W3F1-2010-0051, Technical Specification Bases Update to the NRC for Period July 2, 2009 Through May 24, 2010

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Technical Specification Bases Update to the NRC for Period July 2, 2009 Through May 24, 2010
ML101480065
Person / Time
Site: Waterford Entergy icon.png
Issue date: 05/25/2010
From: Steelman W
Entergy Nuclear South, Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
W3F1-2010-0051
Download: ML101480065 (29)
v • d • e


Text

Entergy Nuclear South Entergy Operations, Inc.

17265 River Road Killona, LA 70057-3093

"=-=ý Entergy Tel 504 739 6685 Fax 504 739 6698 wsteelm@entergy.com William J. Steelman Acting - Licensing Manager Waterford 3 W3F1-2010-0051 May 25, 2010 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Technical Specification Bases Update to the NRC for the Period July 2, 2009 through May 24, 2010 Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

Dear Sir or Madam:

Pursuant to Waterford Steam Electric Station Unit 3 Technical Specification (TS) 6.16, Entergy Operations, Inc. (EOI) hereby submits an update of all changes made to Waterford 3 Technical Specification Bases since the last submittal per letter W3F1-2009-0031 (ADAMS Accession #ML091960049), dated July 9, 2009. This TS Bases update satisfies the requirement listed in 10CFR50.71 (e).

There are no commitments associated with this submittal. Should you have any questions or comments concerning this submittal, please contact William J. Steelman at (504) 739-6685.

Sincerely, WJS/RJ P/ssf Attachment(s): Waterford 3 Technical Specification Bases Revised Pages Aloor

W3F1-2010-0051 Page 2 cc: Mr. Elmo E. Collins, Jr.

Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125 NRC Senior Resident Inspector Waterford Steam Electric Station Unit 3 P.O. Box 822 Killona, LA 70066-0751 U. S. Nuclear Regulatory Commission Attn: Mr. N. Kalyanam Mail Stop O-07D1 Washington, DC 20555-0001 Wise, Carter, Child & Caraway ATTN: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn ATTN: N.S. Reynolds 1700 K Street, NW Washington, DC 20006-3817 Morgan, Lewis & Bockius LLP ATTN: T.C. Poindexter 1111 Pennsylvania Avenue, NW Washington, DC 20004

Attachment to W3FI-2010-0051 Waterford 3 Technical Specification Bases Revised Pages

Attachment to W3F1-2010-0051 Page 1 of 3 Waterford 3 Technical Specification (TS) Bases Revised Pages TS Implementation Affected TS Topic of Change Bases Date Bases Pages Change No.

61 7/30/2009 B 3/4 8-1a Change No. 61 was implemented by EC B 3/4 8-1 b 15945 to make editorial corrections to B 3/4 8-1c TS Bases section 3/4.8.1, 3/4.8.2, and B 3/4 8-1d 3/4.8.3, A.C. Sources, and Onsite Power Distribution Systems. The change addresses the approved changes documented in License Amendment 216, which revis'ed TS 3.8.1.3 and which had been previously incorporated into the bases.

Corrections were made to grammar, spelling, and minor errors made when incorporating the approved change.

The TS Bases changes under this EC remain consistent with License Amendment 217.

62 10/5/09 B 3/4 9-2 Change No. 62 to TS Bases section 3/4.9.6 Refueling Machine was implemented by EC 17724. The change addresses the approved changes documented in License Amendment 220. The Amendment updated TS 3.9.6 to explicitly state to place the refuel machine in a safe condition if it became inoperable with a fuel assembly or control element assembly attached to the mast. The TS 3.9.6 bases change only describes the TS 3.9.6 change.

j,

Attachment to W3F1-2010-0051 Page 2 of 3 TS Implementation Affected TS Topic of Change Bases Date Bases Pages Change No.

63 10/8/2009 B 3/4 3-1 Change No. 63 to TS Bases sections B 3/4 3-1a 3/4.3.1 and 3/4.3.2, Reactor Protective B 3/4 3-1b and Engineered Safety Features B 3/4 3-1c Actuation Systems Instrumentation was B 3/4 3-1d implemented by EC 17731. The change B 3/4 3-1e addresses the approved changes documented in License Amendment 222. This amendment resolved an issue noted from Cycle 15 startup, where a more restrictive calibration assumption for CPC delta T and neutron flux power was noted in Startup Test and Setpoints Transmittal (STST) for Cycle 15 than what was contained in TS Table 4.3-1 Note 2 requirement.

Amendment 222 revised Note 2 to align with STST. Additionally, this amendment clarified that:

(1) No adjustments are required below 15% RATED THERMAL POWER, (2) Certain adjustments should result in CPC power indications being as close as practical to calorimetric power, and (3) Ranges of acceptance criteria and adjustment limits are stated as percentages of RATED THERMAL POWER instead of percentages of current power.

4- 4 +

64 11/17/2009 B 2-1 Change No. 64 to TS Bases section B 2-2a 2.1.1, Reactor Core and 2.1.2, Reactor B 2-6 Trip Setpoints was implemented by EC 18510, and addresses the approved changes documented in License Amendment 224. The bases change reflects the TS 2.1.1.1 change in that the DNBR safety limit has been changed from 1.26 to 1.24 using the WSST-T/ABB-NV critical heat flux correlations. This bases change reiterates the TS 2.1.1.1 change that was reviewed and approved by the NRC.

Attachment to W3F1 -2010-0051 Page 3 of 3 TS Implementation Affected TS Topic of Change Bases Date Bases Pages Change No.

65 11/25/09 IX ( Change No. 65 to TS Bases section XIX 3/4.9.12 and 3/4.9.13, Spent Fuel Pool B 3/4 9-4 Boron Concentration and Spent Fuel B 3/4 9-5 Storage was implemented by EC 18742, and addresses the approved.

changes documented in License Amendment 223. The amendment revised the Waterford 3 TS 5.6 (FUEL STORAGE) to take credit for soluble boron in Region 1 (cask storage pit) and Region 2 (spent fuel pool and refueling canal) fuel storage racks for the storage of both Standard and Next Generation Fuel (NGF) assemblies. The

'amendment also adds new TS 3/4.9.12 (SPENT FUEL POOL (SFP) BORON-CONCENTRATION) which includes a surveillance that ensures the required boron concentration is maintained in the spent fuel storage racks and new TS 3/4.9.13 (SPENT FUEL STORAGE) to reflect the results of the new criticality analysis.

