yankee Nuclear Power Station .

Proposed Technical Specification Change Suppression Pool Water Temperature i

Determination of No Si9nificant Hazards Consideration l

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l May 1998  !

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VERMONT YANKEE NUCLEAR POWER CORPORATION U.S. Nucl::cr Regulatory Commission BVY 98-69 / Attachment 2 / Page 1 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change Suppression Pool Water Temperature Determination of No Significant Hazards Consideration Pursuant to 10CFR50.92, Vermont Yankee Nuclear Power Corporation has reviewed the proposed cnange and concludes that the change does not involve a significant hazards consideration since the proposed change satisfies the criteria in 10CFR50.92(c).

1. The operation of Vermont Yankee Nuclear Power Station in accordance with the proposed amendment. will not involve a significant increase in the probability or conseauencee of an accident previously evaluated.

a) The proposed change to decrease the normal operating suppression pool temperature limit from 100 F to 90 F will assure that the consequences of accidents previously evaluated will not be significantly increased.

A reduction in the normal operating suppression pool temperature limit provides more margin for the suppression pool as a heat sink to at' sorb energy from the reactor vessel following an accident. The effect of higher calculated suppression pool temperatures following an accident as a result of the effect of increased feedwater addition and decreased RHR heat exchanger heat removal does not affect the consequences of accidents previously evaluated.

Certain types of Mark l containment loacing conditions are increased at lower suppression pool temperature, but since the analysis of Mark i loads for Vermont Yankee was based on initial suppression pool temperatures between 70 F and 90 F, the proposed decrease in the normal operating limit to 90 F will not affect the consequences of those particular events, b) The proposed change to decrease the normal operating suppression pool temperature limit from 100 F to 90 F will not affect the probability of accidents occurring. The accidents and transients described in the FSAR are initiated by failures of components which are not in contact with the suppression pool water, therefore a change in the suppression pool temperature will have no affect on the probability of those accidents occurring.

c) The proposed change to restrict operation during testing that adds heat to the suppression pool to no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while above the normal operating temperature limit will have no affect on the consequences of accidents previously evaluated since accidents are not assumed tot be initiated during these modes of operation. This assumption is made in order to assure that plants have testing flexibility at power. In addition to the time limit placed on pool temperature, the plant enters the appropriate limiting condition for operation whenever the RHR system is placed in the suppression poc! cooling mode during power operation.

VERMONT . YANKEE NUCLEAR POWER CORPORATION U.S. Nucl:ar Regulatory Commission BVY 98-69 / Attachment 2 / Page 2 d) The proposed change to restrict operation during testing that adds heat to the suppression pool to no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while above the normal operating temperature limit will have no affect on the probability of an accident occurring.

The accidents and transients described in the FSAR are initiated by failures of componerits which are not in contact with th'e suppression pool water, therefore a change in the duration of time at any particular suppression pool temperature will have no affect on the probability of those accidents occurring.

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

The proposed change to decrease the normal operating suppression pool temperature limit from 100 F to 90 F does not change any accident initiators or the types of accidents analyzed. No new modes of equipment operation or physical plant equipment modifications are proposed. The change in predicted peak suppression pool temperature results from more conservatively calculating the effects of currently analyzed accidents. Therefore this change wil! not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change to restrict operation during testing that adds heat to the suppression pool to no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> witn water temperature above the normal operating temperature limit will allow for appropriate testing of safety related equipment to ensure operability. This testing allowance does not create any new initiating events or transients and does not involve any new modes of operation. Therefore, this change does not create the possibility of a new or differen: kind of accident from those l previously evaluated.

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

The proposed change to decrease the normal operating suppression pool temperature limit from 100'F to 90 F assures that the suppression pool can adequately perform its safety function without a significant decrease in the margin of safety. Each of the accidents affected by suppression pool temperature have been evaluated. The evaluation showed that a higher peak suppression pool temperature was predicted based on analysis assumptions that are more conservative that those used in the  !

current FSAR, but that the increase in peak temperature does not have a impact on l containment loads and equipment operability. The principal effect of an increase in peak suppression pool temperature is the reduction of NPSH margin for the low pressure ECCS pumps. Operator action is credited in throttling the ECCS pump flow rates after 10 minutes for the most limiting scenarios in order to. assure that available NPSH exceeds required NPSH. Operator action after 10 minutes is consistent with Vermont Yankee's design basis and Emergency Ooerating Procedures. The proposed reduction in the normal operating suppression pool temperature limit from 100 F to 90 F will provide more time for operators to take actions, if required.

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VERMONT Y ANKEE NUCLEAR POWER CORPORATION U.S. Nuciser Regul: tory Commission '

BVY 98-69 / Attachment 2 / Page 3 Operation of the facility in accordance with the proposed change to restrict operation during testing that adds heat to the s'uppression pool to no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while above the normal operating temperature limit will not involve a significant reduction in a margin of safety because it restricts the amount of time that the facility can be operated at a suppression pool temperature above the normal operating limit.

Therefore, based on the abov.. evaluation, VY has concluded that this change does not involve a significant hazards consideration.

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