Text

TS Proposed Change No.257, Implementation of Arts/Mella
	ML030850467
Person / Time
Site:	Vermont Yankee
Issue date:	03/20/2003
From:	Balduzzi M; Entergy Nuclear Operations, Entergy Nuclear Vermont Yankee
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	BVY 03-23
	Download: ML030850467 (71)
	v • d • e

Entergy Nuclear Vermont Yankee, LLC Entergy Nuclear Operations, Inc.

185 Old Ferry Road Brattleboro, VF 05302-0500 March 20, 2003 BVY 03-23 U.S. Nuclear Regulatory Commission ATTN: 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. 257 Implementation of ARTS/MELLLA at Vermont Yankee 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 reflects an expanded operating domain resulting from the proposed implementation of the Average Power Range Monitor, Rod Block Monitor Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA). The NRC has approved similar requests for extended operating domains at other Boiling Water Reactor plants. to this letter contains supporting information and a summary of 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 provides the retyped Technical Specification pages. Attachment 5 "Vermont Yankee Nuclear Power Station APRM/RBM/Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)," NEDC-33089P, Revision 0, dated March 2003, contains the safety analysis prepared by General Electric (GE) to support this proposed change. Attachment 5 is considered proprietary information by GE. In accordance with 10CFR2.790(b)(I), an affidavit attesting to the proprietary nature of the information (report) is enclosed with Attachment 5.

Attachment 6 provides a non-proprietary version of the GE report for public disclosure.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. are the licensees of the Vermont Yankee Nuclear Power Station

BVY 03-23 / Page 2 of 2 VY has reviewed the proposed Technical Specification change in accordance with IOCFR50.92 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(c)(9) and does not require an environmental review. Therefore, pursuant to IOCFR51.22(b), no environmental impact statement or environmental assessment needs to be prepared for this change.

VY requests that this license amendment be approved by September 2003 to be effective for Operating Cycle 24. This is to support our core reload analysis schedule for Cycle 24. In addition, it supports the scheduled Extended Power Up-rate and Alternate Source Term license amendment applications.

If you have any questions, please contact Mr. Jim DeVincentis at (802) 258-4236.

Sincerely,

-"-4 "14 Michael A. Balduzzi Vice President, Operations STATE OF VERMONT )

)ss WINDHAM COUNTY )

Then personally appeared before me, Michael A. Balduzzi, who, being duly sworn, did state that he is Vice President, Operations of the Vermont Yankee Nuclear Power Station, that he is duly authorized to execute and file the foregoing document, and that the statements therein are true to the best of his knowledge and belief.

My Commission Expires February(, 2 NOTA Attachments ,.*

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

Docket No. 50-271 BVY 03-23 Attachment I Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 257 Implementation of ARTS/MELLA at Vermont Yankee Supporting Information and Safety Assessment of Proposed Change

BVY 03-23 / Attachment I / Page 1 INTRODUCTION The Proposed Change to the Vermont Yankee Nuclear Power Station (VY) Technical Specifications (TS) reflects an expanded operating domain resulting from the proposed implementation of the Average Power Range Monitor - Rod Block Monitor Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA). The specific TS changes are as follow. Conforming changes were also made to the associated bases.

I. TS SR 2.1.A.1.a (current TS page 6)

Current TS SR 2.1 .A.1 .a is as follows:

S< 0.66 (W- AJV) + 54%

Revise TS SR 2.1 .A.l.a to the following:

Two Loop Operation:

S<0.4W +64.4% for 0%< W <31.1%

S< 1.28W+ 37.0% for 31.1% < W<54.0%

S< 0.66W+ 70.5% for 54.0% < W< 75.0%

With a maximiunm of 120%powerfor W > 75. 0%

Single Loop Operation:

S < 0.4W + 61.2% for 0% < W < 39.1%

S<1.28W +26.8% for39.1%< W<61.9%

S< 0.66tV+ 65.2% for61.9% < W<83.0%

With a maximum of 120% powerfor W > 83. 0%

2. TS SR 2.1.A.I.a (current TS page 7)

Current TS SR 2.1 .A. L.a is as follows:

I AW =difference between two loop and single loop driveflow at the same core flow. This difference must be accountedfor during single loop operation. AW = Ofor two loop operation.

hI the event of operationat > 25 percent Rated Thermal Power with the ration of MFLPD to FRP greaterthan 1.0, the AJPRMgain shall be increasedby the ratio MFLPD/FRPwhere:

MFLPD = mnaximuni fraction of limitingpower density where the limitingpower density is defined in the Core OperatingLimits Report.

BVY 03-23 / Attachment I / Page 2 FRP =fractionof ratedpower (1593 MWIt).

In the event of operation at > 25% Rated Thermal Power the ratioof MFLPD to FRP equal to or less than 1.0, the APRM gain shall be equal to or greaterthan 1.0.

Revise TS SR 2.1 .A. .a to the following:

In the event of operationat > 25% Rated Thermal Power the APRM gain shall be equal to or greater than 1.0.

3. TS Figure 2.1.1 (current TS page 11)

Replace with revised Table.

4. TS SR 2.2 Table 2.2.1 (current TS page 18)

Current TS SR 2.2 Table 2.2.1 is as follows:

2 safety valves Revise TS SR 2.2 Table 2.2.1 as follows:

3 safety valves

5. TS LCO 3.1.B (current TS page 20)

Current TS LCO 3.1.B is as follows:

B. During operationat > Rated Thermal Power with the ratioof MFLPD to FRPgreater than

1.0 either

a. The APRM System gains shall be adjusted by the ratio given in Technical Specification 2.1.A.1, or

b. The power Distributionshall be changed to reduce the ration of MFLPD to FRP.

Revise TS LCO 3.1.13 as follows:

B. Deleted.

6. TS SR 4.1.B (current TS page 20)

Current TS SR 4.1.B is as follows:

B. Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after > 25% Rated Thermal Power and once a day during operations at > 25% Rated Thermal Power thereafter,the maximum fractionof limitingpower density andfraction of ratedpower shall be determinedand the APRM system gains shall be adjustedby the ratio given in Technical Specification 2. .A. L.a.

BVY 03-23 1Attachment I / Page 3 Revise TS SR 4.1 .B as follows:

B. Deleted

7. TS TABLE 3.1.1 (current TS Page 21)

Current TS Table 3.1.1 is as follows:

< 0.66 (W-AW) + 54% with a maximum of 120% (4)

Revise TS Table 3.1.1 as follows:

Two Loop Operation: (4)

S<0.4W + 64.4% forO% < W <31.1%

S< 1.28W+ 37.0% for31.1% < W<54.0%

S< 0.66W+ 70.5% for54.0%< W< 75.0%

With a maximnum of 120% powerfor W> 75.0%

Single Loop Operation: (4)

S < 0.4W + 61.2% for 0% < TV < 39.1%

S< 1.28W+ 26.0% for 39.1% < W< 61.9%

S< 0.66W+ 65.2% for 61.9% < W<83.0%

With a maximum of 120% powerfor W > 83.0%

8. TS TABLE 3.1.1 NOTES (current TS page 24)

Current TS TABLE 3.1.1 NOTES is as follows:

4. "TV" is percent ratedtwo loop drive flow where 100% rateddriveflow is thatflow equivalent to 48 X 106 lbs/hr coreflow. 4W is the difference between the two loop and single loop driveflow at the same coreflow. This difference must be accountedforduringsingle loop operation. AW = Ofor two recirculationloop operation.

Revise TS TABLE NOTES as follows:

4. "W" is percent rated two loop driveflow where 100% rateddrive flow is thatflow equivalent to 48 X 106 lbs/hr coreflow.

BVY 03-23 / Attachment I / Page 4

9. TS TABLE 3.2.5 NOTES (current TS page 52)

Current TS TABLE 3.2.5 NOTES is as follows:

5. "TV" is percent ratedtwo loop driveflow where 100% rateddriveflow is thatflow equivalent to 48 X 106 lbs/hr coreflow. Refer to the Core OperatingLimits Reportfor acceptable values for N. AW is the difference between the two loop andsingle loop drive flow at the same coreflow This difference must be accountedfor duringsingle loop operation. AW = Ofor two recirculationloop operation.

Revise TS TABLE 3.2.5 NOTES as follows:

5. "W" is percent rated two loop driveflow where 100% rateddrive flow is thatflow equivalent to 48 X 106 lbs/hr core flow. Refer to the Core OperatingLimits Reportfor acceptable values for N. AW is the difference between the two loop andsingle loop drive flow at the same coreflow. This difference must be accountedfor during single loop operation. AW = Ofor two recirculationloop operation and =8%for single loop operation.

10. TS TABLE 3.2.5 NOTES (current TS page 52)

Current TS TABLE 3.2.5 NOTES is as follows:

9. With one or two RBM channels inoperable:

a. Verify that the reactor is not operating on a limiting control rod pattern (as described in the Bases for Specification 3.3.B.6, and

b. If one RBM channel is inoperable, restore the inoperable channel to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; and

c. If the required actions and associated completion times of Notes 9.a and 9.b above are not met, or if two RBM channels are inoperable, place one RBM channel in the tripped condition within the next hour.

Revise TS TABLE 3.2.5 NOTES as follows:

9. With one or two RBM channels inoperable:

a. Deleted

b. If one RBM channel is inoperable, restore the inoperable channel to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; and

c. If the required action and associated completion time of Note 9.b above is not met, or if two RBM channels are inoperable, place one RBM channel in the tripped condition within the next hour.

BVY 03-23 / Attachment I / Page 5

11. TS LCO 3.3.B.6 (current TS pages 84 and 85)

Current TS LCO 3.3.B.6 is as follows:

6. During Operation with limiting control rodpatternseither:

(a) Both RBM channelsshall be operable; or (b) Controlrod withdrawalshall be blocked ;or (c) The operatingpower level shall be limited so that the MCPR will remain above thefuel claddingintegrity safety limit assuminga single errorthat results in complete withdrawal of any single operable control rod.

Revise TS LCO 3.3.B.6 as follows.

6. Deleted

12. TS SR 4.3.B.6 (current TS pages 84)

Current TS SR 4.3.B.6 is as follows:

6. When a limiting control rodpattern exists, an instrumentfunctionaltest of the RBMshall be performedpriorto withdrawalof the designatedrod(s) and daily there after.

Revise TS SR 4.3.B.6 as follows.

6. Deleted

13. TS SR 4.4.A.1 (current TS page 92)

Current TS SR 4.4.A.1 reads as follows:

1. Testingpumps and valves is accordance with specification 4.6.E. A minimum flow rate of 35 GPMat 12 75 psig shall be verifiedfor eachpump.

Revise TS SR 4.4.A.1 as follows

2. Testing pumps andvalves is accordancewith specification 4.6.E. A minimum flow rate of 35 GPM at 1320 psig shall be verifiedfor eachpump.

BVY 03-23 / Attachment 1/ Page 6

14. TS LCO 3.6.D.1 (current TS page 120)

Current TS LCO 3.6.D.1 is as follows:

1. Duringreactorpower operatingconditions and whenever the reactorcoolantpressure is greaterthan 150 psig and temperaturegreaterthan 350 TF, both safety valves and at least three ofthe four reliefvalves shall be operable.

Revise TS LCO 3.6.D.1 as follows:

1. During reactorpower operatingconditions and whenever the reactor coolantpressure is greaterthan 150 psig and temperaturegreaterthan 350 OF, all safety valves and at least three of the four reliefvalves shall be operable.

15. TS LCO 3.11.A (current TS page 224)

Current TS LCO 3.11 .A is as follows:

A. Average PlanarLinear Heat GenerationRate (APLHGR)

During operation at > 25% Rated Thermal Power, the APLHGR for each type offuel as afunction of averageplanarexposure shall not exceed the limiting values in the Core OperatingLimits Report....

Revise TS LCO 3.11 .A as follows:

A. Average PlanarLinear Heat GenerationRate (APLHGR)

During operationat > 25% Rated Thermal Power, the APLHGRfor each type offutel as afiunction of averageplanarexposure, power andflow shall not exceed the limiting values in the Core OperatingLimits Report....

16. TS SR 4.11.A (current TS page 224)

Current TS SR 4.11 .A is as follows A. Average PlanarLinearHeat GenerationRate (APLHGR)

The APLHGR for each type offiuel as afunction of averageplanar exposure shall be determined once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after > 25% Rated Thermal Power anddaily during operation at > 25% Rated Thermal Powerthereafter.

