ML20210U248
| ML20210U248 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 10/01/1986 |
| From: | LOUISIANA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20210U193 | List: |
| References | |
| NUDOCS 8610090463 | |
| Download: ML20210U248 (132) | |
Text
CURRENT PLAN POTENTIAL TECHNICAL SPECIFICATICS CHANCES Revisio2 6, 10/1/86
( /01/86)
( 8/21/86)
I 9I # )
(10/01/80 (3)
CROUP A (2)
GROUP B (7) CROUP C (5) CROUP D (10/01/86)
(11/21/86)
(12/25/86)
(01/01/87)
TECH SPEC NATURE OF TECH SPEC NATURE OF TECH SPEC NATURE OF TECH SPEC NATURE OF SECTION CHANGE SECTION CHANGE SECTION CHANGE SECTION CHANGE 3.1.3.4 CEA Drop Time 3.1.2.1 Change in BAM Tank 3.1.1.1 Shutdown Margin for 2.1.1.1 Redefine the Safety 18 month Surveillance Concentration MODES 1-4 may change Limit on DNBR (Submitted to NRC (Submitted to NRC (Submitted to NRC (Submitted to NRC on 6/24/86) on 8/20/86) on 9/25/86) on 10/1/86) 5.3.1 Fuel enrichment limit 3.1.2.2 Change in BAN Tank 3.1.1.2 Shutdown Margin for 2.2.1 DNBR trip limit increases from 3.7 Concentration MODE 5 may change (Submitted to NRC w/o to = 14.10 w/o (Submitted to NRC (Submitted to NRC on 10/1/86)
(Submitted to NRC on 8/20/86) on 9/25/86) on 6/24/86) 5.6.1 Update to reflect new 3.1.2.7 Change in BAM Tank 3.1.2.4 See MODE 5 Shutdown 3.2.4 Revise Figure and Change fuel storage criti-Concentration Margin Change Format. DNBR Margin; cality analyses for (Submitted to NRC (Submitted to NRC Delete Surveillance Cycle 2 on 8/20/86) on 9/25/86)
Requirement 4.2.4.4 (Submitted to NRC (Submitted to NRC on 6/24/86) on 10/1/86) 3.1.2.8 Change in BAM Tank 3.1.2.6 See MODE 5 Shutdown 3.3.1 Revise Table 3.3-2 Concentration Margin Change response times (Submitted to NRC (Subettted to NRC (Submitted to NRC on 8/20/86) on 9/25/86) on 10/1/86) 3.10.1 Surveillance Require-3.1.3.6 Revise Insertion ment 4.10.1.2 to be Limit Figure relaxed (Submitted to NRC (Submitted to NRC on 10/1/86) on 9/25/86) 3.10.3 Special Test Exception 3.1.2.9 Revise Table 3.1-1 for to allow low power boron dilution physics test detection (Submitted to NRC (Submitted to NRC on 9/25/86) 10/1/86) 3.2.1 Linear Heat Rate 8610090 (Submitted to NRC PDR AD 3 861001 an 10/1/86)
K 050093g2 P
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NS20546
m CURRENT Pl.AN POTENTIAL TECHNICAL SPECIFICATICU CHANCES Ravision 6, 10/1/86 f3 1/86 I
II (1)
GROUP G (8)
GROUP H (7)
CROUP E (7)
GROUP F f
. g j
3 TECH SPEC NATURE OF TECH SPEC NATURE OF TECH SPEC NATURE OF TECH SPEC NATURE OF SECTION CHANGE SECTION CHANGE SECTION CHANGE SECTION CHANGE 3.1.1.3 NTC will become more 3.1.3.1 CEA Misalignment 2.2.2 Removal, requested, 6.2.2 Staffing negative at EOC and ACTION statement 51FR 10465, 3/26/86 (Submitted to NRC on more positive at BOC need to be modified CPC Addressable 6/24/86)
(Submitted to NRC on (Submitted to NRC constant 7/15/86) 8/29/86)
(Tech Spec change 3.3.1 RPS, Allow Bypass received)
SG Level High Trip (Submitted to NRC on i
3.1.3.7 Add curve and change 3.3.3.8 Chan8e Table 3.3.3.8 6/24/86) short-term and tran-to list smoke detector sient insertion limits in Control Room 5.3.1 Correct mistake re-(Submitted to NRC on (Submitted to NRC characterizing 1807 gr 7/15/86) 8/29/86) uranium as maximum fuel i
rod loading 3.2.7 ASI ranges will (Submitted to NRC on change 6/24/86) i (Submitted to NRC on 7/15/86) 6.2.2.d Clarify SRO Staffing requirements during 3.3.3.6 Add RVLMS per License refueling operations Conditions and CEOG (Submitted to NRC (Submitted to NRC on on 9/10/86) 7/15/86) 3.6.1.2 Add spare penetration to leak rate testing list i