TS Bases sections affected by the TS change are TS Bases Table of Contents, List of Figures, Section 3.9.12 and 3.9.13. The TS Bases change only reiterates the information that was sent to the NRC for review and subsequently approved.

TECHNICAL SPECIFICATION BASES CHANGE NO. 61 REPLACEMENT PAGE(S)

(4 pages)

Replace the following pages of the Waterford 3 Technical Specification Bases with the attached pages. The revised pages are identified by Change Number 61 and contain the appropriate EC number and a vertical line indicating the areas of change.

Remove Insert B 3/4 8-1a B 3/4 8-1a B 3/4 8-1b B 3/4 8-1 b B 3/4 8-1c B 3/4 8-1c B 3/4 8-1d B 3/4 8-1d

ELECTRICAL POWER SYSTEMS BASES 3/4.8.1, 3/4.8.2. and 3/4.8.3 A.C. SOURCES, AND ONSITE POWER DISTRIBUTION SYSTEMS (Continued)

-#(DRN 03-375, Ch. 19)

The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ens ures that(1) the facility can be maintained in the shutdown or refueling condition for extended time periods and (2) sufficient instrumentation and control capability is available for monitoring and maintaining the unit status. With the minimum AC and DC power sources and assoc iated distribution systems inoperable the ACTION requires the immediate suspension of various activities including operations involving positive reactivity additions that could result in loss of required SHUTDOWN MARGIN (MODE 5) or boron concentration (MODE 6). Suspending positive reactivity additions that could result in failure to meet the minimum SHUTDOWN MARGIN or boron concentration Ii mit is required to assure continued safe operation. I ntroduction of coolant inventory must be from sources that have a boron concentration greater tha n that what would be required in the RCS for minimum SHUTDOWN MARGIN or refueling concentration. Thi s may result in an overall reduction i n boron concentration, but provides acceptable margin to maintaining subcritical operation. Introduction of tem perature changes, including increases when operating with a positive moderator temperature coefficient, must also be evaluated to ensure they do not result in a loss of required SHUTDOWN MARGIN. Suspension of these activities does not preclude completion of actions to establish a safe conservative condition.

.- (DRN 03-375, Ch. 19) 4(EC-10752, Ch. 56)

LCO 3.8.1.3 ACTION a

-+(EC-15945, Ch. 61)

This ACTION ensures that each diesel generator fuel oil storage tank (FOST) contains fuel oil of a sufficient volume to operate each diesel generator for a period of 7 days. An.

administrative limit of greater than 40,033 gallons as sures at least 39,300 usable gal Ions are stored in the tank accounting for volumetric shrink and instrumentation uncertainty. This useable volume is sufficient to operate the diesel, generator for 7 days based on the tim e-dependent loads of the diesel generator follow ing a loss of offsite power and a design bas es accident and includes the capacity to power the engineered safety features in conform ance with Regulatory Guide 1.137 October 1979. The minimum onsite stored fuel oil is sufficient to operate the diesel generator for a period longer than the tim e to replenish the onsite suppl y from the outside sources discussed in FSAR 9.5.4.2.

An additional provision is included in the ACTION which allows the diesel generators to remain operable when their 7 day fuel oil suppl y is not available provided that at least a 6 day supply of fuel oil is available. This provision is acceptable on the basis that replacement fuel oil is onsite within the first 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after falling below the 7 day supply. An administrative limit of greater than 37,696 gallons ass ures at least 37,000 usable gallons are stored in the tank, 4-(EC-10725, Ch. 56; EC-15945, Ch. 61)

AMENDMENT NO. 92,-66, WATERFORD - UNIT 3 B 3/4 8-1a CHANGE NO.. +9, 3-,56, 61

ELECTRICAL POWER SYSTEMS BASES 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, AND ONSITE POWER DISTRIBUTION SYSTEMS (Continued)

-4(EC-10752, Ch. 56)

LCO 3.8.1.3 (Continued)

ACTION a (Continued) accounting for volumetric shrink and instrumentation uncertainty. This useable volume is sufficient to operate the diesel generator for 5 days based on the full continuous load (4400kW) of the diesel generator and is sufficient to operable the diesel generator for greater than 6 days based on the time dependent loads of the diesel generator followi ng a loss of offs ite power and a design basis accident.

ACTION b

This ACTION is entered as a result of a failure to meet the acceptance criterion of particulate limits. Normally, trending of particulate levels allows sufficient time to correct high particulate levels prior to reaching the lim it of acceptability. Poor sample procedures (bottom sampling), contaminated sampling equipment, and errors in laboratory analysis can produce failures that do not follow a trend. Since the presence of particul ates does not mean failure of the fuel oil to burn properly in the diesel engine, and particulate concentration is unlikely to change significantly between surveillance frequency intervals, and proper engine per formance has been recently dem onstrated (within 31 days), it i s prudent to allow a brief period prior to declaring the associated DG inoperable. The 7-day Completion Time allows for further evaluation, re-sam piing, and re-analysis of the D G fuel oil.

4-(EC-15945, Ch. 61)

ACTION c With the new fuel oil properties defined in the Bases for SR 4.8.1.1.2.c not within the required limits, a period of 30 days is allowed for restoring the stored fuel oil properties. T his period provides sufficient time to test the stored fuel oil to determine that the new fuel oil, when mixed with previously stored fuel oil, remains acceptable, or restore the stored fuel oil properties. This restoration may involve feed and bleed procedures, filtering, or combinations of these procedures. Even if a diesel generator start and load was required during this ti me interval and the fuel oil properties were outside limits, there is a high likelihood that the diesel generator would still be capable of perform ing its intended function.

ACTION d

-4(EC-15945, Ch. 61)

This ACTION is entered as a result of the failure to m eet any of the other ACTIONS.