BVY 03-23 / Attachment I / Page 7 Revise TS SR 4.11 .A as follows:

A. Average PlanarLinear Heat GenerationRate (APLHGR)

The APLHGR for each type offuel as afunction of averageplanarexposure, power and flow shall be determinedonce within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after > 25% Rated Thermal Power and daily during operation at > 25% Rated Thermal Power thereafter.

17. TS LCO 3.I1C.1 (current TS page 226)

Current TS LCO 3.11 .C.1. is as follows:

1. During operation at > 25% Rated Thermal Power the MCPR operatingvalue shall be equal to or greaterthan the MCPR limitsprovided in the Core OperatingLimits Report.

Forsingle recirculationloop operation,the MCPR Limits at ratedflow are alsoprovided in the Core OperatingLimits Report. Forcoreflows other than rated, the Operating MCPR Limit shall be the above value multiplied by K1 where K1 is provided in the Core OperatingLimits Report. If at any time...

Revise TS LCO 3.1 1.C.I as follows:

1. During operationat > 25% Rated Thermal Power the MCPR operatingvalue shall be equal to or greaterthan the MCPR limits provided in the Core OperatingLimits Report.

Forsingle recirculationloop operation,the MCPR Limits at ratedflow are also provided in the Core OperatingLimits Report. If at any time...

18. TS SR 4.11.C (current TS page 226)

Current TS SR 4.11 .C is as follows.

C. Minimum CriticalPower Ratio (MCPR)

MCPR shall be determined once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after > 25% Rated Thermal Power, daily during operation at > 25% Rated Thermal Power thereafter,andfollowing any change in power level or distributionthat would cause operation with a limiting control rodpatternas describedin the bases for Specification 3.3.B.6.

Revise TS SR 4.11.C.1 as follows:

C. Minimum CriticalPowerRatio (MCPR)

MCPR shall be determinedonce within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after > 25% Rated Thermal Power and daily during operation at > 25% Rated Thermal Power thereafter.

BVY 03-23 / Attachment I / Page 8

29. TS Section 6.6.C.2 (current TS page 259)

Current TS Section 6.6.C.2 is as follows:

2. The Kf coreflow adjustmentfactor for Specification 3.11.C.

Revise TS Section 6.6.C.2 as follows Delete the above informationfrom this section.

BVY 03-23 / Attachment 1 / Page 9 BACKGROUND Many factors restrict the flexibility of a Boiling Water Reactor during power ascension from the low-power / low-core flow condition to the high-power / high-core flow condition. Once rated power is achieved, periodic adjustments must also be made to compensate for reactivity changes due to Xenon effects and fuel burnup. Some of the factors currently existing at VY that restrict plant flexibility in quickly achieving rated power are:

1. The currently licensed allowable operating power/flow map; and

2. The Average Power Range Monitor (APRM) flow-biased flux scram and flow-biased rod block setdown requirements.

This proposed TS change reflects operation of VY in a region which is above the rated rod line.

The current operating envelope is modified to include the extended operating region bounded by the rod line which passes through the 100% power/ 75% flow point (i.e., approximately the 120.8% rod line), the rated power line and the rated load line, as shown in Attachment 5 "Vermont Yankee Nuclear Power Station APRM/RBM/Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)," NEDC-33089P, Revision 0, dated March 2003.

The technical analysis is referred to as the Maximum Extended Load Line Limit Analysis (MELLLA) and the extended operating domain is referred to as the Maximum Extended Load Line Limit (MELLL) region. The analysis for cycle specific operating limits associated with the MELLL region will be performed as part of the cycle specific reload analysis. In addition, the implementation of the Average Power Range Monitor, Rod Block Monitor Technical Specification (ARTS) will increase plant operating efficiency by updating the thermal limits requirements and improving plant instrumentation responses and accuracy.

The improvement with operation in the MELLL region and the other TS improvements in the proposed change include;

1) A power dependent Maximum Critical Power Ratio (MCPR) thermal limit similar to that used in Boiling Water Reactor (BWR) type 6 plants is implemented as an update to reactor thermal limit administration.

2) The Average Power Range Monitor (APRM) trip set down and design total peaking factor requirement is to be replaced by more direct power and flow-dependent thermal limits to reduce the need for manual setpoint adjustments and to allow more direct thermal limit administration. This improves man/machine interface, thermal limit administration, increases reliability and provides more accurate protection of plant safety.

3) The Rod Withdrawal Error analysis was performed assuming no credit for the Rod Block Monitor (RBM) control rod block functions. The new RBM setpoints will be based on providing operational flexibility in the MELLLA region.

BVY 03-23 / Attachment I / Page 10 SAFETY ASSESSMENT The proposed change will allow VY to operate in an expanded operating domain. Operation in the expanded operating domain is based on a Maximum Extended Load Line Limit Analysis (MELLLA) performed by General Electric (GE) using methods described in Attachment 5, "Vermont Yankee Nuclear Power Station APRM/RBM/Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)," NEDC-33089P, Revision 0, dated March 2003.

The current operating envelope is modified to include the extended operating region bounded by the rod line which passes through the 100% power / 75% core flow point (i.e. approximately the 120.8 % rod line), the rated power line and the rated load line. Plant operating efficiency is increased by the Average Power Range Monitor - Rod Block Monitor Technical Specifications (ARTS) program which updates the thermal limits requirements and improves plant instrumentation and accuracy.

The safety analyses and system evaluations performed to justify operation in the MELLLA region consist of a non-fuel dependent portion and a fuel dependent portion that is fuel cycle dependent. In general, the limiting Anticipated Operational Occurrences (AOOs) MCPR calculation and the reactor vessel overpressure protection analysis are fuel dependent. These analyses, as discussed in Attachment A, are based on the assumption of a representative GEI3 and GE14 core (Cycle 23 core design). Subsequent cycle-specific analyses will be performed by VY in conjunction with the reload licensing activities. The non-fuel dependent evaluations such as containment response are based on the current hardware design and plant geometry, and as such they are applicable to VY. The limiting AOOs, were reviewed for the MELLLA region based on a review of existing thermal analysis limits at plants similar to VY and use of generic power-dependent and generic flow-dependent MCPR and Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits/setpoints. For the fuel-dependent evaluations of reactor pressurization events, these reviews indicate that there is a small difference in the operating limit minimum critical power ratio (OLMCPR) for operation in the MELLLA region and the Current Licensed Thermal Power (CLTP) condition (100% of CLTP / 100% of rated core flow). The actual operating limit is calculated on a cycle specific basis to bound the entire operating domain. The analysis results also indicate that performance in the MELLLA region is within allowable design limits for overpressure protection, loss-of-coolant accident (LOCA), containment dynamic loads, flow-induced vibration and reactor internals structural integrity, and Anticipated Transient Without Scram (ATWS) licensing criteria.

The analyses which justify operation in the MELLLA region under the stated conditions are discussed in Attachment 5 and its supporting references. These analyses include fuel performance event evaluations, mechanical evaluations of the reactor internals, structural vibration assessment, LOCA evaluations, and containment loads evaluations. NRC-approved or industry-accepted computer codes and calculational techniques are used in the ARTS/MELLLA analyses. Physical changes to the plant to accommodate the expanded operating region include upgrading Flow Trip Reference Cards and installing an additional Spring Safety Valve. These modifications will be engineered to meet applicable design and regulatory criteria.

BVY 03-23 / Attachment 1 / Page 11 The ARTS definition of a limiting control rod pattern (LCRP) is one for which the RBM is required to prevent violating a thermal limit in the event of a Rod Withdrawal Error (RWE) event. Because the VYNPS ARTS RWE basis does not credit the RBM, the LCRP concept is no longer meaningful for VYNPS.

The maximum Standby Liquid Control System (SLCS) pump discharge pressure during the limiting ATWS event is 1320 psig. This value is based on a peak reactor vessel lower plenum pressure of 1290 psia that occurs during the event. SLCS equipment design is not impacted by the increased pressure and there is adequate margin to prevent the SLCS relief valve from lifting. With a nominal SLCS relief valve setpoint of 1400 psig, there is a margin of 80 psi between the peak SLCS pump discharge pressure and the relief valve nominal setpoint.

In conclusion, based on the considerations discussed above and detailed in Attachment 5, (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.

Docket No. 50-271 BVY 03-23 Attachment 2 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 257 Implementation of ARTS/MELLLA at Vermont Yankee Determination of No Significant Hazards Consideration

BVY 03-23 / Attachment 2 / Page 1 Description of amendment request:

The proposed TS changes will allow VY to operate in an expanded operating domain. Operation in the expanded operating domain is based on a Maximum Extended Load Line Limit Analysis (MELLLA) performed by General Electric using methods described in "Vermont Yankee Nuclear Power Station APRM/RBM/Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTSIMELLLA)," NEDC-33089P. The current operating envelope is modified to include the extended operating region bounded by the rod line that passes through the 100% power / 75% core flow point, the rated power line, and the rated load line. Plant operating efficiency is increased by the Average Power Range Monitor - Rod Block Monitor Technical Specifications (ARTS) changes which updates thermal limits requirements and improves plant instrumentation and accuracy. Physical changes to the plant to accommodate the expanded operating region include upgrading Flow Trip Reference Cards and installing an additional Spring Safety Valve. These modifications will be engineered to meet applicable design and regulatory criteria.

Basis for No Significant Hazards Determination:

Pursuant to 10CFR50.92, Vermont Yankee (VY) 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 IOCFR50.92(c). These criteria require that 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 probability or consequences of an accident previously evaluated.

The proposed change involves allowing VY to operate in an expanded operating domain. Physical changes provide for enhanced instrument performance or were the result of safety analyses that support mitigation of design bases accidents. There are no changes to radioactive source terms or release pathways. Operation within the expanded operating domain has been evaluated and the impact on plant accidents was found to be within acceptable parameters. The proposed change does not result in any significant change in the availability of logic systems or safety-related systems themselves. Required protective functions will be maintained. The proposed change does not degrade plant design, operation, or the performance of any safety system assumed to function in the accident analysis.

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

BVY 03-23 / Attachment 2 / Page 2

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

The proposed change, which allows VY to operate in an expanded operating domain, does not introduce any new accident initiators or failure mechanisms because the change and the affects on existing structures, systems and components have been evaluated and found to not have any adverse affects. The proposed change will not substantially impose new requirements or eliminate any existing requirements.

Therefore, the proposed change 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 change, which allows VY to operate in an expanded operating domain, does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. There is no impact on the conclusions of any safety analysis. The proposed change does not involve any increase in calculated off-site dose consequences. Operability of protective instrumentation and the associated systems is assured, and performance of equipment will not be significantly affected.

Therefore, there is no significant reduction in the margin of safety as a result of this proposed change.

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 I0CFR50.92(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 03-23 Attachment 3 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 257 Implementation of ARTS/MELLLA at Vermont Yankee Marked-up Version of the Current Technical Specifications

VYNPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Applicability: Applicability:

Applies to the interrelated Applies to trip setting of the variable associated with fuel instruments and devices which are thermal behavior. provided to prevent the nuclear system safety limits from being exceeded.

Objective: Objective:

To establish limits below which To define the level of the process variable at which automatic the integrity of the fuel cladding is preserved. protective action is initiated.