(Submitted to NRC on
+
l 3.10.2 Add 3.1.3.7 for 8/29/86)
]
part-length CEAs y
(Submitted to NRC on 3.3.3.3 Change location of
.I 7/15/86) seismic monitors (Submitted to NRC on 3
8/29/86) 3.3.3.10 Clarify liquid effluent monitoring instriasenta-tion action statements (Submitted to NRC on 10/1/86) 3.3.3.11 Clarify gaseous effluent monitoring instrumentation action statements (Submitted to NRC on 10/1/86)
w NPF-38-43 m
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DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-43 This is a request to revise Technical Specifications 2.1.1.1, " Safety Limits, Reactor Core - DNBR"; 2.2.1, " Safety Limits, Reactor Trip Setpoints; and the associated Bases of these specifications.
Existing Specification See Attachment A.
Proposed Specification See Attachment B.
Description The proposed change would revise Technical Specification 2.1.1.1, " Safety Limits, Reactor Core - DNBR"; 2.2.1, " Safety Limits, Reactor Trip Setpoints"; and the associated Bases to these Specifications. The reason for these changes is to account for a different method of treating uncertainties in the CE-1 heat flux correlation and DNBR calculation as applied to Waterford 3's 16 x 16 fuel assem-blies. Specifically, the uncertainties associated with fuel manufacturing varia-tions, as well as certain thermal-hydraulic uncertainties, will be combined using the methodology associated with the Statistical Combination of Uncertainties (SCU). These methods have been previously applied on other CE plants and have been approved by the NRC.
Use of the SCU methodology results in an increase in the minimum allowable value of the DNBR safety limit and the reactor trip setpoint for DNBR from a value of 1.205 to a Cycle 2 value of 1.260. This new value ensures, with a 95% confidence level, that if the hot channel in the core reaches the DNBR safety limit, there is still a 95% probability that DNBR has not occurred. This is the same probability / confidence level that applied to the 1.205 DNBR value that was used during Cycle 1.
Thus, although some uncertainties have been removed from the actual CPC calculation of core heat flux (and corresponding DNBR), they have been factored into the determination of the reactor trip setpoint for low DNBR as well as the associated DNBR Safety Limit.
The proposed change would also modify the pressurizer pressure range over which the DNBR algorithm used by the CPCs is valid. The new range is from 1860 to 2375 psia and is only slightly different from the range used during Cycle 1 (1845 to 2355 psia). This change is being submitted to make the Waterford 3 parameter range consistent with ranges for other plants that utilize the CPC system.
In addition, NOTE 6 to Table 2.2-1 will be deleted. This is consistent with license amendment NPF-38-08 in which the list of CPC Addressable Constants was deleted. The CPC software has been designed with automatic acceptable input checks against limits that are specified by the CPC functional design specifica-tions. Thus, inclusion of the addressable constants and software limit value described in NOTE 6 is redundant.
NS41171
Safety Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:
1.
Will operation of the facility in accordance with this proposed change significantly increase the probability or consequences of any accident previously evaluated?