4-(EC-10725, Ch. 56; EC-15945, Ch. 61)

WATERFORD - UNIT 3 B 3/4 8-1 b CHANGE NO. 56,61

ELECTRICAL POWER SYSTEMS BASES 3/4.8.1. 3/4.8.2, and 3/4.8.3 A.C. SOURCES, AND ONSITE POWER DISTRIBUTION SYSTEMS (Continued) 4(EC-10752, Ch. 56)

SR 4.8.1.3.1 This SR provides verification that there is an adequate inventory of fuel oil in the storage tanks to support each E DG's operation for 7 days at full load. The 7 day period is sufficient time to place the unit in a safe shutdown condition and to bring in repleni shment fuel from an off site location. The 31 day Frequency is adequate to ensure that a suffi cient supply of fuel oil is available, since low level alarms are provided and unit operators would be aw are of any large uses of fuel oil during this period.

SR 4.8.1.3.2 SR 4.8.1.3.2 provides a means of determining whether new fuel oil i s of the appropriate grade and has not been contaminated with substances that would have an immediate, detrimental impact on diesel engine combustion. If results from the tests are within acceptable limits, the fuel oil may be added to the storage tanks w ithout concern for contaminating the entire volume of fuel oil in the storage tanks. The tests are to be conducted prior to adding the new fuel to the storage tanks, but in no case is the time between receipt of the new fuel and conducting the tests to exceed 31 days. The tests, limits and applicable ASTM Standards are as follows:

a. Sample the new fuel oil in accordance with ASTM D4057-06.

+-(EC-15945, Ch. 61)

b. Verify in accordance with the tests specified in ASTM D975-7b that the sam pie has a kinematic viscosity at 40 0 C of > 1.9 centistokes and < 4.1 centistokes, and a flash point

>1250 F,

c. Verify in accordance with A STM D1 298 or ASTM D4052 that the sam pie has an absolute specific gravity of 60/600 F of >0.85 and <0.885 or an API gravity at 60 0 F of

>28.40 and <350 and

d. Verify that the new fuel oil has a clear and bright appearance with proper color when tested in accordance with ASTM D4176-04 or water and sediment content within limits when tested in accordance with ASTM D2709-96.

-+(EC-15945, Ch. 61)

Failure to meet any of the above lim its is cause for rejecting the new fuel oil, but does not represent a failure to m eet the LCO since the fuel oil is not added to the storage tanks.

4-(EC-15945, Ch. 61)

Within 31 days following the initial new fuel oil sample, the fuel oil is analyzed to establish that the other properties specified in Table 1 of ASTM D975-7b are met for Grade 2-D 4-(EC-10725, Ch. 56)

WATERFORD - UNIT 3 B 3/4 8-1 c CHANGE NO. 56,61

ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, AND ONSITE POWER DISTRIBUTION SYSTEMS (Continued)

-#(EC-10752, Ch. 56)

SR 4.8.1.3.2 (Continued) new fuel oil when tested in accordance with ASTM D975-7b. The 31 day period is acceptable because the fuel oil properties of interest, even if they were not within stated limits, would not have an immediate effect on diesel generator operation. This Surveillance ensures the availability of high quality fuel oil for the diesel generators.

-4(EC-15945, Ch. 61)

Fuel oil degradation during I ong term storage shows up as an increase in particulate, due mostly to oxidation. The presence of particulate does not mean the fuel oil will not burn properly in a diesel engine. The particulate can cause fouling of filters and fuel oil injection equipment which can causeIengi ne failure.

Particulate concentrations will be determined in accordance with ASTM D6217-98. This method involves a gravimetric determination of total particulate concentration in the fuel oil and has a limit of <10 mg/I. It is acceptable to obtain a field sample for subsequent laboratory testing in lieu of field testing.

4-(EC-15945, Ch. 61)

The frequency of this test takes into consideration fuel oil degradation trends that indicate that particulate concentration in unlikely to change significantly between test intervals.

4-(EC-10725, Ch. 56)

WATERFORD - UNIT 3 B 3/4 8-1 d CHANGE NO. 56,61

TECHNICAL SPECIFICATION BASES CHANGE NO. 62 REPLACEMENT PAGE(S)

(1 page)

Replace the following page of the Waterford 3 Technical Specification Bases with the attached page. The revised page is identified by Change Number 62 and contain the appropriate EC number and a vertical line indicating the areas of change.

Remove Insert B 3/4 9-2 B 3/4 9-2

REFUELING OPERATIONS BASES 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS (Continued) 4(DRN 03-178, Ch. 21) closure" rather than "containment OPERABILITY." Containment closure means that all potential escape paths are closed or capable of being closed. Since there is no potenti al~for containment pressurization, the Appendix J leakage criteria and tests are not required.

During CORE ALTERATIONS or movement of irradiated fuel within the containment, the escape of radioactivity to the environment is minimized when the LCO requirements are met.

The equipment door, personnel airlock doors, or penetrations may be open during movement of irradiated fuel in the containm ent ahnd during CORE ALTERATIONS provided the equipment door, a minimum of one door in the airlock, and penetrations are capable of bei ng closed by an isolation valve, blind flange or manual valve, or capable of being cl osed on a containment purge isolation signal (CPIS) initiated by the required radiation m onitors in the event of a fuel handling accident. An OPERABLE containment purge isolation valve consists of a containment purge valve capable of isolating on manual initiation and on a containment purge isolation test signal from each of the required radiation monitoring instrumentation channels.

(Note that Technical Specifications 3/4.3.3, Radiation Monitoring, and 3/4.9.9, Containm ent Purge Isolation System, are also applicable.) Should a fuel handling accident occur inside containment, the equipment door, a minimum of one personnel airloc k door and the open penetrations will be closed. For closure, the equipment door will be held in place by a minimum of four symmetrically-placed bolts. The containment purge lines are automatically closed upon a CPIS if the fuel handling accident releases activity above prescribed levels. Closure of at least on of the containment purge isolation valves is sufficient to provide closure of the penetration. Containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be achieved by an OPERABLE automatic isolation valve, or by a manual isolation valve or blind flange.

4-(DRN 03-178, Ch. 21) 3/4.9.5 COMMUNICATIONS

  • The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during CORE ALTERATIONS.

3/4.9.6 REFUELING MACHINE 4 (EC-17724, Ch. 62)

The OPERABILITY requirements for'the refueling machine ensure that: (1) the refueling machine will be used for movement of CEAs and fuel assemblies, (2) each hoist has sufficient load capacity to lift a CEA or fuel assembly, and (3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged dur ing lifting operations. The Technical Specification Actions 'a.' and 'b.' statements allow the movement of a fuel assembly or CEA to safe condition using administrative controls in the event of a refueling machine failure.