Specification:

A. Trip Settings A. Bundle Safety Limit (Reactor The limiting safety system Pressure >800 psia and Core Flow >10% of Rated) trip settings shall be as specified below:

When the reactor pressure is

>800 psia and the core flow is 1. Neutron Flux Trip Settings greater than 10% of rated:

a. APRM Flux Scram Trip Setting (Run Mode)

1. A Minimum Critical Power II Ratio (MCPR) of less than When the mode switch 1.10 (1.12 for Single Loop is in the RUN Operation) shall constitute position, the APRM violation of the Fuel flux scram trip Cladding Integrity Safety setting shall be as Limit (FCISL).

shown on Figure 2.1.1 I and shall be:

6-ee .<er7 iq - -- rý -W where:

S = setting in percent of rated thermal power (1593 MWt)

W = percent rated two loop drive flow where 100%

rated drive flow is that flow equivalent to 48 x,10 6 lbs/hr core flow Amendment No. 4-4, 4-7, ", 40, -94, 44g, 4-&G, 4-&9, 176 6

Insert Two loop operation:

S5 0.4W+ 64.4% for 0% < W : 31.1%

S<1.28W+ 37.0% for 31.1% < W __.54.0%

S_5 0.66 W + 70.5% for 54.0% < W _<75.0%

with a maximum of 120.0% power for W > 75.0%

Single loop operation:

S_5 0.4W+ 61.2% for 0% < W < 39.1%

S<1.28W+ 26.8% for 39.1% < W : 61.9%

S_< 0.66 W + 65.2% for 61.9% < W :5 83.0%

with a maximum of 120.0% power for W > 83.0%

VYNPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING B. Core Thermal Power Limit W = dif renc(

(Reactor Pressure < 800 psia beween t' or Core Flow < 10%of Rated) oop andnoc I ./ single When the reactor pressure is *" driv *flop

<800 psia or core flow <10%athsa of rated, the core thermal cor/te flow power shall not exceed 25% of Ti i rated thermal power. mu/Thstdf C. Power Transient ric*ngtsi

\ /loop oper*

To ensure that the safety AW = 0/,*o limit established inlop-pr Specificat ion 1.IlA o pr and l.1B is not exceeded, In the event of each required scram shall operation at >25% F be initiated by its expected Thermal Power*Q-i-F Rated scram signal. The safety fratio of FPD to limit shall be assumed to FRP g ter than/

be exceeded when scram is 1.0 the APRM ggi accomplished by means b *increased* shall other than the expected *he rati: PpD scram signal.FR D. Whenever the reactor is wher shutdown with irradiated fuel in the reactor vessel, MFLPD =mxm the water level shall not fai be less than 12 inches above the top of the enriched fuelliti when it is seated in the core. w r be equal to than 1.0.

Amendment No. 4-, @-9, 4-, &1, -64, 68&, 44, VA-, 188 7

474, 94, a84, 211 i1 Amendment No.

VYNPS FIGURE 2.1.1 APRM FLOW REFERENCE SCRAM SETTING 130o 120 1 Two loop operation 110 100 90 I Single loop operation

'U " 80 S70 x

-J

'A z 60 0

I-- 5 zUJ 50 40 301 APRM Flow Biased Scram 20 10 I

Setpoints shall be less than or equal to 17111 17 values shown on the graph.

11 I-117 77H 7 771 1 1I I

o4 0 20 40 60 80 100 120 RECIRCULATION FLOW (% RATED)

Amendment No. 14, 94, 1-8-7, 2-l-l- 11

VYNPS BASES:

2.1 FUEL CLADDING INTEGRITY A. Trip Settings The bases for individual trip settings are discussed in the following paragraphs.

1. Neutron Flux Trip Settings

a. APRM Flux Scram Trip Setting (Run Mode)

The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (1593 MWt).

Because fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel. Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting.

Analyses are performed to demonstrate that the APRM flux scram over h_ age of settings from a maximum of 120% to the the minimum flow Isettin h f g.Z provide protection from fuel safety bias limit for all abnormal operational transients including those that may result in a thermal hydraulic instability.

The relatonship An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is between recirculation reached. The APRM scram trip setting was determined by an analysis maneuvering drive flow and reactor of margins required to provide a reasonable range for would increase during operation. Reducing this operating margin core flow is non-linear the frequency of spurious scrams which have an adverse effect on atlow core fl6ws. reactor safety because of the resulting thermal stresses. Thus, the APRM scram trip setting was selected because Safety it provides Therefore, separate adequate margin for the fuel cladding integrity Limit yet APRM flow biased allows operating margin that reduces the possibility of unnecessary scram trip setting scrams. f is set to ensure acceptable equations are provided _ . . ..T - (14 t n t r-o be adjusted to ansir-e that the LNGA core los.

for lowforlo cre flows. The scram trip an mst

__________________transient peak is not-increased fOr any combFiatioGnofILP d reactor core thermal po*wr. If the scram requires a change due tc an ab-no-rmal peaking coanditioen, it w41ll be accomplished by increasing the APRM gain by the rati. in Speification 2!-A.!..,

thus aesuring a reactor Sr ...

at o r .than

.. design ..vrpo.r conditions. For single recirculation loop operation, the APRM flux scram trip setting is reduced in accordance with the analysis presented in NEDO-30060, February 1983. This adjustment accounts for the difference between the single loop and two loop drive flow at the same core flow, and ensures that the margin of safety is not reduced during single loop operation.

Analyses of the limiting transients show th no scram adjustment is required to assure fuel cladding integrity n the transient is initiated from the operating limit MCPR defined in e Core nneratina Limits Renort. _____________

The single loop operation equations are based on a bounding (maximum) difference between two loop and single loop drive flow at the same core flow of 8%.

Amendment No. ;4., 2-5, -39, 4-, -64, 04, 444, 146 14

VYNPS 1.2 SAFETY LIMIT 2.2 LIMITING SAFETY SYSTEM SETTING 1.2 REACTOR COOLANT SYSTEM 2.2 REACTOR COOLANT SYSTEM Applicability:.

Applicability:

Applies to limits on reactor Applies to trip coolant system pr "essure. settings for controlling reactor system pressure.

Objective:

To establish a liinit below which the integrity of the reactor To provide for protective action coolant system is not threatened in the event that the principal process variable approaches a I due to an overpresssure safety limit.

condition.

Specification:

The reactor coolan.t system pressure shall not exceed 1335 psig at any t ime when irradiated fuel is present in the reactor vessel A. Reactor coolant high pressure scram shall be less than or equal to 1055 psig.

B. Primary system relief and safety valve settings shall be as specified in I

Table 2.2.1.

TABLE 2.2.1 Primary System Relief and Safety Valve Settings I

03. Note:

(1) As-left setpoint tolerance

+/-1%.

As-found setpoint tolerance

+/-32.

Amendment No. 4-1,160 18

VYNPS 3.1 LIMITING CONDITIONS FOR 4.1 SURVEILLANCE REQUIREMENTS OPERATION 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM Applicability: Applicability:

Applies to the operability of Applies to the surveillance of plant instrumentation and control the plant instrumentation and systems required for reactor control systems required for safety. reactor safety.

Objective: Objective:

To specify the limits imposed on To specify the type and frequency of surveillance to be applied to plant operation by those instrument and control systems those instrument and control required for reactor safety. systems required for reactor safety.

Specification: Specification:

A. Plant operation at any power A. Instrumentation systems level shall be permitted in shall be functionally accordance with Table 3.1.1. tested and calibrated as The system response time from indicated in Tables 4.1.1 the opening of the sensor and 4.1.2, respectively contact up to and including the opening of the scram solenoid relay shall not exceed 50 milliseconds.

B.ui o o at >25%---- B. S enceJit1fln 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months afte Amendment No. 61, a-6-4, 1&&*, 211 20

VYNPS TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT REQUIREMENTS Required Modes in Which Functions Must ACTIONS When be Operating Minimum Number Minimum Operating Conditions For Instrument Operation Trip Function Channels Per Are Not Trip Settings Refuel (1) Startup (12) Run Trip System (2) Satisfied (3)

1. Mode Switch in X Shutdown (SA-SI) X X 1 A

2. Manual Scram X X (5A-S3A/B) X 1 A

3. IRM (7-41(A-F))

High Flux <120/125 X X INOP x X 2 A

4. 2 A APRM (APRM A-F)

High Flux (flow bias) x A or B I 2 High Flux (reduced) <15% x INOP x 2 A x x A or B 2(5)

5. High Reactor <1055- psig X X x Pressure 2 A (PT-2-3-55 (A-D) (M))

6. High Drywell <2.5 psig X X x Pressure 2 A (PT-5-12 (A-D) (M))

x x x

7. Reactor Low (6) >127.0 inches Water Level 2 A (LT-2-3-57A/B (M))

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

8. Scram Discharge <21 gallons X X X 2 Volume High Level A (LT-3-231 (A-H) (M)) (per volume)

Amendment No. 4, 44, 44, '64, ?4,-6,;*, -4, 4G, 4-94, --84, 187 21

TS Insert :

Two loop operation: (q)

/_ O.4W+ 64.4% for 0% < W __31.1%

51.28W+ 37.0% for 31.1% < W _<54.0%

_0.66 W + 70.5% for 54.0% < W

75.0%

with a maximum of 120.0% power for W > 75.0%

Single loop operation:ýY)

< 0.4W+ 61.2% for 0% < W _ 39.1%

51.28W+ 26.8% for 39.1% < W :* 61.9%

0.66 W + 65.2% for 61.9% < W _<83.0%

with a maximum of 120.09/ power for W > 83.0%

file: TS-TRM-UFSAR Inserts VY A-M T506rO.doc

VYNPS TABLE 3.1.1 NOTES (Cont'd)

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

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.

within 8 c) Reduce turbine load and close main steam line isolation valves hours.

d) Reduce reactor power to less than 30% of rated within B hours.

flow is that

4. "W" is percent rated two loop drive flow where 100% rated drive flow equivalent to 48 x 10dIbs/hr c re flow. p o t .4ret*

tworiruand ion-gelp n.

dhppr~ive LPRM inputs per level

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

shall be

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

scram

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

11. Not used.

mode switch to be

12. While performing refuel interlock checks which require the scram need not be operable provided:

in Startup, the reduced APRM high flux

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.

24 Amendment No. ++, *-A, 4, &&, 44,, 94G, 94, 464, 4r-7, I", .4q9,212

VYNPS BASES: 4.1 (Cont'd)

LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local flux profile for appropriate representative input to the APRM System. The 2,000 megawatt-days per short ton (MWD/T) frequency is based on operating experience with LPRM sensitivity changes, and that the resulting nodal power uncertainty, combined with other identified uncertainties, remains less than the total uncertainty (i.e., 8.7%) allowed by the GETAB safety limit analysis.

B.--The ratje-o MFLPD to RP shall be checked oncýe.per da en operating Sat _* %Rated Th~r Power t trmine ýýe APRP-gins requix-ecur

\checking ese paramet s daily is *equate..,,he 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months *llowancel C levels, i Amendment No. .4&, 4-84, 191 33a

VYNPS TABLE 3.2.5 NOTES

1. Deleted.

2. Deleted.

3. Deleted.

4. Deleted.

5. vWx is percent rated two loop drive flow where 100% rated drive flow is that flow equivalent to 48 x 106 lbs/hr core flow. Refer to the Core Operating Limits Report for acceptable values for N. 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 operatio(

a _ = s $o e- Loo P

6. Not used.

7. The trip may be bypassed when the reactor power is <30% of Rated Thermal Power. An RBM channel will be considered inoperable if there are less than half the total number of normal inputs from any LPRM level.

8. With the number of operable channels less than the required number, place the inoperable channel in the tripped condition within one hour.

9. With one or two RBM channels inoperable:

1 +/-ftlft~..i~

1~i~ll C0721=1 L_ý a 4~

b. If one RBM channel is inoperable, restore the inoperable channel to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; and 5

c. If the required action* and associated completion timei of Notet;

*n 9.b above . not met, or if two RBM channels are inoperable, place RBM channe[in the tripped condition within the next hour.

is

10. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into associated Limiting Conditions for Operation and required action notes 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 Control Rod Block initiation capability.

11. Deleted.

12. Required to be operable when the reactor mode switch is-Tin the shutdown position.

13. With one or more Reactor Mode Switch - Shutdown Position channels inoperable, immediately suspend control rod withdrawal and immediately initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies.

Amendment No. "4, , 46, .4G, 4, -76, a-8, 211 52

VYNPS PI "BASES : 3.2 (Cont'd) control and/or bypass valves to open, resulting in a rapid depressurizatio and cooldown of the reactor vessel. The 800 psig trip setpoint limits the depressurization such that no excessive vessel thermal stress occurs as a NJ result of a pressure regulator malfunction. This setpoint was selected far

enough below normal main steam line pressures to avoid spurious primary containment isolations.

Low condenser vacuum has been added as a trip of the Group 1 isolation valves to prevent release of radioactive gases from the primary coolant through condenser. The setpoint of 12 inches of mercury absolute was selected to provide sufficient margin to assure retention capability in the condenser when gas flow is stopped and sufficient margin below normal operating values.

The HPCI and/or RCIC high flow and temperature instrumentation is provided The HPCI and RCIC steam -Y to detect a break in the HPCI and/or RCIC piping.

supply pressure instrumentation is provided to isolate the systems when pressure may be too low to continue operation. These isolations are for "Ik equipment protection. However, they also provide a diverse signal into %

are included indicate a possible system break. These instruments Technical Specifications because of the potential for possible system initiation failure if not properly tested. Tripping of this instrumentation t results in actuation of HPCI and/or RCIC isolation valves, i.e., Group 6 valves. A time delay has been incorporated into the RCIC due steam flow trip IZ to pressure logic to prevent the system from inadvertently isolating spikes which may occur on startup. The trip settings are such that core uncovering is prevented and fission product release is within limits.