Response: No.
The proposed change to the CPC low DNBR trip setpoint has been considered in all transients that require protectinn from this trip.
In addition, those design basis accidents which use she Safety Limit setting to predict the number of fuel pins which experience DNB have been evaluated using this new limit. All Anticipated Operattonal Occurrences (A00s) result in a DNBR which remains above the Cycle 2 safety limit of 1.26.
All acci-dents which result in a DNBR less than 1.26 have been evaluated to ensure acceptable fuel performance and that the off-site doses do not exceed the guidelines specified in 10 CFR 100. Thus, the proposed change will not significantly increase the probability or consequences of any accident previously evaluated.
2.
Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not affect the logic used by the CPCs to perform their design function of protecting the core from a violation of Specified Acceptable Fuel Design Limits (SAFDLs) during an A00. Several of the uncertainties used in the CPC calculation of hot channel DNBR have been removed from the actual heat flux calculation and statistically combined with other uncertainties to determine a new DNBR Safety Limit and low -
DNBR trip setpoint. The new Safety Limit provides the same probability /
confidence level that DNB will not occur during an A00. Thus, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
Will operation of the facility in accordance with the proposed change involve a significant reduction in the margin of safety?
Response: No.
The intent of this Specification is to prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. This is accomplished by maintaining the DNBR in the hottest coolant channel in the core at a value greater than the safety limit during normal operation and all A00s.
The CPCs perform this protective function by continuously (monitoring the DNBR in the " hot channel". The changes made to the CPCs i.e., removal of some uncertainties) have been accounted for in the SCU methodology by p
increasing the DNBR Safety Limit / Trip Setpoint from 1.205 to 1.260. This limit ensures, with a 95/95 probability / confidence level, that the hot channel will not experience DNB during an A00. Thus, the proposed change will not involve a significant reduction of the margin of safety.
Safety and Significant Hazards Determination Based on the above Safety Analysis, it is concluded that:
(1) the proposed change does not constitute a significant hazards consideration as defined by 10 CFR 50.92; (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.
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e NPF-38-43 ATTACHMENT A
I
- 2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE ONSR 2.1.1.1 Tite ONBR of the reactor core shall be maintained greater than or equal to 1.20.
APPLICABILITY: MODES 1 and 2.
ACTION:
Whenever the DN8R of the reactor has decreased to less than 1.20, be in HOT STAN08Y within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the' requirements of Specification 6 7
' PEAK LINEAR' HEAT RATE 2.1.1.2 fuel shall be maintained less than or equal to 21.0 kW/ft.The peak APPLICABILITY: MODES 1 and 2.
ACTION:
Whenever the peak linear heat rate (adjusted for fuel rod dynamics) of the the requirements of Specification 6.7.1. fuel has exceeded 21.0 kW/ft, b REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.
APPLICABILITY:
MODES 1, 2, 3, 4, and 5.
ACTION:
MODES 1 and 2 in HOT STAN08Y with the Reactor Coolant System pressure w within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.
MODES 3, 4, and 5 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, and comply with the requirements of Specification 6.7.1. reduce the WATERFORD - UNIT 3 2-1 y----
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1 1
2.1 and 2.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 2.1.1 REACTOR CORE The restrictions of these safety limits prevent overheating of the fuel cladding and possible cladding perforation which 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 %ithin the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature, and (2) maintaining the dynamically adjusted peak linear heat rate of the fuel at or less than 21.0 kW/ft which will not cause fuel centerline melting in any fuel rod.
First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the maximum clad surface temperature is only slightly greater than the coolant saturation temperature.
The upper boundary of the nucleate boiling., 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.
Correlations predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB ratio (DNBR), defined as the ratio of the predicted DNB heat flux at a particular core location to the actua1 heat flux at that location, is indicative of the margin to DNB.
The minimum value of DNBR during normal operational occurrences is limited to 1.20 for the CE-1 correlation and is established as a Safety Limit.