4- (EC-17724, Ch. 62)

AMENDMENT NO. 144,-148, WATERFORD - UNIT 3 B 3/4 9-2 CHANGE NO. 19,21 62

TECHNICAL. SPECIFICATION BASES CHANGE NO. 63 REPLACEMENT PAGE(S)

(6 pages).

Replace the following pages of the Waterford 3 Technical Specification Bases with the attached pages. The revised pages are identified by Change Number 63 and contain the appropriate EC number and a vertical line indicating the areas of change.

Remove Insert B 3/4 3-1 B 3/4 3-1 B 3/4 3-1a B 3/4 3-1a B 3/4 3-1b B 3/4 3-1b B 3/4 3-1c B 3/4 3-1'c B 3/4 3-1d B 3/4 3-1d

-B 3/4 3-le

3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEMS INSTRUMENTATION The OPERABILITY of the Reactor Protective and Engineered Safety Features Actuation Systems instrumentation and bypasses ensures that (1) the associated Engineered Safety Features Actuation action and/or reactor trip will be initiated when the param eter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient sy stem functional capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity ass umed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses.

The redundancy design of the Control Element Assembly Calculators (CEAC) provides reactor protection in the event one or both CEACs become inoperable. If one CEAC is in test or inoperable, verification of C EA position is performed at least every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the second CEAC fails, the CPCs will use DNBR and LPD penalty factors to restrict reactor operation to some maximum fraction of RATED THERMAL POWER. If this maximum fraction is exceeded, a reactor trip will occur.

-*(EC-17731, Ch. 63)

Note 2 of Table 4.3-1 provides requirements for the periodic c alibration of CPC power indications using calorimetric power as the calibration standard.

No calibration of CPC power indications are required at less than 15% RATED THERMAL POWER since inherent conservatis ms in the CPC calculations at these power levels compensate for any potential decali bration. Significant differences between CP C power indications and calorimetric power observed during surveillances should always be investigated to'determ ine the cause of the deviation. T he most accurate calorimetric power indicati on available at the time of calibration should be used.

Between 15% and 80% power, if the daily surveillance finds that a CPC power indication is greater than the calorimetric power indication by more than 10% RTP, it should be adjusted to be within 8.0% and 10.0% RTP above the calorimetric. If the CPC power indications have been calibrated properly to the calorim etric power indication at high power (meani ng 80% or above),

then the most appropriate thing to do is not calibrate CPC powers below 80% power if they are conservative relative to calorimetric. In the extremely unlikely event that a CPC power indication is found to be more than 10.0% RTP higher than the calorimetric, it should be adjusted as little as possible to meet the requirem ents of the Technical Specifications. If this situation were to occur, it is likely that there is an anomaly in the calibration data or instrumentation. The safety and setpoint analysis does not explicitly address this situation because it is an unreasonable scenario without some other anomaly in the measurements, calibration or instrumentation. The 4-(EC-17731, Ch. 63)

AMENDMENT NO. 69, 113, 143, 154, WATERFORD - UNIT 3 B 3/4 3-1 CHANGE NO. 63

BASES (Cont'd) 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURE SAFETY ACTUATION SYSTEMS INSTRUMENTATION (Continued) 4(EC-17731, Ch. 63) probability of being greater than +10.0% from calibration following a power reduction from a calibrated condition, recalibrating to between +8.0% and +10.0% and then having a power increasing event which requires a CPC trip and having CPC be non-conservative at the point the trip is needed is too low to consider it as being within the CPC design basis.

At or above 80% RATED THERMAL POWER, the Note 2 phrase "as close as practical to calorimetric power" im plies that the as-left difference between the affected CP C power indication and calorimetric power should be as near to 0% RATED THERMAL POWER as possible. '  %

CPCs use the addressabl e constant PCALIB to determine power dependent biases for use in its calculations. Thus, when calibrations of CPC power indications are performed, it may be necessary to adjust the CPC constant PCALIB as described below:

While operating below 80% RATED THERMAL POWER, whenever the calibration of either CPC neutron flux power.or CPC AT power is adjusted, PCALIB must be set equal to the lower of the power level (in % RATED THERMAL POWER) of that adjustment and the power level (in % RATED THERMAL POWER) of the most recent calibration adjustment (or verification) of the other power indicat ion (the one not being calibrated).

PCALIB can be set to the current power level (in % RATED THERMAL POWER) whenever both CPC neutron flux power an d CPC LT power are adjusted or verified to be within the Technical Specification requirements at that power level.

PCALIB can be set to 100.0 whenever both CPC neutron flux power and CPC LT power have been adjusted or verified to be w.ithin the Technical Specification requirements at or above 80% RATED THERMAL POWER (plus uncertainty).

  • PCALIB must be set to 20.0 prior to initial power ascension following refueling.

4-(EC-17731, Ch. 63)

Table 3.3-3 ACTION 19 allows for continued operation in the applicable MODE(S) with one of the Refueling Water Storage Pool (RWSP) - Low or Steam Generator LP Emergency Feedwater Actuation Signal (EFAS) channels inoperable provided the c hannel is placed in the bypass or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. If an inoperable channel of the RWS P - Low or Steam Generator AP EFAS channel is required to be placed in the tripped condition within one hour, then within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> the channel must either be restored to OPERABLE status or be placed in the bypassed condition. The bypassed channel must be restored to OPERABLE status prior to entering the applicable MODE(S) following the next MODE 5 entry. With one of

'the RWSP - Low or Steam Generator LP (EFAS) channels inoperable and in bypass, and a failure occurs or repairs are necessary on one of the rem aining channels, ACTION 20 must be entered.

AMENDMENT NO. 6.,113,143153 WATERFORD - UNIT 3 B 3/4 3-1a +54-CHANGE NO. 63

3/4.3 INSTRUMENTATION BASES (Cont'd) 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURE SAFETY ACTUATION SYSTEMS INSTRUMENTATION (Continued)

ACTION 1 9a is annotated with a 3.0.4 exemption to allow the changing of MODES even though one channel is bypass ed. MODE changes between M ODES 1 and 4 with this configuration are allowed, to permit maintenance and testing on the inoperable channel. In this configuration, the protection system is in a two-out-of-th ree logic, and the probability of a random failure affecting two of the OPERABLE channels is remote. The tripped condition does not have this annotation as a single failure could cause the Emergency Core Cooling System and Containment Spray System suctions to be supplied from the Safety Injection System Sump prematurely and loss of the Low Pressure Safety Injection Systems with a premature Recirculation Actuation Signal (RAS) or with an inadvertent EFAS could cause the automatic isolation of a faulted steam generator from Emergency Feedwater (EFW) to not occur as assumed by the Waterford 3 safety analys is.