The instrumentation which initiates ECCS action is arranged in a dual channel system. Permanently installed circuits and equipment may be used t "

IO trip instrument channels. In the nonfail safe systems which require energizing the circuitry, tripping an instrument channel may take the form '

of providing the required relay function by use of permanently installedwit circuits. This is accomplished in some cases by closing logic circuits the aid of the permanently installed test jacks or other circuitry which would be installed for this purpose.

o The Rod Block Monitor (RBM) control rod block functions ar rovide rev excessive con rol roe wit drawal so thatSCTRc oeL ot decrease b ow th c ingi tel grity s0 y It1 The is credit in the Conti us R ithdr Dunn ower R e Opera n transie or pre ntn ecssi xeuel control dwith awal bef therfni adding.

itegri safet imit (M )or t fuel ro echanica verpower its are e ceeded. The REM per li 'is cla d to provi e protecti 7at va e is clamped Limits cycle specific; greater therefore, thanit 100% rate core is located in theflow.Core The Operating Report.

in For single recirculation loop operation, the RBM trip setting is reduced This accordance with the analysis presented in NEDO-30060, February 1983.

adjustment accounts for the difference between the single loop of and two loop flow at the same core flow, and ensures that the margin safety is drive not reduced during single loop operation.

During hot shutdown, cold shutdown, and refueling when the reactor mode to be switch is required to be in the shutdown position, the core is assumed subcritical with sufficient shutdown margin; therefore, no positive reactivity insertion events are analyzed. The Reactor Mode Switch-Shutdown the mode Position control rod withdrawal block, required to be operable with switch in the shutdown position, ensures that the reactor remains the subcritical by blocking control rod withdrawal, thereby preserving to be assumptions of the safety analysis. Two channels are required B4, .94, 44-4, Q, 211 77 Amendment No. -9, 5, 469,

VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION pressure, are fully inserted, no more than two rods may be moved.

(c) Out-of-sequence control rods in each distinct RWM group shall be selected and the annunciator of the selection errors verified.

(d) An out-of-sequence control rod shall be withdrawn no more than three notches and the rod block function verified.

4. Control rod patterns and 4. The control rod pattern the sequence of and sequence of withdrawal or insertion withdrawal or insertion shall be established shall be verified to such that the rod drop comply with accident limit of Specification 3.3.B.4.

280 cal/g is not exceeded.

5. Control rods shall not 5. Prior to control rod be withdrawn for startup withdrawal for startup or refueling unless at or during refueling, least two source range verification shall be channels have an made that at least two observed count rate source range channels greater than or equal to have an observed count three counts per second./. rate of at least three counts per second.

Amendment No. 34, 61 84

VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION

7. The scram discharge volume drain and vent valves shall be verified open at least once per month. These valves may be closed intermittently for testing under administrative control.

C. C. Scram Insertion Times 1.1 The average scram time, 1. After refueling outage based on the and prior to operation de-energization of the above 30% power with scram pilot valve reactor pressure above solenoids of all 800 psig all control operable control rods in rods shall be subject to the reactor power scram-time measurements operation condition from the fully withdrawn shall be no greater position. The scram than: times for single rod scram testing shall be Drop-Out % Inserted Avg. Scram measured without of From Fully Insertion reliance on the control Position Withdrawn Time (sec) rod drive pumps.

46 4.51 0.358 2. During or following a 36 25.34 0.912 controlled shutdown of 26 46.18 1.468 the reactor, but not 06 87.84 2.686 more frequently than 16 weeks nor less The average of the scram frequently than 32 weeks insertion times for the intervals, 50% control three fastest control rod drives in each rods of all groups of quadrant of the reactor four control rods in a core shall be measured two by two array shall for scram times be no greater than: specified in Specification 3.3.C.

Drop-Out % Inserted Avg. Scram All control rod drives of From Fully Insertion shall have experienced Position Withdrawn Time (sec) scram-time measurements each year. Whenever 50%

46 4.51 0.379 of the control rod 36 25.34 0.967 drives scram times have 26 46.18 1.556 been measured, an 06 87.84 2.848 e,.aluation shall be made to provide reasonable assurance that proper control rod drives performance is being maintained. The results of measurements performed on the control rod drives shall be submitted in the start up test report.

Amendment No. -14, , -49, :74, 73 85

VYNPS BASES: 3.3 & 4.3 (Cont'd)

2. The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage of the primary coolant system. The design basis is given in Subsection 3.5.2 of the FSAR, and the design evaluation is given in Subsection 3.5.4. This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing.

3. In the course of performing normal startup and shutdown procedures, a pre-specified sequence for the withdrawal or insertion of control

-' rods is followed.- Control rod dropout accidents which might lead to significant core damage, cannot occur if this sequence of rod withdrawals or insertions is followed. The Rod Worth Minimizer restricts withdrawals and insertions to those listed in the pre-specified sequence and provides an additional check that the reactor operator is following prescribed sequence. Although beginning a reactor startup without having the RWM operable would entail unnecessary risk, continuing to withdraw rods if the RWM fails subsequently is acceptable if a second licensed operator verifies the withdrawal sequence. Continuing the startup increases core power, reduces the rod worth and reduces the consequences of dropping any rod. Withdrawal of rods for testing is permitted with the RWM inoperable, if the reactor is subcritical and all other rods are fully inserted. Above 20% power, the RWM is not needed since even with a single error an operator cannot withdraw a rod with sufficient worth, which if dropped, would result in anything but minor consequences.

4. Refer to the "General Electric Standard Application for Reactor Fuel (GESTAR II)," NEDE-24011-P-A, (the latest NRC-approved version will be listed in the COLR).

5. The Source Range Monitor (SRM) system provides a scram function in I noncoincident configuration.

visual indication of neutron level.

It does provide the operator with a The consequences of reactivity accidents are a function of the initial neutron flux. The requirement of at least three counts per second assures that any transient, should it occur, begins at or above the initial value of 10.8 of rated power used in the analyses of transients from cold conditions. One operable SRM channel is adequate to monitor the approach to criticality, therefore, two operable SRM's are specified for added conservatism.

6. , The Rod BLack Monitor (RBM) is designed to automatically prevent fuel mge inýt~vent of erroneous rod withdrawal from cations igh powe ensity during high Power level oper tign. During reactor ration with certai miting contr od patterns, the suwithdrawal of uconti signated sin control rod cou

! i reyst in a violatio stern the MCPR saf limit or the I which astic st amn limit. A iting control d pattern is a trnwhc results in t core being o: er mit (i . operat g on a limiting ue for APLHGRGXRGs, or MCPR. D Ing e 0'. uch patte~r ,it is judged, atft testing of t RBM systeW nor to withdrawal of suchprs will provide ed assurance that proper withdrawal does occur. It is e responsibility the Nucl ar Engineer to i tify these lim* ing patterns and designi.at rods eitherhen the patte are initially e lished o, s they develop due to the occurrence of inoperable control rods.

- be~cvfe0 Amendment No. &9,- 6-1, *0, 4-, "- 99 S, 1+/-1

VYNPC 3.4- LIMITING CONDITIONS FOR 4.4 SURVEILLANCE REQUIREMENTS OPERATION 3.4 REACTOR STANDBY LIQUID CONTROL 4.4 REACTOR STANDBY LIQUID CONTROL SYSTEM SYSTEM Applicability: Applicability:

Applies to the operating status Applies to the periodic testing of the Reactor Standby Liquid requirement for the Reactor Control System. Standby Liquid Control System.

Objective: Objective:

To assure the availability of an To verify the operability of the independent reactivity control Standby Liquid Control System.

mechanism.

Specification: Specification:

A. Normal Operation A. Normal Operation Except as specified in 3.4.B The Standby Liquid Control below, the Standby Liquid System shall be verified Control System shall be operable by:

operable when the reactor mode switch is in either the 1. Testing pumps and valves "Startup/Hot Standby" or in accordance with "Run" position, except to Specification 4.6.E. A allow testing of instrumentation associated with the reactor mode switch minimum flow rate of 35 gpm at<4 psig shall be veriied for I

I interlock functions each pump.

provided:

2. Verifying the continuity

1. Reactor coolant of the explosive charges temperature is less than at least monthly.

or equal to 2120 F; In addition, at least once

2. All control rods remain during each operating cycle, fully inserted in core the Standby Liquid Control cells containing one or System shall be verified more fuel assemblies; operable by:

and

3. Testing that the setting

3. No core alterations are of the pressure relief in progress. valves is between 1400 and 1490 psig.

4. Initiating one of the standby liquid control loops, excluding the primer chamber and inlet fitting, and verifying that a flow path from a pump to the reactor vessel is available.

Both loops shall be tested over the course of two operating cycles.

I Amendment No. 4-G-, 4-2-8-, 4-64, 175 92

VYNPS BASES:

3.4 & 4.4 REACTOR STANDBY LIQUID CONTROL SYSTEM t C6 A. Normal Operation The design objective of the Reactor Standby Liquid Control System it provide the capability of bringing the reactor from full power to a cold, xenon-free shutdown assuming that none of the withdrawn control rods can be inserted. To meet this objective, the Standby Liquid Control System is designed to inject a quantity of boron which produces a concentration of 800 ppm of natural boron in the reactor core in less than 138 minutes. An 800 ppm natural boron concentration in the reactor core is required to bring the reactor from full power to a 5%

Ak subcritical condition. An additional margin (25% of boron) is added for possible imperfect mixing of the chemical solution in the reactor water. A minimum quantity of 3850 gallons of solution having a 10.1%

natural sodium pentaborate concentration is required to meet this shutdown requirement.

The time requirement (138 minutes) for insertion of the boron solution was selected to override the rate of reactivity insertion due to cooldown of the reactor following the xenon poison peak. For a required minimum pumping rate of 35 gallons per minute, the maximum net storage volume of the boron solution is established as 4830 gallons.

In addition to its original design basis, the Standby Liquid Control System also satisfies the requirements of IOCFR50.62(c)(4) on anticipated transients without scram (ATWS) by using enriched boron. The ATWS rule adds hot shutdown and neutron absorber (i.e., boron-10) injection rate requirements that exceed the original Standby Liquid Control System design basis. However, changes to the Standby Liquid Control System as a result of the ATWS rule have not invalidated the original design basis.

With the reactor mode switch in the 'Run" or "Startup/Hot Standby, position, shutdown capability is required. With the mode switch in "Shutdown,' control rods are not able to be withdrawn since a control rod block is applied. This provides adequate controls to ensure that the reactor remains subcritical.

With the mode switch in "Refuel,' only a single control rod can be withdrawn from a core cell containing fuel assemblies. Determination of adequate shutdown margin by Specification 3.3.A ensures that the reactor will not become critical. Therefore, the Standby Liquid Control System is not required to be operable when only a single control rod can be withdrawn.

Pump operability testing (by recirculating demineralized water to the test tank)in accordance with Specification 4.6.E is adequate to detect if failures have occurred. Flow, relief valve, circuitry, and trigger assembly testing at the prescribed intervals assures a high reliability of system operation capability. Recirculation of the borated solution is done during each operating ?ycle to ensure one suction line from the boron tank is clear. In additi at least once during each operating cycle, one of the standby liquid contro loops will be initiated to verify that a flow path from a pump to the reacrr vessel is available by pumping demineralized water into the reactor ves el.

B. ;eration With inoperable Components proper operation of the system. If one pumping circuit is found to Amendment No. -I-0, -144, a, 175 1 97 The maximum SLCS pump discharge pressure during the limiting ATWS event is 1320 psig. This value is based on a peak reactor vessel lower plenum pressure of 1290 psia that occurs during the limiting ATWS event at the time of SLCS initiation, i.e., 120 seconds into the event. There is adequate margin to prevent the SLCS relief valve from lifting. With a nominal SLCS relief valve setpoint of 1400 psig, there is a margin of 80 psi between the peak SLCS pump discharge pressure and the relief valve nominal setpoint.

VYNPS 4.6 SURVEILLANCE REQUIREMENTS 3.6 LIMITING CONDITIONS FOR OPERATION D. Safety and Relief Valves D. Safety and Relief Valves During reactor power 1. Operability testing of 1.