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 br maintained only if the design and operating conditions are appropriate throughout the life of the fuel rods.
Volume changes which accompany the solid to liquid phase change are significant and require accommodation.
Another consideration involves the redistribution of the fuel which depends on the extent of the melting and the physical state of the fuel rod at the time of melting.
Because of the above factors, the steady state value of the peak linear heat rate which would not cause fuel centerline melting is established as a Safety Limit.
To account for fuel rod dynamics 1
(lags), the directly indicated linear heat rate is dynamically adjusted.
Limiting safety system settings for the Low DNBR, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High Linear Power Level trips, and limiting conditions for operation on DNBR and kW/ft margin are specified such that there is a high degree of confidence that the specified l
acceptable fuel design limits are not exceeded during normal operation and design basis anticipated operational occurrences.
WATERFORD - UNIT 3 8 2-1
SAFETY LIMITS AND LXMKTENG SAFETY SYSTEM SETTlNGS
- 2. 2 LIMITING SAFETY SYST'EM SETTINGS REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.
APPLICABILITY: As shown for each channel in Table 3.3-1.
ACTION:
With a reactor protective instrumentation setpoint les's conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION s'atement requirement of Specifica-t tion 3.3.1 until the channel is restored to OPERA 8LE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
W t
e WATERFORD - UNIT 3 2-2 AMEN 0 MENT NO.
5
TABLE 2.2-1 5
m" REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS E
Y FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 1.
Manual Reactor Trip Not Applicable Not Applicable w
2.
Linear Power Level - High Four Reactor Coolant Pumps i 110.1% of RATED THERMAL POWER 1 110.7% of RATED THERMAL POWER Operating 3.
Logarithmic Power Level - High (1) 1 0.257% of RATED THERMAL POWER i 0.275% of RATED THERMAL POWER 4
4.
Pressurizer Pressure - High 5 2365 psia i 2372 psia 5.
Pressurizer Pressure - Low 1 1684 psia (2)
> 1644 psia (2)
[
6.
Containment Pres'sure - High i 17.1 psia i 17.3 psia 7.
Steam Generator Pressure - Low
> 764 psia (3) 1 748 psia (3) 8.
Steam Generator Level - Low 1 27.4% (4) 1 26.7% (4) 9.
Local Power Density - High 1 21.0 kW/ft (5) i 21.0 kW/ft (5) 10.
DNBR - Low 1 1.205 (5)(6) 1 1.205 (5)(6) 11.
Steam Generator Level - High i 87.7% (4) 5 88.4% (4) 12.
Reactor Protection System Logic Not Applicable Not Applicable 13.
Reactor Trip Breakers Not Applicable Not Applicable 14.
Core Protection Calculators Not Applicable Not Applicable 15.
CEA Calculators Not Applicable Not Applicable 16.
Reactor Coolant Flow - Low 1 23.8 psid (7) 1 23.6 psid (7)
TABLE 2.2-1 (Continued)
REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TABLE NOTATIONS (1) Trip may be manually bypassed above 10 4% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is less than or equal to 10 4% of RATED THERMAL POWER.
(2) Value may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is reduced, provided the margin between the pressurizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached.
Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.
(3) Value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or equal to 200 psi; the setpoint shall be f
increased automatically as steam generator pressure is increased until the trip setpoint is reached.
(4) % of the distance between steam generator upper and low level instrument nozzles.
(5) As stored within the Core Protection Calculator (CPC).
Calculation of the trip setpoint includes measurement, calculational and processor uncertainties, and dynamic allowances.
Trip may be manually bypassed below 10 4% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 10 4% of RATED THERMAL POWER.
(6) The minimum allowable value of the addressable constant BERR1 in each l
OPERABLE channel is 1.146.
(7) The setpoint may be altered to disable trip function during testing pursuant to Specification 3.10.3.
WATERFORD - UNIT 3 2-4 l
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The Reactor Coolant System components are designed to Section III, 1974 Edition, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psia) of design pressure.