Table 3.3-3 ACTION 20 allows for continued operation in the applicable MODE(S) with two of the RWSP-Low or Steam Generator AP (EFAS) channels inoperable provided that one of the inoperable channels is bypassed and the other inoperable channel i s placed in the tripped condition within one hour.

One of the inoperable RWS P-Low channelsmust be restored to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to allow removal of the channel from the tripped condition. The allowed time is acceptable because operating experience has dem onstrated the low probability of the following sequence of events occurring: the need to place one RW SP-Low channel in the tripped condition while another RWSP-Low channel is in bypass, the receipt of a valid Safety Injection Actuation Signal Actuation, and a coincident failure of one of the two remaining OPERABLE RWSP-Low channels. These conditions could cause the Emergency Core Cooling System and Containment Spray System suctions to be supplied from the Safety Injection System Sump, prematurely due to containment pressure being higher than RWSP outlet pressure and loss of the Low Pressure Safety Injection Systems.

One of the inoperable Steam Generator AP (EFAS) channels must be restored to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to allow removal of the channel from the tripped condition.

The allowed time is acceptable because operating experience has demonstrated the low probability of the following sequence of events occurring: the need to place one S team Generator AP (EFAS) channel in the tripped condition while another Steam Generator AP (EFAS) is in bypass, coincident with a failure of one of the two remaining OPERABLE Steam Generator AP (EFAS) channels, and the occurrence of a MSLB or FWLB. These conditions could cause the automatic isolation of a faulted steam generator from Emergency Feedwater (EFW) to not occur as assumed by the Waterford safety analysis.

AMENDMENT NO. 113, 143, 154,-

WATERFORD - UNIT 3 B 3/4 3-1b CHANGE NO. 63

3/4 INSTRUMENTATION BASES (Cont'd) 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION (Continued)

When one of the inoperable channels i s restored to OPERABLE status, subsequent operation in the applic able MODE(S) may continue in accordance with the provisions of ACTION 19.

Because of the interaction between process measurement circuits and associated functional units as listed in the ACTIONS 19 and 20, placement of an inoperable channel of Steam Generator Level in the bypass or trip condition results in corresponding placements of Steam Generator LP (EFAS) instrumentation. Depending on the num ber of applicable inoperable channels, the provisions of ACTIONS 19 and 20 and the aforesaid scenarios f or Steam Generator A*P (EFAS) would govern.

The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The

.periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. The quarterly frequency for the channel functional tests for these systems comes from the analyses presented in topical report CE N-327: RPS/ESFAS Extended Test Interval Evaluation, as supplemented.

Testing frequency for the Reactor T rip Breakers (RT Bs) is described and analyzed in CEN NPSD-951. The quarterly RTB channel functional test and RPS logic channel functional test are scheduled and performed such that RTBs are verified OPERABLE at least every 6 weeks to accommodate the appropriate vendor recommended interval for cycling of each RTB.

RPS\ESFAS Trip Setpoints values are determined by means of an explicit setpoint calculation analysis. A Total Loop Uncertainty (T LU) is calculated for each RPS/ESFAS instrument channel. The Trip Setpoint is then determined by adding or subtracting the T LU from the Analytical Limit (add TLU for decreasing process value; subtract TLU for increasing process value). The Allowable Value is determined by adding an allowance between the Trip Setpoint and the Analytical Limit to account for RPS/ESFAS cabinet Periodic Test Errors (PTE) which are present du ring a CHANNEL FUNCTIONAL TEST. PTE combines the RPS/ESFAS cabinet reference accuracy, calibration equipm ent errors (M&TE), and RPS/ESFAS cabinet bistable Drift. Periodic testing assures that actual setpoints are within their Allowable Values. A channel is inoperable if its actual setpoint is not within its Allowable Value and corrective action must be taken. Operation with a trip set less conservative than its setpoint, but within its specified ALLOWABLE VALUE is acceptable on the basis that the difference between each trip Setpoint and the ALLOWABLE VALUE is equal to or less than the Periodic Test Error allowance assumed for each trip in the safety analy ses.

The measurement of response ti me at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the safety analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable.

AMENDMENT NO. +54 T-SeR .99-4 WATERFORD - UNIT 3 B 3/4 3-1c CHANGE NO. 4,- 9, 27, 57, 63

3/4 INSTRUMENTATION BASES (Cont'd) 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION (Continued)

Response time may be verified by any series of sequential, overlapping, or total channel measurements, including allocated sensor response ti me, such that the response time is verified. Allocations for sensor response times may be obtained from records of test results, vendor test data, or vendor engineering specific ations. Topical Report CE NPSD-1 167-A, "Elimination of Pressure Sensor Response Time Testing Requirements," provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the topical report. Respose time verification for other sensor types must be demonstrated by test. The allocation of sensor response times must be verified prior to placi ng a new component in operation and reverified after m aintenance that may adversely affect the sensor response tim e.

TABLE 3.3-1, Functional Unit 13, Reactor Trig Breakers The Reactor Trip Breakers Functional Unit in Table 3.3-1 refers to the reactor trip breaker channels. There are four reactor trip breaker channels. Two reactor trip breaker channels with a coincident trip logic of one-out-of-two taken twic e (reactor trip breaker channels A or B, and C or D) are required to produce a trip. Each reactor trip breaker channel consists of two reactor trip breakers. For a reactor trip breaker channel to be considered OP-ERABLE, both of the reactor trip breakers of that reactor trip breaker channel must be capable of performing their safety function (disrupting the flow of power in its respective trip leg). The safety function is satisfied when the reactor trip breaker is capable of automatically opening, or otherwise opened or racked-out.

If a racked-in reactor trip breaker is not capable of automatically opening, the ACTION for an inoperable reactor trip breaker channel shall be entered. The ACTION shall not be exited unless the reactor trip breaker capability to automatically open is restored, or the reactor trip breaker is opened or racked-out.