Safety and Relief Valves operating conditions and shall be in accordance whenever the reactor with Specification 4.6.E.

coolant pressure is point of the The lift greater than 150 psig and

e *ature greater than safety and relief valves shall be set as specified Z/_ 350 0 F, be-*i safety valves in Specification 2.2.B.

/* 0 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 350OF within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

E. Structural Integrity and E. Structural Integrity and Operability Testing Operability Testing

1. Inservice inspection of The structural integrity and safety-related components the operability of the shall be performed in safety-related systems and accordance with components shall be Section XI of the ASME maintained at the level Boiler and Pressure required by the original Vessel Code and acceptance standards applicable Addenda as throughout the life of the required by 10 CFR 50, plant. Section 50.55a(g), except where specific written relief has been granted by the NRC.

Inservice inspection of piping, identified in NRC Generic Letter 88-01, shall be performed in accordance with the staff positions on schedule, methods, and personnel and sample expansion included in the Generic Letter or in accordance with alternte measures approved by NRC Staff.

1,-7-f 4-7, 196 120 Amendment No. 4-3, 48, 5, 99, 1-2, 4-3-9,-

VYNPS 3.11 LIMITING CONDITIONS FOR 4.11 SURVEILLANCE REQUIREMENTS OPERATION 4.11 REACTOR FUEL ASSEMBLIES 3.11 REACTOR FUEL ASSEMBLIES Applicability:

Applicability:

The Surveillance Requirements The Limiting Conditions for apply to the parameters which Operation associated with the monitor the fuel rod operating fuel rods apply to these conditions.

parameters which monitor the fuel rod operating conditions.

Objective:

The Objective of the The Objective of the Limiting Surveillance Requirements is to Conditions for Operation is to specify the type and frequency assure the performance of the of surveillance to be applied to fuel rods. the fuel rods.

S cifications:

Specifications: landflow A. Average Planar Linear Heat ,power' A. Avera e Planar L near Heat Generation Rate (APLHGR) and flow Generation Rate AAPLHGR)

During operation at >25t The APLHGR for each type o Rated Thermal Po er,--the fuel as a function of I as a for each APLHGRfunction ype of fuel o average average planar exposure shall be determined once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after >25%

planar exposure shall not Rated Thermal Power and exceed the limiting values daily during operation at provided in the Core >25% Rated Thermal Power Operating Limits Report. thereafter.

For single recirculation loop operation, the limiting values shall be the values provided in the Core Operating Limits Report listed under the heading

",Single Loop Operation." If at any time during operation at >25% Rated Thermal Power it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, APLHGR(s) shall be returned to within prescribed limits within two (2) hours; otherwise, the reactor shall be brought to

<25% Rated Thermal Power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits-88 224 S-9, 4-, 6+-. 9, 94, .- 8-4, 44&, 144, 1 Amendment No.

VYNPS LIMITING CONDITIONS FOR 4.11 SURVEILLANCE REQUIREMENTS 3.11 OPERATION Power Ratio C. Minimum Critical Power Ratio C. Minimum Critical (MCPR) (MCPR)

MCPR shall be determined /

1. During operation at >25%

Rated Thermal Power the once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months aftex/*

MCPR operating value >25% Rated Thermal Power, xar shall be equal to or daily during operation at greater than the MCPR >25% Rated Thermal Power limits provided in the Ehereafte and f lowingl-F Core Operating Limits ny an p-in wer le 1 or

,stri io hat w d Report. -For single h a recirculation loop caus ope ion operation, the MCPR iti cont ro patt n as crib in e b s Limits at rated flow are also provided in the for Specification Core Operating p'Fr coreLimits

+/-w S~~Report.

_* - r7,_he Op *e i**R n Lim' I~~1SAIJethe ab e va e SReport.If at any time during operation at >25%

Rated Thermal Power it is determined by normal surveillance that the limiting value for MCPR is being exceeded, MCPR(s) shall be returned to within the prescribed limits within two (2) hours; otherwise, the reactor power shall be brought to <25% Rated Thermal Power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

226 Amendment No. 4-,, -4,-4, t , -t-, loo

VYNPS BASES:

3. 11 FUEL RODS A. Average Planar Linear Heat Generation Rate (APLHGR)

Refer to the appropriate topical reports listed in for analyses methods. Specification 6.6.C (Note: All exposure increments in this Technical Specification

.section are expressed in terms of megawatt-days per short ton.)

The MAPLHGR reduction factor for single recirculation is based on the assumption that the coastdown flow fromloop operation the recirculation loop would not be available during a postulatedunbroken break in the active recirculation loop. See Core Operating large j NS E _7 Report for the cycle-specific reduction factor.

Limits B. Linear Heat Generation Rate (LHGR)

Refer to the appropriate topical reports listed in Specification for analyses methods. 6.6.C C. Minimum Critical Power Ratio (MCPR)

Operating Limit MCPR

1. The MCPR operating limit is a cycle-dependent parameter be determined for a number of different combinations which can modes, initial conditions, and cycle exposures in order of operating reasonable assurance against exceeding the Fuel Cladding to provide Safety Limit (FCISL) for potential abnormal occurrences. Integrity operating limits are justified by the analyses, the The MCPR which are presented in the current cycle's Supplemental results of Licensing Report. Reload Refer to the appropriate topical reports listed in Specification 6.6.C for analysis methods. The increase in MCPR operating limits for single loop operation accounts for core flow measurement and TIP reading uncertainties. increased J SR -2 Power and flow dependent LHGR limits are implemented using LHGRFAC multipliers on the standard LHGR limits.

The LHGRFAC multipliers are identical to the MAPFAC multipliers.

Amendment No. -1, 4-4, -7, 4;a, jS, 14, G, 47F4, B,-9 -, 171 227

l INSERT !

Flow dependent MAPLHGR limits, MAPFAC(F), were designed to assure adherence to all fuel thermal-mechanical design bases. The same transient events used to support the MCPR(F) operating limits were analyzed, and the resulting overpowers were statistically evaluated as a function of the initial and maximum core flow. From the bounding overpowers, the MAPFAC(F) limits were derived such that the peak transient LHGR would not exceed fuel mechanical limits. The flow-dependent MAPLHGR limits are cycle-independent and are specified in terms of multipliers, MAPFAC(F), to be applied to the rated MAPLHGR values.

"Power-dependentMAPLHGR limits, expressed in terms of a MAPLHGR multiplier, MAPFAC(P), are substituted to assure adherence to the fuel thermal-mechanical design bases. Both incipient centerline melting of fuel and plastic strain of the cladding are considered in determining the power dependent MAPLHGR limit. Generally, the limiting criterion is incipient centerline melting. The power-dependent MAPFAC(P) multipliers were generated using the same database as used to determine the MCPR multiplier (Kp). Appropriate MAPFAC(P) multipliers are selected based on plant specific transient analyses with suitable margin to assure applicability to future reloads.

These limits are derived to assure that the peak transient MAPLHGR for any transient is not increased above the fuel design bases values.

INSERT 2 Flow-dependent MCPR limits, MCPR(F), are necessary to assure that the Safety Limit MCPR (SLMCPR) is not violated during recirculation flow increase events. The design basis flow increase event is a slow (maximum two pump runout rate of 1% /dsecond) recirculation flow increase event which is not terminated by scram, but which stabilizes at a new core power corresponding to the maximum possible core flow. Flow runout events were analyzed along a constant xenon, constant feedwater temperature flow control line assuming a quasi steady-state plant heat balance. The ARTS-based MCPR(F) limit is specified as an absolute value and is cycle-independent. The operating limit is based on a maximum core flow limiter setting of 109.5% in the Recirculation Flow Control System.

Above the power at which the scram is bypassed (Pbypass), bounding power-dependent trend functions have been developed. This trend function, Kp, is used as multiplier to the rated MCPR operating limits to obtain the power-dependent MCPR limits, MCPR(P).

Below the power at which the scram is automatically bypassed (Below Pbypass), the MCPR(P) limits are actual absolute Operating Limit MCPR (OLMCPR) values, rather than multipliers on the rated power OLMCPR.

VYNPS BASES:

4.11 FUEL RODS A. The APLHGR, LHGR and MCPR shall be checked daily when operating at

>25% Rated Thermal Power to determine if fuel burnup, or control rod movement has caused changes in power distribution. Since changes due to burnup are slow, and only a few control rods are removed daily, a daily check of power distribution is adequate. For a limiting value to occur below 25% of rated thermal power, an unreasonably large peaking factor would be required, which is not the case for operating control rod sequences. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after thermal power >25%

Rated Thermal Power is achieved is acceptable given the large inherent margin to operating limits at low power levels.

B. At certain times during plant startups and power changes the plant technical staff may determine that surveillance of APLHGR, LHGR and/or MCPR is necessary more frequently than daily. Because the necessity for such an augmented surveillance program is a function of a number of interrelated parameters, a reasonable program can only be determined on a case-by-case basis by the plant technical staff. The check of APLHGR, LHGR and MCPR will normally be done using the plant process computer. In the event that the computer is unavailable, the check will consist of either a manual calculation or a comparison of existing core conditions to those existing at the time of a previous check to determine if a significant change has occurred.

If a reactor power distribution limit is exceeded, an assumption regarding an initial condition of the DBA analysis, transient analyses, or the fuel design analysis may not be met. Therefore, prompt action should be taken to restore the APLHGR, LHGR or MCPR to within the required limits such that the plant operates within analyzed conditions and within design limits of the fuel rods. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months completion time is sufficient to restore the APLHGR, LHGR, or MCPR to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the APLHGR, LHGR, or MCPR out of specification.

C. Minimum Critical Power Ratio (MCPR) - Surveillance Requirement At core thermal power levels less than or equal to 25%, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% thermal power level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR above 25% rated thermal power is sufficient since Isign power distribution shifts are very slow hen thre-have not been cant pow 1aculatipaCPR whe or control od chang-. .*. l* rpr*irement for l ensures hat MCP <ill

imiti Entro~ ro° be eown fo/<win-chae d tte in s ap er ached

-owe sha (regardess of m nitude) kat cou d pl ce-op rationrat-*s' Amendment No. *1, 188

VYNPS include a tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man-rem exposure according to work and job functions, 1/ e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.

The dose assignment to various duty functions may be estimates based on Self-Reading Dosimeter (SRD), TLD or film badge measurement.

Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work functions.

B. Monthly Operating Reports Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis no later than the fifteenth of each month following the calendar month covered by the report.

These reports shall include a narrative summary of operating experience during the report period which describes the operation of the facility.

C. Core Operating Limits Report The core operating limits shall be established and documented in the Core Operating Limits Report (COLR) before each reload cycle or any remaining part of a reload cycle for the following:

I. The Average Planar Linear Heat Generation Rates (APLHGR) for Specifications 3.11.A and 3.6.G.la,

2. The Kf core flow adjustment factor for Specification 3.11.C__>,

a /<The Minimum Critical Power Ratio (MCPR) for e .LJ .. ~1 ~

n~

A~LLU~ CJ.. I Sf The Linear Heat Generation Rates (LHGR) for I 3~j Specifications 2.l.A.la and 3.11.B, and The Power/Flow Exclusion Region for Specifi cations 3.6.J.la and 3.6.J.lb.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

Report, E. E. Pilat, "Methods for the Analysis of Boiling Water Reactors Lattice Physics," YAEC-1232, December 1980 (Approved by NRC SER, dated September 15, 1982).

1/ This tabulation supplements the requirements of 20.2206 of 10 CFR Part 20.

Amendment No. 42-, 446-,!-4&, 4--7, 211 259

Docket No. 50-271 BVY 03-23 Attachment 4 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 257 Implementation of ARTS/MELLLA at Vermont Yankee Retyped Technical Specification Pages

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.

Current Page New Page 6 6 7 7 11 11 14 14 18 18 20 20 21 21 22 22 24 24 33a 33a 52 52 77 77 84 84 85 85 90 90 92 92 97 97 98 98 120 120 224 224 226 226 227 227

--- 227a 228 228 259 259

VYNPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Applicability: Applicability:

Applies to the interrelated Applies to trip setting of the variable associated with fuel instruments and devices which are thermal behavior. provided to prevent the nuclear system safety limits from being exceeded.

Objective: Objective:

To establish limits below which To define the level of the process the integrity of the fuel variable at which automatic cladding is preserved. protective action is initiated.