The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.
The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation.
2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit.
The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.
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 drift allowance assumed for each trip in the safety analyses.
The DNBR - Low and Local Power Density - High are digitally generated trip setpoints based on Limiting Safety System Settings of 1.20 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 equipment.
The Allowable Values for these trips are therefore the same as the Trip Setpoints.
To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the DNBR - Low and Local Power Density -
High trips include the measurement, calculational and processor uncertainties and dynamic allowances as defined in CEN-147(S)-P " Functional Design Specifica-tion for a Core Protection Calculator," January 1981; CEN-148(S)-P " Functional Design Specification for a Control Element Assembly Calculator," January 1981 and Software Change Package - CEN-197(c)-P "CEAC/CPC Software Modifications for Waterford-3 SES," March 1982.
WATERFORD - UNIT 3 8 2-2
SAFETY LIMITS AND LIMITlNG SAFETY SYSTEM SETTINGS BASES
~
Local Power Density - High (Continued)
The local power density (LPD), the trip variable, calculated by the CPC incorporates uncertainties and dynamic compensation routines.
These uncertainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core peak LPD is sufficiently less than the fuel design limit such that the increase in actual core peak LPD after the trip will not result in a violation of the peak LPD Safety Limit.
CPC uncertainties related to peak LPD are the same types used for DNBR calculation.
Dynamic compensation for peak LPD is provided for the effects of core fuel centerline temperature delays (relative to changes in power density), sensor time delays, and protection system equipment time delays.
DNBR - Low The DNBR - Low trip is provided to prevent the DNBR in the limiting coolant channel in the core from exceeding the fuel design limit in the event of anticipated operational occurrences.
The DNBR - Low trip incorporates a low pressurizer pressure floor of 1845 psia.
At this pressure a DNBR - Low trip will automatically occur.
This low pressure trip also provides protection against steam generator tube rupture events.
The DNBR is calculated in the CPC utilizing the following information:
a.
Nuclear flux power and axial power distribution from the excore neutron flux monitoring system; b.
Reactor Coolant System pressure from pressurizer pressure measurement; Differential temperature (Delta T) power from reactor coolant temperature c.
and coolant flow measurements; l
d.
Radial peaking factors from the position measurement for the CEAs; e.
Reactor coolant mass flow rate from reactor coolant pump speed; f.
Core inlet temperature from reactor coolant cold leg temperature l
measurements.
f l
The DNBR, the trip variable, calculated by the CPC incorporates various l
uncertainties and dynamic compensation routines to assure a trip is initiated l
prior to violation of fuel design limits.
These uncertainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core DNBR is sufficiently greater than the fuel design limit such that the decrease l
WATERFORD - UNIT 3 B 2-5 l
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
~
BASES DNBR - Low (Continued) in actual core DNBR after the trip will not result in a violation of the DNBR Safety Limit of 1.20.
CPC uncertainties related to DNBR cover CPC input measurement uncertainties, algorithm modelling uncertainties, and computer equipment processing uncertainties.
Dynamic compensation is provided in the CPC calculations for the effects of coolant transport delays, core heat flux delays (relative to changes in core power), sensor time delays, and protection system equipment time delays.
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 7 580*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
> 1845 psia f.
Pressurizer Pressure-High 7 2355 psia g.
Integrated Radial Peaking Factor-Low
> 1.28 h.
Integrated Radial Peaking Factor-High
< 4.28 i.
Quality Margin-Low
>0 Steam Generator Level - High The Steam Generator Level - High trip is provided to protect the turbine from excessive moisture carry over.
Since the turbine is automatically I
tripped when the reactor is tripped, this trip provides a reliable means for providing protection to the turbine from excessive moisture carry over.
This trip's setpoint does not correspond to a Safety Limit and no credit was taken in the safety analyses for operation of this trip.