4(EC-12084, Ch. 57)

TABLES 3.3-3 and 4.3-2, Functional Unit 6, Loss of Power (LOV)

The Loss of Power Functional Unit 6 in Tables 3.3-3 and 4.3-2 refers to the undervoltage relay channels that detect a loss of bus voltage on the 4kV (A3 & B3) and 480V (A31 & B31) safety buses and a sustained degraded voltage condition on 4k V (A3 & B3) safety buses. The intent of these relays is to ensure that the Emergency Diesel Generator starts on a loss of voltage or a sustained degraded voltage conditi on. The response ti me SR in TS 3.3.2 ensures that Bus A3 and B3 undervoltage relays trip and generate a Loss of V oltage (LOV) signal in 2 seconds for.

initiation of the EDG start. The response time for Bus AB3 and AB31 relays is not as critical as the Bus A3 and B3 undervoltage relays. B us AB3 and AB31 undervoltage relays [4KVEREL3AB-1A(1B)(1C) and SSDEREL31AB-1A(1B)(1C)] strip bus loads upon an undervoltage condition to preclude any perturbations whic h might affect the A and B buses and prepare the bus to be energized by an E DG with subsequent loading by the sequenc er. Bus AB3 and AB31 4-(EC-12084, Ch. 57)

WATERFORD - UNIT 3 B 3/4 3-1 d CHANGE NO. 5iL,- 63

3/4 INSTRUMENTATION BASES (Cont'd)

TABLES 3.3-3 and 4.3-2. F unctional Unit 6, Loss of Power (LOV (Conti nued)

-#(EC-12084, Ch. 57) undervoltage relays do not provide an EDG start signal. Therefore, TS 3/4.3.2, Tables 3.3-3 and 4.3-2 functional unit 6 requirem ents, are not applicable to AB3 Bus and AB31 Bus undervoltage relays.

If an AB Bus undervoltage relay becomes inoperable, initiate a condition report and consider operability of the associated E DG based on the AB Bus loads when evaluati ng the failure.

4-(EC-12084, Ch. 57) 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that: (1) the radiation levels are continually measured in the areas served by the indivi dual channels; (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recom mendations of Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," December 1980 and NURE G-0737,.

"Clarification of TMI Action Plan Requirements," November 1980.

WATERFORD - UNIT 3 B 3/4 3-1e CHANGE NO. 63

TECHNICAL SPECIFICATION BASES CHANGE NO. 64 REPLACEMENT PAGE(S)

(3 pages)

Replace the following pages of the Waterford 3 Technical Specification Bases with the attached pages. The revised pages are identified by Change Number 64 and contain the appropriate EC number and a vertical line indicating the areas of change.

Remove Insert B2-1 B2-1 B 2-2a B 2-2a B 2-6 .B 2-6

'SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 2.1.1 REACTOR CORE

-+(DRN 02-458)

The restrictions of these safety limits prevent overheating of the fuel c ladding and possible cladding perforation w hich would-,result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by (1) restricting fuel operation to within the nucleate boi ling regime where the heat transfer coeffici ent is large and the cladding s urface temperature is slightly above the coolant saturation temperature, and (2) maintaining the peak fuel centerline temperature below the melting point.

First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the m aximum clad surface temperature is only slightly greater than the coolant saturation temperature. The upper boundary of the nucleate boi ling regime is termed "departure from nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat transfer coefficient, which would result in higher cladding temperatures and the possibility of cladding failure.

-4(EC-18510, Ch. 64)

Correlations predict DN B and the location of D NB for axially uniform and non-uniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the predicted DN B heat flux at a particular core location to the actual heat flux at that location, is indicative of the margin to DNB. The minimum value of DNBR during normal operational occurren ces is limited to 1.24 for the WSSV-T and ABB-NV correlations and is established as a Safety Limit. This value is based on a statistical combination of uncertainties. It includes uncertainties in the Critical Heat Flux (CHF) correlation, allowances for rod bow and hot channel factors (related to fuel manufacturing variations) and allowances for other hot channel calcul ative uncertainties (C EN-356(V)-P-A, "Modified Statistical Combination of Uncertainties," Revision 01-P-A, May 1988).

-(EC-18510, Ch. 64)

-+(DRN 02-458)

Second, operation with a peak linear heat rate below that which would cause fuel centerline melting maintains fuel rod and cladding integrity. Above this peak linear heat rate level (i.e., with some melting in the center), fuel rod integrity would be maintained only if the design and operating conditions are appropriatethroughout the life of the fuel rods. Volume changes which accompany the solid to liquid phase change are significant and require accom modation.

Another consideration involves the redistribution of the fuel which depends on the extent of the melting and thephysical state of the fuel rod at the time of melting. Because of the above factors, fuel centerline melting is established as a Safety Limit. The design melting point of new fuel with no burnable poison is 5080'F. The melting point is adjusted downward from this temperature depending on the am ount of burnup and am ount and type of burn able poison in the

-(DRN 02-458)

CHANGE NO.+--, 64 WATERFORD - UNIT 3 B 2-1 AMENDMENT NO. +-1,

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SETPOINTS (Continued)

A Total Loop Uncertainty (TLU) is calculated for each RPS instrument channel. The Trip setpoint is determined by adding or subtracting the T LU from the. Analytical Limit (add TLU for decreasing process value; subtract TLU for increasing process value). The Allowable Value is determined by adding an allowance betw een the Trip Setpoint and the Analytical Limit to account for RPS cabinet Periodic Test Errors (PTE) which are present during a CHA NNEL FUNCTIONAL TEST. PTE combines RPS cabinet reference accuracy, calibration equipment errors (M&TE), and RPS cabinet bistable drift. Periodic testing assures that actual set points are within their Allowable Values. A channel is inoperable if its actual setpoint is not within its Allowable Value and corrective action must be taken. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the PTE allowance assumed for each trip in the safety analyses.

-+(EC-18510, Ch. 64)

The DNBR - Low and Local Power Density - High are digitally generated trip setpoints based on Limiting Safety System Settings of 1.26 and 21.0 kW/ft, respectively. Since these trips are digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to trips generated by analog type equi pment. The Allowable Values for these trips are therefore the same as the Trip Setpoints. The CPC power adjustment addressable constant BERR1 is used such that the CP C DNBR trip setpoint of 1.26 using the CE -1 critical heat flux correlation assures that the bounding safety lim it DNBR of 1.24 for the WSSV-T and ABB-NV correlations wi ll not be exceeded during norm al operations and AO Os.