Specification:

A. Trip Settings A. Bundle Safety Limit (Reactor Pressure >800 psia and Core The limiting safety system Flow >10% of Rated) trip settings shall be as specified below:

When the reactor pressure is

>800 psia and the core flow is 1. Neutron Flux Trip Settings greater than 10% of rated:

a. APRM Flux Scram Trip

1. A Minimum Critical Power Setting (Run Mode)

Ratio (MCPR) of less than 1.10 (1.12 for Single Loop When the mode switch Operation) shall constitute is in the RUN violation of the Fuel position, the APRM Cladding Integrity Safety flux scram trip Limit (FCISL). setting shall be as shown on Figure 2.1.1 and shall be:

Two loop operation:

S* 0.4W+ 64.4% for 0% < W 5 31.1%

S5 1.28W+ 37.0% for 31.1% < W 5 54.0%

S5 0.66W+ 70.5% for 54.0% < W

75.0%

With a maximum of 120.0% power for W >

75.0%

Single loop operation:

S5 0.4W+ 61.2% for 0% < W

39.1%

S5 1.28W+ 26.8% for 39.1% < W

61.9%

S5 0.66W+ 65.2% for 61.9% < W

83.0%

With a maximum of 120.0% power for W >

83.0%

where:

S = setting in percent of rated thermal power (1593 MWt)

-1, 4-, -64, ý -94,,4, 9 45, !59, 176 6 Amendment No.

VYNPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING B. Core Thermal Power Limit W = percent rated (Reactor Pressure < 800 psia two loop drive or Core Flow < 10% of Rated) flow where When the reactor pressure is 100% rated

<800 psia or core flow <10% drive flow is of rated, the core thermal that flow power shall not exceed 25% of equivalent to rated thermal power. 48 x 106 lbs/hr core C. Power Transient flow To ensure that the safety In the event of limit established in operation at >25% Rated Specification 1.1A Thermal Power the APRM and 1.1B is not exceeded, gain shall be equal to each required scram shall or greater than 1.0.

be initiated by its expected scram signal. The safety limit shall be assumed to be exceeded when scram is accomplished by means other than the expected scram signal.

D. Whenever the reactor is shutdown with irradiated fuel in the reactor vessel, the water level shall not be less than 12 inches above the top of the enriched fuel when it is seated in the core.

4, 39, 44-, 4-1, 64, -64, 94, 18,

. 7 Amendment No.

VYNPS FIGURE 2.1.1 APRM FLOW REFERENCE SCRAM SETTING 130 120 Tolo 120 -froperation 110 100 90 Single loop operation S 80 LU

-J L

z 60 0

i M

z 40 30 APRM Flow Biased Scram 20 Setpoints shall be less than or equal to values shown on the graph.

10- 11111.

(II 04 0 20 40 60 80 100 120 RECIRCULATION FLOW (% RATED)

Amendment No. 4-, S4, 18ý;, 2-, 11

VYNPS BASES:

2.1 FUEL CLADDING INTEGRITY A. Trip Settings The bases for individual trip settings are discussed in the following paragraphs.

1. Neutron Flux Trip Settings

a. APRM Flux Scram Trip Setting (Run Mode)

The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (1593 MWt). Because fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel. Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting.

Analyses are performed to demonstrate that the APRM flux scram over the range of settings from a maximum of 120% to the minimum flow biased setting provide protection from the fuel safety limit for all abnormal operational transients including those that may result in a thermal hydraulic instability.

An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached. The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation. Reducing this operating margin would increase the frequency of spurious scrams which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APRM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity Safety Limit yet allows operating margin that reduces the possibility of unnecessary scrams. The relationship between recirculation drive flow and reactor core flow is non-linear at low core flows. Therefore, separate APRM flow biased scram trip setting equations are provided for low core flows.

The scram trip is set to ensure acceptable transient response. For single recirculation loop operation, the APRM flux scram trip setting is reduced in accordance with the analysis presented in NEDO-30060, February 1983. This adjustment accounts for the difference between the single loop and two loop drive flow at the same core flow, and ensures that the margin of safety is not reduced during single loop operation. The single loop operation equations are based on a bounding (maximum) difference between two loop and single loop drive flow at the same core flow of 8%.

Analyses of the limiting transients show that no scram adjustment is required to assure fuel cladding integrity when the transient is initiated from the operating limit MCPR defined in the Core Operating Limits Report.

, 4-, &1, 94, 4, 14 Amendment No. 44, VYNPS 1.2 SAFETY LIMIT 2.2 LIMITING SAFETY SYSTEM SETTING 1.2 REACTOR COOLANT SYSTEM 2.2 REACTOR COOLANT SYSTEM Applicability: Applicability:

Applies to limits on reactor Applies to trip settings for coolant system pressure. controlling reactor system pressure.

Objective: Objective:

To establish a limit below which To provide for protective action the integrity of the reactor in the event that the principal coolant system is not threatened process variable approaches a due to an overpressure safety limit.

condition.

Specification: Specification:

The reactor coolant system pressure shall not exceed 1335 psig at any time when irradiated fuel is present in the reactor vessel.

A. Reactor coolant high pressure scram shall be less than or equal to 1055 psig.

B. Primary system relief and safety valve settings shall be as specified in Table 2.2.1.

TABLE 2.2.1 Primary System Relief and Safety Valve Settings Lift Number and Type Setting(I) of Valve(s) I 1 safety relief valve 1080 psig 2 safety relief valves 1090 psig 1 safety relief valve 1100 psig 3 safety valves 1240 psig Note:

(1) As-left setpoint tolerance +/-1%.

As-found setpoint tolerance

+/-3%.

18 Amendment No. 14-,64E

VYNPS 3.1 LIMITING CONDITIONS FOR 4.1 SURVEILLANCE REQUIREMENTS OPERATION 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM Applicability: Applicability:

Applies to the operability of Applies to the surveillance of plant instrumentation and the plant instrumentation and control systems required for control systems required for reactor safety. reactor safety.

Objective: Objective:

To specify the limits imposed on To specify the type and plant operation by those frequency of surveillance to be instrument and control systems applied to those instrument and required for reactor safety. control systems required for reactor safety.

Specification: Specification:

A. Plant operation at any power A. Instrumentation systems level shall be permitted in shall be functionally tested accordance with Table 3.1.1. and calibrated as indicated The system response time in Tables 4.1.1 and 4.1.2, from the opening of the respectively.

sensor contact up to and including the opening of the scram solenoid relay shall not exceed 50 milliseconds.

B. Deleted. B. Deleted.

Amendment No. 41, 164, 188,

-1Q*2 20

VYNPS TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT REQUIREMENTS Required ACTIONS When Modes in Which Minimum Minimum Functions Must be Number Conditions Operating Operating For Instrument Operation Channels Per Are Not Refuel Startup Run Trip System Satisfied Trip Function Trip Settings (3)

(1) (12) (2)

X X 1 A

1. Mode Switch in X Shutdown (5A-Sl)

X X X 1 A

2. Manual Scram (5A-S3A/B)

3. IRM (7-41(A-F))

X X A High Flux <120/125 2 X X 2 A INOP

4. APRM (APRM A-F)

High Flux Two loop operation: (4) X 2 A or B (flow bias) S5 0.4W+ 64.4% for 0% < W *31.1%

S5 1.28W+ 37.0% for 31.1% < W *54.0%

S: 0.66W+ 70.5% for 54.0% < W S75.0%

With a maximum of 120.0% power for W > 75.0%

Single loop operation: (4)

S! 0.4W+ 61.2% for 0% < W 39.1%

S! 1.28W+ 26.8% for 39.1% < W 61.9%

S! 0.66W+ 65.2% for 61.9% < W 83.0%

With a maximum of 120.0% power for W > 83.0%

High Flux X X 2 A (reduced)

X 2(5) A or B INOP X X X X 2 A

5. High Reactor <1055 psig Pressure (PT-2-3-55 (A-D)

(M))

-4, 4-, 64, .4&, 4, *4,

-4, -94, 94, S4I6, -4*G, -378- 21 Amendment No.

VYNPS TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT REQUIREMENTS (Continued)

Required ACTIONS When Modes in Which Functions Must Minimum Number Minimum be Operating Operating Conditions For Instrument Operation Channels Per Are Not Trip Function Trip Settings Refuel (1) Startup (12) Run Trip System (2) Satisfied (3)

6. High Drywell <2.5 psig x x x 2 A Pressure (PT-5-12(A-D) (M))

7. Reactor Low (6) >127.0 inches x x x 2 A Water Level (LT-2-3-57A/B(M))

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

8. Scram Discharge <21 gallons x x x 2 A Volume High Level (per volume)

(LT-3-231 (A-H) (M))

9. Deleted

10. Main steamline <10% valve x 4 A or C isolation valve closure closure (POS-2-80A-Al,Bl POS-2-86A-Al,B1 POS-2-80B-AI,B2 POS-2-86B-Al,B2 POS-2-80C-A2,Bl POS-2-86C-A2,Bl POS-2-80D-A2,B2 POS-2-86D-A2,B2)

11. Turbine control (9) (10) x 2 A or D valve fast closure (PS- (37-40))

12. Turbine stop valve <10% valve(10) x 2 A or D closure closure (SVOS (1-4))

Amendment No. 4-64, *4*-, 42 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.

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.

-,4*,2 , 6-, :74, i4, !)4, 94, 4,64, , 3:86, *-9, 24 Amendment No.

VYNPS BASES: 4.1 (Cont'd)

LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local flux profile for appropriate representative input to the APRM System. The 2,000 megawatt-days per short ton (MWD/T) frequency is based on operating experience with LPRM sensitivity changes, and that the resulting nodal power uncertainty, combined with other identified uncertainties, remains less than the total uncertainty (i.e., 8.7%) allowed by the GETAB safety limit analysis.

&4&, 19 4-4, 33a Amendment No.

VYNPS TABLE 3.2.5 NOTES

1. Deleted.

2. Deleted.

3. Deleted.

4. Deleted.

5. "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. Refer to the Core Operating Limits Report for acceptable values for N. 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 and = 8% for single loop operation.

6. Not used.

7. The trip may be bypassed when the reactor power is <30% of Rated Thermal Power. An RBM channel will be considered inoperable if there are less than half the total number of normal inputs from any LPRM level.

8. With the number of operable channels less than the required number, place the inoperable channel in the tripped condition within one hour.

9. With one or two RBM channels inoperable:

a. Deleted.

b. If one RBM channel is inoperable, restore the inoperable channel to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; and

c. If the required action and associated completion time of Note 9.b above is not met, or if two RBM channelswithin are inoperable, place one RBM channel in the tripped condition the next hour.

10. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into associated Limiting Conditions for Operation and required action notes 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 Control Rod Block initiation capability.

11. Deleted.

12. Required to be operable when the reactor mode switch is in the shutdown position.

13. With one or more Reactor Mode Switch - Shutdown Position channels inoperable, immediately suspend control rod withdrawal and immediately initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies.

q-G, *, 94, -l,-6, --a4, 52 Amendment No. 64, -,

VYNPS BASES: 3.2 (Cont'd) control and/or bypass valves to open, resulting in a rapid depressurization and cooldown of the reactor vessel. The 800 psig trip setpoint limits the depressurization such that no excessive vessel thermal stress occurs as a result of a pressure regulator malfunction. This setpoint was selected far enough below normal main steam line pressures to avoid spurious primary containment isolations.

Low condenser vacuum has been added as a trip of the Group 1 isolation valves to prevent release of radioactive gases from the primary coolant through condenser. The setpoint of 12 inches of mercury absolute was selected to provide sufficient margin to assure retention capability in the condenser when gas flow is stopped and sufficient margin below normal operating values.

The HPCI and/or RCIC high flow and temperature instrumentation is provided to detect a break in the HPCI and/or'RCIC piping. The HPCI and RCIC steam supply pressure instrumentation is provided to isolate the systems when pressure may be too low to continue operation. These isolations are for equipment protection. However, they also provide a diverse signal to indicate a possible system break. These instruments are included in Technical Specifications because of the potential for possible system initiation failure if not properly tested. Tripping of this instrumentation results in actuation of HPCI and/or RCIC isolation valves, i.e., Group 6 valves. A time delay has been incorporated into the RCIC steam flow trip logic to prevent the system from inadvertently isolating due to pressure spikes which may occur on startup. The trip settings are such that core uncovering is prevented and fission product release is within limits.