Its functional capability at the specified trip setting is required to enhance the overall reliability l
of the Reactor Protection System.
l Reactor Coolant Flow - Low The Reactor Coolant Flow - Low trip provides protection against a reactor coolant pump sheared shaft event and a steam line break event with a loss-of-offsite power.
A trip is initiated when the pressure differential across the primary side of either steam generator decreases below a nominal setpoint of 23.8 psid.
The specified setpoint ensures that a reactor trip occurs to prevent violation of local power density or DNBR safety limits under the stated conditions.
I WATERFORD - UNIT 3 8 2-6 i
NPF-38-43 ATTACHMENT B l-I l
I
t 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE DN8R 2.1.1.1 The DN8R of the reactor core shall be maintained greater than or equal to APPLICABILITY:
MODES 1 and 2.
ACTION:
/.14 Whenever the DN8R of the reactor has decreased to less than STAN08Y within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6 71
(
, be in HOT
- PEAK LINEAR' HEAT RATE 2.1.1.2 fuel shall be maintained less than or equal to 21.0 kW/ft.Tt.e pea APPLICA8ILITY: MODES 1 and 2.
ACTION:
Whenever the peak linear heat rate (adjusted for fuel rod dynamics) of the the requirements of Specification 6.7.1. fuel has exceeded 21.0 kW/ft, b 4
REACTOR COOLANT SYSTEM PRESSURE i
2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.
APPLICA8ILITY: MODES 1, 2, 3, 4, and 5.
ACTION:
MODES 1 and 2 in HOT STAN08Y with the Reactor Coolant System pressure w within I hour, and comply with the requirements of Specification 6.7.1.
MODES 3, 4, and 5 i
Whenever the Reactor Coolant System pressure has exceeded 2750 psia, and comply with the requirements of Specification 6.7.1. reduce the l
WATERFORD - UNIT 3 2-1
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l 2.1 and 2.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i
BASES 2.1.1 REACTOR CORE The restrictions of these safety limits prevent overheating of the fuel cladding and possible cladding perforation which 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 boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature, and (2) maintaining the dynamically adjusted peak linear heat rate of the fuel at or less than 21.0 kW/ft which will not cause fuel centerline melting in any fuel rod.
First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the maximum clad surface temperature is only slightly greater than the coolant saturation temperature.
The upper boundary of the nucleate boiling 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.
i Correlations predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB ratio (DNBR), defined as the ratio of the predicted DNB 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 occurrences is limited to for the CE-1 correlation and is established as a Safety Limit.
fjg Second, operation with a peak linear heat rate below that which would l
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),
fuet rod integrity would be" maintained ~6nly if the design and operating conditions are appropriate throughout the life of the fuel rods.
Volume changes which accompany the solid to liquid phase change are significant and require accommodation. Another consideration involves the redistribution of the fuel which depends on the extent of the melting and the physical state of the fuel rod at the time of melting.
Because of the above factors, the steady state value of the peak linear heat rate which would not cause fuel centerline melting is established as a Safety Limit.
To account for fuel rod dynamics (lags), the directly indicated linear heat rate is dynamically adjusted.
Limiting safety system settings for the Low DNBR, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High Linear Power Level trips, and limiting conditions for operation on DNBR and kW/ft margin are specified such that there is a high degree of confidence that the specified acceptable fuel design limits are not exceeded during normal operation and design basis anticipated operational occurrences.
WATERFORD - UNIT 3 B 2-1
4 TABLE 2.2-1 E
h REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS 85 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 1.
Manual Reactor Trip Not Applicable Not Applicable w
2.
Linear Power Level - High Four Reactor Coolant
- Pumps 5 110.1% of RATED THERMAL POWER 1 110.7% of RATED THERMAL POWER Operating 3.
Logarithmic Power Level - High (1) $ 0.257% of RATED THERMAL POWER 5 0.275% of RATED THERMAL POWER 4.
Pressurizer Pressure - High 1 2365 psia 1 2372 psia 5.