.- (EC-18510, Ch. 64)

To maintain the margins of safety assumed in the safety analyses, the calculations of the

  • trip variables for the DNB R - Low and Local P ower Density -High trips incl ude the measurement, calculational and processor uncertainties and dynamic allowances as defined in the latest applicable revision of CEN-305-P, "Functional Design Requirements for a Core Protection Calculator" and; CEN-304-P, "Functional Design Requirements for a Control Element Assembly Calculator."

WATERFORD - UNIT 3 B 2-2a CHANGE NO. +2-,64

BASES DNBR - Low (Continued)

-#(EC-18510, Ch. 64) in actual core DNBR after the trip will not result in a violation of the DNB R Safety Limit of 1.24.

CPC uncertainties related to DNBR cover CPC input measurement uncertainties, al gorithm modelling uncertainties, and computer equipment processing uncertainties. Dynamic compensation is provided in the CPC calculations for the effects of cool ant transport delays, core heat flux delays (relative to changes in core power), sensor time delays, and protection system equipment time delays.

4-(EC-18510, Ch. 64)

The DNBR algorithm used in the CPC is valid only within the limits indicated below and operation outside of these limits will result in a CPC initiated trip.

a. RCS Cold Leg Temperature-Low > 495°F
b. RCS Cold Leg Temperature-High < 580°F
c. Axial Shape Index-Positive Not more positive than +0.5
d. Axial Shape Index-Negative Not more negative than -0.5
e. Pressurizer Pressure-Low > 1860 psia
f. , Pressurizer Pressure-High < 2375 psia
g. Integrated Radial Peaking Factor-Low > 1.28
h. Integrated Radial Peaking Factor-High < 7.00 Quality Margin-Low > 0

-4(DRN 04-1243, Ch. 38)

The CPCs contain several auxiliary trip functions which are credited in the, safety analysis.

These trips manifest themselves as DNBR trips however they are making the trip determination on parameters other than DNBR.

The CPC Variable Overpower T rip (VOPT) is provided to include a trip on power which is compensated for the decali brating effects of changes in coolant temperature in the reactor vessel downcormer. Additionally, the trip setpoint is allowed to change with slow changes in plant power. Thus at intermediate steady state powers, the plant is protected by a power trip which is a small distance above steady state power levels. The rate at which the'autom atic increases and decreases in the setpoi nt may change are limited and accounted for in the safety anal ysis.

TheOCPCs contain a trip which detects asymmetries in cold leg loop temperatures resulting from as asymmetric steam generator transient. The trip occurs if the cold leg asymmetry exceeds 11 OF.

The CPCs contain a trip monitoring margin to saturation conditions in the hot legs. A trip will be generated if margin to saturation is Iess than 13 °F.

The CPCs contain a direct trip on low RCP speed. The trip will occur if the RCP speed drops below 0.965.

4-(DRN 04-1243, Ch. 38)

AMENDMENT NO. 1-2, WATERFORD - UNIT 3 B 2-6 CHANGE NO. 4--3, 64

TECHNICAL SPECIFICATION BASES CHANGE NO. 65 REPLACEMENT PAGE(S)

(4 pages)

Replace the following pages of the Waterford 3 Technical Specification Bases with the attached pages. The revised pages are identified by Change Number 65 and contain the appropriate EC number and a vertical line indicating the areas of change.

Remove Insert IX IX XIX XIX B 3/4 9-4 B 3/4 9-4 B 3/4 9-5

-*(DRN 05-747, Ch. 40)

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.8 ELECTRICAL POWER SYSTEMS (Continued) 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS O P E RA T ING ........................................................................ 3/4 8-13 SHUTDOW N ................................................................. 3/4 8-15 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES .................... 3/4 8-16 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION AND BYPASS DEVICES .................................. 3/4 8-52 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRA TION .................................................... 3/4 9-1 3/4.9.2 INSTRUM ENTATIO N .............................................................. 3/4 9-2 3/4.9.3 D ECA Y T IME .......................................................................... 3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS ........................ 3/4 9-4 3/4.9.5 CO MMUNICATIO NS ................................................................ 3/4 9-5 3/4.9.6 REFUELING MACHINE ........................................................... 3/4 9-6 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING .................... 3/4 9-7 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION HIGH W ATER LEVEL .......................................................... 3/4 9-8 LOW W ATER LEVEL........................................................... 3/4 9-9 3/4.9.9 CONTAINMENT PURGE VALVE ISOLATION SYSTEM ......... 3/4 9-10 3/4.9.10 WATER LEVEL - REACTOR VESSEL FUEL ASSEM BLIES ............................................................ 3/4 9-11 C E A s ..................................................................................... 3/4 9-12 3/4.9.11 WATER LEVEL - SPENT FUEL POOL ...................................... 3/4 9-13

-+(EC-18742, Ch. 65) 3/4.9.12 SPENT FUEL POOL (SFP) BORON CONCENTRATION ............ 3/4 9-13a 3/4.9.13 SPENT FUEL STORAGE ............................................................ 3/4 9-13b 4-(EC-18742, Ch. 65) 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOW N MARG IN................................................................. 3/4 10-1 4-(DRN 05-747, Ch. 40)

-#(DRN 05-747, Ch. 40)

WATERFORD - UNIT 3 IX AM[NDM,[NT NO. 176 4-(DRN 05-747, Ch. 40) CHANGE NO. 4-, 65

4(DRN 05-747, Ch. 40)

INDEX LIST OF FIGURES FIGURE PAGE 3.1-1 REQUIRED STORED-BORIC ACID VOLUME AS A FUNCTION OF CONCENTRATION (VOLUME OF ONE BAMT) ....................... 3/4 1-14 3.1-2 REQUIRED STORED BORIC ACID VOLUME AS A FUNCTION OF CONCENTRATION (COMBINED VOLUME OF TWO BAMT)... 3/4 1-14a 3.4-1 D E LET E D ..................................................................................... 3/4 4-27 3.4-2 WATERFORD UNIT 3 HEATUP CURVE - 32 EFPY REACTOR COOLANT SYSTEM PRESSURE - TEMPERATURE LIMIT S ....................................................................................... 3/4 4 -30 3.4-3 WATERFORD UNIT 3 COOLDOWN CURVE - 32 EFPY REACTOR COOLANT SYSTEM PRESSURE -TEMPERATURE LIMIT S ....................................................................................... 3/4 4 -3 1 3.6-1 D E LET E D .................................................................................... 3/4 6-12

-#(EC-15515, Ch. 60) 4.7-1 DELETED 4-(EC-15515, Ch. 60) 5.1-1 EX C LUS IO N A R EA ...................................................................... 5-2 5.1-2 LOW POPULATION ZONE .......................................................... 5-3 5.1-3 SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND I LIQ U ID EFFLU EN TS .................................................................