The instrumentation which initiates ECCS action is arranged in a dual channel system. Permanently installed circuits and equipment may be used to trip instrument channels. In the nonfail safe systems which require energizing the circuitry, tripping an instrument channel may take the form of providing the required relay function by use of permanently installed circuits. This is accomplished in some cases by closing logic circuits with the aid of the permanently installed test jacks or other circuitry which would be installed for this purpose.

The Rod Block Monitor (RBM) control rod block functions are no longer credited in the Rod Withdrawal Error (RWE) Analysis. The RBM setpoints are based on providing operational flexibility in the MELLLA region.

For single recirculation loop operation, the RBM trip setting is reduced in accordance with the analysis presented in NEDO-30060, February 1983. This adjustment accounts for the difference between the single loop and two loop drive flow at the same core flow, and ensures that the margin of safety is not reduced during single loop operation.

During hot shutdown, cold shutdown, and refueling when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical with sufficient shutdown margin; therefore, no positive reactivity insertion events are analyzed. The Reactor Mode Switch-Shutdown Position control rod withdrawal block, required to be operable with the mode switch in the shutdown position, ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis. Two channels are required to be Amendment No. -,, 1 .4, &4, .94, -8, 2-2,4-1 77

VYNPS 4.3 SURVEILLANCE REQUIREMENTS 3.3 LIMITING CONDITIONS FOR OPERATION pressure, are fully inserted, no more than two rods may be moved.

(c) Out-of-sequence control rods in each distinct RWM group shall be selected and the annunciator of the selection errors verified.

(d) An out-of-sequence control rod shall be withdrawn no more than three notches and the rod block function verified.

4. The control rod pattern

4. Control rod patterns and and sequence of the sequence of withdrawal or insertion withdrawal or insertion shall be verified to shall be established comply with such that the rod drop Specification 3.3.B.4.

accident limit of 280 cal/g is not exceeded.

5. Prior to control rod

5. Control rods shall not withdrawal for startup be withdrawn for startup or during refueling, or refueling unless at verification shall be least two source range made that at least two channels have an source range channels observed count rate have an observed count greater than or equal to rate of at least three three counts per second. counts per second.

I 6. Deleted. 6. Deleted.

I 84 Amendment No. -24, 6-

VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION

7. The scram discharge volume drain and vent valves shall be verified open at least once per month. These valves may be closed intermittently for testing under administrative control.

C. Scram Insertion Times C. Scram Insertion Times 1.1 The average scram time, 1. After refueling outage based on the and prior to operation de-energization of the above 30% power with scram pilot valve reactor pressure above solenoids of all 800 psig all control operable control rods in rods shall be subject to the reactor power scram-time measurements operation condition from the fully withdrawn shall be no greater position. The scram than: times for single rod scram testing shall be Drop-Out % Inserted Avg. Scram measured without of From Fully Insertion reliance on the control Position Withdrawn Time (sec) rod drive pumps.

46 4.51 0.358 2. During or following a 36 25.34 0.912 controlled shutdown of 26 46.18 1.468 the reactor, but not 06 87.84 2.686 more frequently than 16 weeks nor less The average of the scram frequently than 32 weeks insertion times for the intervals, 50% control three fastest control rod drives in each rods of all groups of quadrant of the reactor four control rods in a core shall be measured two by two array shall for scram times be no greater than: specified in Specification 3.3.C.

Drop-Out % Inserted Avg. Scram All control rod drives of From Fully Insertion shall have experienced Position Withdrawn Time (sec) scram-time measurements each year. Whenever 50%

46 4.51 0.379 of the control rod 36 25.34 0.967 drives scram times have 26 46.18 1.556 been measured, an 06 87.84 2.848 evaluation shall be made to provide reasonable assurance that proper control rod drives performance is being maintained. The results of measurements performed on the control rod drives shall be submitted in the start up test report.

Amendment No. 4-4, 2 23-,*&0, 448 85

VYNPS BASES: 3.3 & 4.3 (Cont'd)

2. The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage of the primary coolant system. The design basis is given in Subsection 3.5.2 of the FSAR, and the design evaluation is given in Subsection 3.5.4. This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing.

3. In the course of performing normal startup and shutdown procedures, a pre-specified sequence for the withdrawal or insertion of control rods is followed. Control rod dropout accidents which might lead to significant core damage, cannot occur if this sequence of rod withdrawals or insertions is followed. The Rod Worth Minimizer restricts withdrawals and insertions to those listed in the pre-specified sequence and provides an additional check that the reactor operator is following prescribed sequence. Although beginning a reactor startup without having the RWM operable would entail unnecessary risk, continuing to withdraw rods if the RWM fails subsequently is acceptable if a second licensed operator verifies the withdrawal sequence. Continuing the startup increases core power, reduces the rod worth and reduces the consequences of dropping any rod. Withdrawal of rods for testing is permitted with the RWM inoperable, if the reactor is subcritical and all other rods are fully inserted. Above 20% power, the RWM is not needed since even with a single error an operator cannot withdraw a rod with sufficient worth, which if dropped, would result in anything but minor consequences.

4. Refer to the "General Electric Standard Application for Reactor Fuel (GESTAR II)," NEDE-24011-P-A, (the latest NRC-approved version will be listed in the COLR).

5. The Source Range Monitor (SRM) system provides a scram function in noncoincident configuration. It does provide the operator with a visual indication of neutron level. The consequences of reactivity accidents are a function of the initial neutron flux. The requirement of at least three counts per second assures that any transient, should it occur, begins at or above the initial value of 10-8 of rated power used in the analyses of transients from cold conditions. One operable SRM channel is adequate to monitor the approach to criticality, therefore, two operable SRM's are specified for added conservatism.

6. Deleted.

5, -9, 6-1, =G, 164, BVY 99 55, 219 90 Amendment No.

VYNPS 3.4 LIMITING CONDITIONS FOR 4.4 SURVEILLANCE REQUIREMENTS OPERATION 3.4 REACTOR STANDBY LIQUID CONTROL 4.4 REACTOR STANDBY LIQUID CONTROL SYSTEM SYSTEM Applicability: Applicability:

Applies to the operating status Applies to the periodic testing of the Reactor Standby Liquid requirement for the Reactor Control System. Standby Liquid Control System.

Objective: Objective:

To assure the availability of an To verify the operability of the independent reactivity control Standby Liquid Control System.

mechanism.

Specification: Specification:

A. Normal Operation A. Normal Operation Except as specified in 3.4.B The Standby Liquid Control below, the Standby Liquid System shall be verified Control System shall be operable by:

operable when the reactor mode switch is in either the 1. Testing pumps and valves "Startup/Hot Standby" or in accordance with "Run" position, except to Specification 4.6.E. A allow testing of minimum flow rate of instrumentation associated 35 gpm at 1320 psig with the reactor mode switch shall be verified for interlock functions each pump.

provided:

2. Verifying the continuity

1. Reactor coolant of the explosive charges temperature is less than at least monthly.

or equal to 212° F; In addition, at least once

2. All control rods remain during each operating cycle, fully inserted in core the Standby Liquid Control cells containing one or System shall be verified more fuel assemblies; operable by:

and

3. Testing that the setting

3. No core alterations are of the pressure relief in progress. valves is between 1400 and 1490 psig.

4. Initiating one of the standby liquid control loops, excluding the primer chamber and inlet fitting, and verifying that a flow path from a pump to the reactor vessel is available.

Both loops shall be tested over the course of two operating cycles.

Amendment No. -02, , 464, 4-7-69 92

VYNPS BASES:

3.4 & 4.4 REACTOR STANDBY LIQUID CONTROL SYSTEM A. Normal Operation The design objective of the Reactor Standby Liquid Control System (SLCS) is to provide the capability of bringing the reactor from full power to a cold, xenon-free shutdown assuming that none of the withdrawn control rods can be inserted. To meet this objective, the Standby Liquid Control System is designed to inject a quantity of boron which produces a concentration of 800 ppm of natural boron in the reactor core in less than 138 minutes. An 800 ppm natural boron concentration in the reactor core is required to bring the reactor from full power to a 5% Ak subcritical condition. An additional margin (25%

of boron) is added for possible imperfect mixing of the chemical solution in the reactor water. A minimum quantity of 3850 gallons of solution having a 10.1% natural sodium pentaborate concentration is required to meet this shutdown requirement.

The time requirement (138 minutes) for insertion of the boron solution was selected to override the rate of reactivity insertion due to cooldown of the reactor following the xenon poison peak. For a required minimum pumping rate of 35 gallons per minute, the maximum net storage volume of the boron solution is established as 4830 gallons.

In addition to its original design basis, the Standby Liquid Control System also satisfies the requirements of IOCFR50.62(c) (4) on anticipated transients without scram (ATWS) by using enriched boron.

The ATWS rule adds hot shutdown and neutron absorber (i.e., boron-10) injection rate requirements that exceed the original Standby Liquid Control System design basis. However, changes to the Standby Liquid Control System as a result of the ATWS rule have not invalidated the original design basis.

With the reactor mode switch in the "Run" or "Startup/Hot Standby" position, shutdown capability is required. With the mode switch in "Shutdown," control rods are not able to be withdrawn since a control rod block is applied. This provides adequate controls to ensure that the reactor remains subcritical. With the mode switch in "Refuel,"

only a single control rod can be withdrawn from a core cell containing fuel assemblies. Determination of adequate shutdown margin by Specification 3.3.A ensures that the reactor will not become critical.

Therefore, the Standby Liquid Control System is not required to be operable when only a single control rod can be withdrawn.

Pump operability testing (by recirculating demineralized water to the test tank) in accordance with Specification 4.6.E is adequate to detect if failures have occurred. Flow, relief valve, circuitry, and trigger assembly testing at the prescribed intervals assures a high reliability of system operation capability. The maximum SLCS pump discharge pressure during the limiting ATWS event is 1320 psig. This value is based on a peak reactor vessel lower plenum pressure of 1290 psia that occurs during the limiting ATWS event at the time of SLCS initiation, i.e., 120 seconds into the event. There is adequate margin to prevent the SLCS relief valve from lifting. With a nominal SLCS relief valve setpoint of 1400 psig, there is a margin of 80 psi between the peak SLCS pump discharge pressure and the relief valve nominal setpoint.

Recirculation of the borated solution is done during each operating cycle to ensure one suction line from the boron tank is clear. In addition, at least once during each operating cycle, one of the standby liquid control loops will be initiated to verify that a flow path from a pump to the reactor vessel is available by pumping demineralized water into the reactor vessel.

Amendment No. 4-1, -I4, &, 175 97

VYNPS BASES: 3.4 & 4.4 (Cont'd)

B. Operation With Inoperable Components Only one of the two standby liquid control pumping circuits is needed for proper operation of the system. If one pumping circuit is found to and be inoperable, there is no immediate threat to shutdown capability, reactor operation may continue while repairs are being made. Assurance from the that the system will perform its intended function is obtained with ASME results of the pump and valve testing performed in accordance Section XI requirements C. Standby Liquid Control System Tank - Borated Solution of The solution saturation temperature varies with the concentration the sodium pentaborate. The solution shall be kept at least 10°F above to guard against boron precipitation. The 10°F saturation temperature margin is included in Figure 3.4.2. Temperature and liquid level alarms for the system are annunciated in the Control Room.

vary Once the solution has been made up, boron concentration will not unless more boron or water is added. Level indication and alarm a

indicate whether the solution volume has changed which might indicate Considering these factors, the possible solution concentration change.

test interval has been established.

Sodium pentaborate concentration is determined within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the addition of water or boron, or if the solution for temperature drops below specified limits. The 24-hour limit allows 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of mixing, subsequent testing, and notification of shift personnel.

on a Boron concentration, solution temperature, and volume are checked to assure a high reliability of operation of the system frequency should it ever be required. Isotopic tests of the sodium pentaborate are performed periodically to ensure that the proper boron-10 atom percentage is being used.

IOCFR50.62(c) (4) requires a Standby Liquid Control System with a of 13 minimum flow capacity and boron content equivalent to 86 gpm weight percent natural sodium pentaborate solution in the 251-inch reactor pressure vessel reference plant. Natural sodium pentaborate solution is 19.8 atom percent boron-10. The relationship expressed in Specification 3.4.C.3 also contains the ratio M251/M to account for the difference in water volume between the reference plant and Vermont Yankee. (This ratio of masses is 628,300 lbs./401,247 lbs.)