Pressurizer Pressure - Low 1 1684 psia (2) 1 1644 psia (2) 3 6.
Containment Pressure - High 1 17.1 psia i 17.3 psia l
7.
Steam Generator Pressure - Low
> 764 psia (3) 1 748 psia (3) 8.
Steam Generator Level - Low 1 27.4% (4) 1 26.7% (4) 9.
Local Power Density - High
< 21.0 kW/ft (5)
< 21.0 kW/ft (5) 10.
DNBR - Low 1
)Mt 3
5)Mr)-
l 11.
Steam Generator Level - High
$ 87.7% (4) 5 88.4% (4) 12.
Reactor Protection System Logic Not Applicable Not Applicable 13.
Reactor Trip Breakers Not Applicable Not Applicable 1
14.
Core Protection Calculators Not Applicable Not Applicable 1
15.
CEA Calculators Not Applicable Not Applicable 16.
Reactor Coolant Flow - Low 1 23.8 psid (7) 1 23.6 psid (7) a
TABLE 2.2-1 (Continued)
REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TABLE NOTATIONS (1) Trip may be manually bypassed above 10 4% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is less than or equal to 10 4% of RATED THERMAL POWER.
(2) Value may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is reduced, provided the margin between the pressurizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached.
Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.
(3) Value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or equal to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.
(4) % of the distance between steam generator upper and low level instrument nozzles.
l l
(5) As stored within the Core Protection Calculator (CPC).
Calculation of l
the trip setpoint includes measurement, calculational and processor uncertainties, and dynamic allowances.
Trip may be manually bypassed below 10 4% of RATED THERMAL POWER; bypass shall be automatically removed l
when THERMAL POWER is greater than or equal to 10 4% of RATED THERMAL POWER.
l (6) '5: c'
- r :!!rt:51: :! : ef th: ddr::::51: ::::t:nt SEP91 '
02:5 l
v"0PCCAOLC ch;nn;I i-1.110.
(7) The setpoint mai be altered to disable trip function during testing pursuant to Specification 3.10.3.
Note 6
haS been C eIe b0cl WATERFORD - UNIT 3 2-4
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The Reactor Coolant System components are designed to Section III, 1974 Edition, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psia) of design pressure.
The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.
The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation.
2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit.
The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.
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 drift allowance assumed for each trip in the safety analyses.
The DNBR - Low and Local Power Density - High are digital generated trip setpoints based on Limiting Safety System Settings of 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 equipe.ent.
The Allowable Values for these trips are therefore the same as the Trip Setpoints.
i To maintain the margins of safety assumed in the safety analyses, the I
calculations of the trip variables for the DNBR - Low and Local Power Density -
l High trips include the measurement, calculational and processor uncertainties and dynamic allowances as defined in CEN-147(S)-P " Functional Design Specifica-tion for a Core Protection Calculator," January 1981; CEN-148(S)-P " Functional Design Specification for a Control Element Assembly Calculator," January 1981 and Software Change Package - CEN-197(c)-P "CEAC/CPC Software Modifications for Waterford-3 SES," March 1982.
I WATERFORD - UNIT 3 8 2-2
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Local Power Density - High (Continued)
The local power density (LPD), the trip variable, calculated by the CPC incorporates uncertainties and dynamic compensation routines.
These uncertainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core peak LPD is sufficiently less than the fuel design limit such that the increase in actual core peak LPD after the trip will not result in a violation of the peak LPD Safety Limit.
CPC uncertainties related to peak LPD are the same types used for DNBR calculation.
Dynamic compensation for peak LPD is provided for the effects of core fuel centerline temperature delays (relative to changes in power density), sensor time delays, and protection system equipment time delays.
DNBR - Low The DNBR - Low trip is provided to revent the DNBR in the limiting coolant channel in the core from exce ing the fuel design limit in the event of anticipated operational occurrenc s.
The DN8R - Low trip incorporates a low pressurizer pressure floor of psia.