5-4

-4(EC-18742, Ch. 65) 5.6-1 ALTERNATIVE CHECKERBOARD STORAGE ARRANGEMENTS 5-6a 5.6-2 ACCEPTABLE BURNUP DOMAIN FOR UNRESTRICTED STORAGE OF IRRADIATED FUEL IN REGION 2 OF THE S PENT FUEL PO O L.................................................................. 5-6b 5.6-3 ACCEPTABLE BURNUP DOMAIN FOR IRRADIATED FUEL IN A CHECKERBOARD ARRANGEMENT WITH FUEL OF 5 WT% ENRICHMENT, OR LESS, AT 27 GWD/MTU BURNUP, OR HIGHER, IN REGION 2 OF THE SPENT FUEL P OO L............................................................................................ 5 -6 c 5.6-4 EXAMPLES OF CONTIGUOUS CHECKERBOARD CONFIGURATIONS WHICH MEET INTERFACE REQUIREMENTS ........................ 5-6d i-(EC-18742, Ch. 65) 6.2-1 D E LE T E D ..................................................................................... 6-3 6 .2-2 D E LE T E D .................................................................................... . 6-4

+-(DRN 05-747, Ch. 40)

-4(DRN 05-747, Ch. 40)

WATERFORD - UNIT 3 XIX AM[NDM[NT N). 13, 27, 102, 1,4, 188, 4-(DRN 05-747, Ch. 40) "96,*99 CHANGE NO. 40-,66, 65

REFUELING OPERATIONS BASES 3/4.9.9 CONTAINMENT PURGE VALVE ISOLATION SYSTEM (Continued) 4(DRN 03-233, Ch. 22)

The containment purge valve isolation system consists of the containment purge isolation valves (CAP-103, CAP-104, CAP-203 and CAP-204), the containment purge and exhaust isolation radiation monitors (one required per train as specified in TS 3/4.3.3), the containment purge isolation signal logic and manual isolation logic.

The ACTION statement to close each of the containment purge penetrations may be met by closing at least one valve per penetration (reference Technical Specification 3/4.9.4 and its Basis).

4-(DRN 03-233, Ch. 22) 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and SPENT FUEL POOL

-,(DRN 05-131 Ch. 39)

The restrictions on minimum water level ensure that sufficient water depth is available such that the iodine released as a resultof a rupture of an irradiated fuel ass embly is reduced by a factor of at least 200. Gap fractions are assumed in accordance with Regulatory Guide 1.183 guidance.

The minimum water depth is consistent with assumptions of the safety analys is.

4-(DRN 05-131, Ch. 39)

-#(EC-18742, Ch. 65) 3/4.9.12 and 3/4.9.13 SPENT FUEL POOL BORON CONCENTRATION and SPENT FUEL STORAGE TS 5.6, "FUEL STORAGE," reflects the results of the criticality analysis, crediting soluble boron and allowing more flexibility in storing the more reactive Next Generation Fuel (NGF) assemblies in the spent fuel storage racks. The Waterford 3 SFP criticality analysis used a design acceptance criteria of effective (neutron) m ultiplication factor (keff) no greater than 0.995, if flooded with unborated water, and k eff no greater than 0.945, if flooded with borated water. T his provides an additional 0.005 A keff analytical margin to the regulatory requirement. T his approach provides sufficient margin to offset minor non-conservatisms to provide reasonable assurance that the regulatory requirements are met. Each storage configuration has a geom etric arrangement which must be maintained so that the S FP criticality analysis remains valid..

The spent fuel pool (S FP) criticality analysis credits 524 parts per million (ppm) of soluble boron to maintain keff less than 0.95 in the SFP during normal conditions, and 870 ppm under the worst-case accident conditions. The analysis determined that a misloading event in the spent fuel checkerboard loading pattern would have t he largest reactivity increase, requiring 870 ppm of soluble boron to meet the regulation. The boron dilution analysis identified a number of assorted sources for slow addition of unborated water to the SF P that could possibly continue undetected for an extended period of time. The maximum flow from any of these sources was determined to be 2 gpm, and dilution of the S FP from 1900 ppm to 870 ppm soluble boron would take approxim ately 72 days. Slow dilution by undetected sources is adequately addressed by sampling the SFP on the 7-day frequency of SR 4.9.12. Higher flow-rate diluti on scenarios would be identified through various alarms and building walkdowns, and could be addressed within a sufficient time to preclude 4-(EC-18742, Ch. 65)

WATERFORD - UNIT 3 B 3/4 9-4 CHANGE NO. 19, 21, 22,39',-65

-,(EC-18742, Ch. 65)

REFUELING OPERATIONS BASES 3/4.9.12 and 3/4.9.13 SPENT FUEL POOL BORON CONCENTRATION and SPENT FUEL STORAGE (Continued)'

dilution of the SFP to 870 ppm soluble room. Adequate safety is maintained in the case of a high flow-rate dilution of the SFP in accordance with 10 CFR 50.68(b)(4) because keff must remain below 1.0 (subcritical), even if the S FP were flooded with unborated water:

Three qualified storage configurations are allowed for Region 2 F uel Storage locations, based on burnup versus enrichm ent restrictions: 1) uniform loading of assemblies, 2) checkerboard loading of hi gh and low reactivity assemblies, and 3) checkerboard loading of fresh assemblies and empty cells. The storage configurations may be interspersed with each other throughout the SFP, provided that the geom etric interface requirements are met.

Checkerboard loading is not required for Region I Fuel Storage locations.

"-(EC-18742, Ch. 65)

WATERFORD - UNIT 3 B 3/4 9-5 CHANGE NO.'65

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