To comply with the ATWS rule, the combination of three Standby Liquid Control System parameters must be considered: boron concentration, Standby Liquid Control System pump flow rate, and boron-10 enrichment.

Fixing the pump flow rate in Specification 3.4.C.3 at the minimum flow rate of 35 gpm conservatively establishes a system parameter that can be used in satisfying the ATWS requirement, as well as the original system design basis. If the product of the expression in Specification 3.4.C.3 is equal to or greater than unity, the Standby Liquid Control System satisfies the requirements of 10CFR50.62(c) (4).

2-G4 98 Amendment No. 44-2-, 4-24, 445-,

VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION D. Safety and Relief Valves D. Safety and Relief Valves

1. During reactor power 1. Operability testing of operating conditions and Safety and Relief Valves whenever the reactor shall be in accordance coolant pressure is with Specification 4.6.E.

greater than 150 psig and The lift point of the temperature greater than safety and relief valves shall be set as specified I 350 0 F, all safety valves and at least three of the four relief valves shall in Specification 2.2.B.

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 350OF within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

E. Structural Integrity and E. Structural Integrity and Operability Testing Operability Testing

1. Inservice inspection of The structural integrity and safety-related components the operability of the shall be performed in safety-related systems and accordance with components shall be Section XI of the ASME maintained at the level Boiler and Pressure required by the original Vessel Code and acceptance standards applicable Addenda as throughout the life of the required by 10 CFR 50, plant. Section 50.55a(g), except where specific written relief has been granted by the NRC.

Inservice inspection of piping, identified in NRC Generic Letter 88-01, shall be performed in accordance with the staff positions on schedule, methods, and personnel and sample expansion included in the Generic Letter or in accordance with alternate measures approved by NRC Staff.

Amendment No. *-1, -14, 4-., 94, 4-2, 4-9,4-64, 4-7-2, 4-4, 49& 120

VYNPS 3.11 LIMITING CONDITIONS FOR 4.11 SURVEILLANCE REQUIREMENTS OPERATION 4.11 REACTOR FUEL ASSEMBLIES 3.11 REACTOR FUEL ASSEMBLIES Applicability:

Applicability:

The Surveillance Requirements The Limiting Conditions for apply to the parameters which Operation associated with the monitor the fuel rod operating fuel rods apply to these conditions.

parameters which monitor the fuel rod operating conditions.

Objective:

The Objective of the The Objective of the Limiting Surveillance Requirements is to Conditions for Operation is to specify the type and frequency assure the performance of the of surveillance to be applied fuel rods. to the fuel rods.

Specifications:

A. Average Planar Linear Heat A. Average Planar Linear Heat Generation Rate (APLHGR)

Generation Rate (APLHGR)

The APLHGR for each type of During operation at fuel as a function of

>25% Rated Thermal average planar exposure, Power, the APLHGR for each type of fuel as a function of average power, and flow shall be determined once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after >25% Rated I

planar exposure, power, I and flow shall not exceed the limiting values provided in the Thermal Power and daily during operation at >25%

Rated Thermal Power thereafter.

Core Operating Limits Report. For single recirculation loop operation, the limiting values shall be the values provided in the Core Operating Limits Report listed under the heading "Single Loop Operation." If at any time during operation at >25% Rated Thermal Power it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, APLHGR(s) shall be returned to within prescribed limits within two (2) hours; otherwise, the reactor shall be brought to <25% Rated Thermal Power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

44-, 14-6p, 4-81-8 224 Amendment No. 47, -,

4-7, *6-4, 44,

,94,

VYNPS 4.11 SURVEILLANCE REQUIREMENTS 3.11 LIMITING CONDITIONS FOR OPERATION C. Minimum Critical Power Ratio C. Minimum Critical Power Ratio (MCPR)

(MCPR)

MCPR shall be determined

1. During operation at >25% once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after Rated Thermal Power The MCPR operating value shall be equal to or

>25% Rated Thermal Power and daily during operation at 11

>25% Rated Thermal Power greater than the MCPR thereafter.

limits provided in the Core Operating Limits Report. For single recirculation loop operation, the MCPR Limits at rated flow are also provided in the Core Operating Limits I Report. If at any time during operation at >25%

Rated Thermal Power it is determined by normal surveillance that the limiting value for MCPR is being exceeded, MCPR(s) shall be returned to within the prescribed limits within two (2) hours; otherwise, the reactor power shall be brought to <25% Rated Thermal Power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

1-64 226 Amendment No. *4, 44-, 84, 94, 14,

VYNPS BASES:

3.11 FUEL RODS A. Average Planar Linear Heat Generation Rate (APLHGR)

Refer to the appropriate topical reports listed in Specification 6.6.C for analyses methods.

(Note: All exposure increments in this Technical Specification section are expressed in terms of megawatt-days per short ton.)

The MAPLHGR reduction factor for single recirculation loop operation is based on the assumption that the coastdown flow from the unbroken recirculation loop would not be available during a postulated large break in the active recirculation loop. See Core Operating Limits Report for the cycle-specific reduction factor.

Flow dependent MAPLHGR limits, MAPFAC(F), were designed to assure adherence to all fuel thermal-mechanical design bases. The same transient events used to support the MCPR(F) operating limits were analyzed, and the resulting overpowers were statistically evaluated as a function of the initial and maximum core flow. From the bounding overpowers, the MAPFAC(F) limits were derived such that the peak transient LHGR would not exceed fuel mechanical limits. The flow dependent MAPLHGR limits are cycle-independent and are specified in terms of multipliers, MAPFAC(F), to be applied to the rated MAPLHGR values.

Power-dependent MAPLHGR limits, expressed in terms of a MAPLHGR multiplier, MAPFAC(P), are substituted to assure adherence to the fuel thermal-mechanical design bases. Both incipient centerline melting of fuel and plastic strain of the cladding are considered in determining the power dependent MAPLHGR limit. Generally, the limiting criterion is incipient centerline melting. The power-dependent MAPFAC(P) multipliers were generated using the same database as used to determine the MCPR multiplier (Kp). Appropriate MAPFAC(P) multipliers are selected based on plant-specific transient analyses with suitable margin to assure applicability to future reloads. These limits are derived to assure that the peak transient MAPLHGR for any transient is not increased above the fuel design bases values.

B. Linear Heat Generation Rate (LHGR)

Refer to the appropriate topical reports listed in Specification 6.6.C for analyses methods.

Power and flow dependent LHGR limits are implemented using LHGRFAC multipliers on the standard LHGR limits. The LHGRFAC multipliers are identical to the MAPFAC multipliers.

C. Minimum Critical Power Ratio (MCPR)

Operating Limit MCPR

1. The MCPR operating limit is a cycle-dependent parameter which can be determined for a number of different combinations of operating modes, initial conditions, and cycle exposures in order to provide reasonable assurance against exceeding the Fuel Cladding Integrity Safety Limit (FCISL) for potential abnormal occurrences. The MCPR operating limits are justified by the analyses, the results of which are presented in the current cycle's Supplemental Reload Amendment No. 4--, 4-;L, 44O, *4, -94, 45-, 4-G-, - I4, BV*r 9-9 * , 171 227

VYNPS BASES:

3.11 FUEL RODS (Continued)

Licensing Report. Refer to the appropriate topical reports listed in Specification 6.6.C for analysis methods. The increase in MCPR operating limits for single loop operation accounts for increased core flow measurement and TIP reading uncertainties.

Flow-dependent MCPR limits, MCPR(F), are necessary to assure that the Safety Limit MCPR (SLMCPR) is not violated during recirculation flow increase events. The design basis flow increase event is a slow (maximum two pump runout rate of 1%/second)recirculation flow increase event which is not terminated by scram, but which stabilizes at a new core power corresponding to the maximum possible core flow. Flow runout events were analyzed along a constant xenon, constant feedwater temperature flow control line assuming a quasi steady-state plant heat balance. The ARTS-based MCPR(F) limit is specified as an absolute value and is cycle-independent. The operating limit is based on the maximum core flow limiter setting of 109.5% in the Recirculation Flow Control System.

Above the power at which the scram is bypassed (Pbypass), bounding power-dependent trend functions have been developed. This trend function, Kp, is used as multiplier to the rated MCPR operating limits to obtain the power-dependent MCPR limits, MCPR(P). Below the power at which the scram is automatically bypassed (Below Pbypass), the MCPR(P) limits are actual absolute Operating Limit MCPR (OLMCPR) values, rather than multipliers on the rated power OLMCPR.

Amendment No. 227a

VYNPS BASES:

4.11 FUEL RODS at A. The APLHGR, LHGR and MCPR shall be checked daily when operating if fuel burnup, or control rod

>25% Rated Thermal Power to determine has caused changes in power distribution. Since changes due movement rods are removed daily, a to burnup are slow, and only a few control is adequate. For a limiting value daily check of power distribution to occur below 25% of rated thermal power, an unreasonably large operating peaking factor would be required, which is not the case for control rod sequences. The 12 hour allowance after thermal power >25%

Thermal Power is achieved is acceptable given the large inherent Rated margin to operating limits at low power levels.

plant B. At certain times during plant startups and power changes the that surveillance of APLHGR, LHGR and/or technical staff may determine more frequently than daily. Because the necessity MCPR is necessary for such an augmented surveillance program is a function of a number of interrelated parameters, a reasonable program can only be The determined on a case-by-case basis by the plant technical staff.

be done using the plant check of APLHGR, LHGR and MCPR will normally process computer. In the event that the computer is unavailable, the of check will consist of either a manual calculation or a comparison existing core conditions to those existing at the time of a previous check to determine if a significant change has occurred.

If a reactor power distribution limit is exceeded, an assumption regarding an initial condition of the DBA analysis, transient analyses, or the fuel design analysis may not be met. Therefore, prompt action should be taken to restore the APLHGR, LHGR or MCPR to within the required limits such that the plant operates within The 2 analyzed conditions and within design limits of the fuel rods.

time is sufficient to restore the APLHGR, LHGR, or hour completion MCPR to within its limits and is acceptable based on the low the probability of a transient or DBA occurring simultaneously with APLHGR, LHGR, or MCPR out of specification.

C. Minimum Critical Power Ratio (MCPR) - Surveillance Requirement reactor At core thermal power levels less than or equal to 25%, the will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, core flow increase would only place operation in a any inadvertent more conservative mode relative to MCPR. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% thermal power level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement since for calculating MCPR above 25% rated thermal power is sufficient power distribution shifts are very slow during normal operation.

228 Amendment No. *44, 44&

VYNPS include a tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man-rem exposure according to work and job functions, 1/ e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.

The dose assignment to various duty functions may be estimates based on Self-Reading Dosimeter (SRD), TLD or film badge measurement. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work functions.

B. Monthly Operating Reports Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis no later than the fifteenth of each month following the calendar month covered by the report. These reports shall include a narrative summary of operating experience during the report period which describes the operation of the facility.

C. Core Operating Limits Report The core operating limits shall be established and documented in the Core Operating Limits Report (COLR) before each reload cycle or any remaining part of a reload cycle for the following:

1. The Average Planar Linear Heat Generation Rates (APLHGR) for Specifications 3.11.A and 3.6.G.la,

2. The Minimum Critical Power Ratio (MCPR) for Specifications 3.11.C and 3.6.G.la,

3. The Linear Heat Generation Rates (LHGR) for Specifications 2.l.A.la and 3.11.B, and

4. The Power/Flow Exclusion Region for Specifications 3.6.J.la and 3.6.J.lb.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

Report, E. E. Pilat, "Methods for the Analysis of Boiling Water Reactors Lattice Physics," YAEC-1232, December 1980 (Approved by NRC SER, dated September 15, 1982).

20.

1/ This tabulation supplements the requirements of 20.2206 of 10 CFR Part 11-6, 446, 1-5, a4, 2 259 Amendment No. 4-2-,

v t e Letter Sequence Request
TAC:MB8070, (Approved, Closed)
Initiation Request Acceptance... Administration Withholding Request Acceptance, Withholding Request Acceptance Meeting, Meeting, Meeting, Meeting Questions 21 May 2003 Answers Results Approval Other: BVY 05-002, Corrected Technical Specification Bases Page and Administrative Changes to Facility Operating License Page, ML040920352
MONTHYEAR2003-05-21 [Table View]

ML030850467

Text

Subject:

1.0 either

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