At this pressure a DNBR - Low l
trip will automatically occur.
This low pressure trip.also provides protection against steam generator tube rupture events.
The DNBR is calculated in the CPC utilizing the following information:
a.
Nuclear flux power and axial power distribution from the excore neutron flux monitoring system; b.
Reactor Coolant System pressure from pressurizer pressure measurement; c.
Differential temperature (Delta T) power from reactor coolant temperature j
and coolant flow measurements; d.
Radial peaking factors from the position measurement for the CEAs; e.
Reactor coolant mass flow rate from reactor coolant pump speed; j
f.
Core inlet temperature from reactor coolant cold leg temperature l
measurements.
The DNBR, the trip variable, calculated by the CPC incorporates various uncertainties and dynamic compensation routines to assure a trip is initiated prior to violation of fuel design limits.
These uncertainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core DNBR is sufficiently greater than the fuel design limit such that the decrease l
WATERFORD - UNIT 3 8 2-5 l
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES DNBR - Low (Continued) f.26 in actual core DNBR fter the trip will not result in a violation of the DNBR P
Safety Limit of CPC uncertainties related to DNBR cover CPC input l
measurement uncertainties, algorithm modelling uncertainties, and computer equipment processing uncertainties.
Dynamic compensation is provided in the CPC calculations for the effects of coolant transport delays, core heat flux delays (relative to changes in core power), sensor time delays, and protection system equipment time delays.
i The DN8R 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 7 580*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
> 40 W psia f.
Pressurizer Pressure-High 7
psia N
.g.
Integrated Radial Peaking i
l Factor-Low
> l.28
^
Integrated Radial Peaking gg h.
Factor-High
< 4.28 j
f.
Quality Margin-Low
>0 l
Steam Genera _ tor Level - High The Steam Generator Level - High trip is provided to protect the turbine from excessive moisture carry over.
Since the turbine is automatically tripped when the reactor is tripped, this trip provides a reliable means for providing protection to the turbine from excessive, moisture carry over.
This trip's setpoint does not correspond to a Safety Limit and no credit was taken in the safety analyses for operation of this trip.
Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System.
Reactor Coolant Flow - Low The Reactor Coolant Flow - Low trip provides protection against a reactor coolant pump sheared shaft event and a steam line break event with a loss-of-offsite power.
A trip is initiated when the pressure differential across the primary side of either steam generator decreases below a nominal setpoint of 23.8 psid.
The specified setpoint ensures that a reactor trip occurs to prevent violation of local power density or DNBR safety limits under the stated conditions.
WATERFORD - UNIT 3 8 2-6
NPF-38-44
DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-44 This is a request to revise Technical Specification 3.2.4, " Power Distribution Limits, DNBR Margin", the associated Surveillance Requirement 4.2.4.4, and the Bases for this Specification (3/4.2.4).
Existing Specification See Attachment A.
Proposed Specification See Attachment B.
Description The proposed change would revise Technical Specification 3.2.4, " Power Distribu-tion Limits, DNBR Margin", and the associated Surveillance Requirement 4.2.4.4.
The reason for this change is that Cycle 2 core parameters and CPC software are different than they were for Cycle 1.
When COLSS is in service and at least one CEA Calculator (CEAC) is operable, the DNBR margin is maintained at an acceptable level by ensuring that the COLSS cal-culated core power is less than the COLSS calculated Power Operating Limit (POL) based on DNBR. Maintaining the core power below the DNBR-based POL ensures that the Specified Acceptable Fuel Design Limits (SAFDLS) will not be violated during an Anticipated Operational Occurrence (A00).
If neither CEAC is operable, the CPCs lack the CEA position information necessary to ensure a reactor trip when necessary.
In this case, Specification 3.2.4b requires that the COLSS calculated core power shall be maintained at 19% below the COLSS calculated power operating limit to compensate for the potential error in the CPC DNBR calculation. The proposed revision would decrease this required adjustment to 13% as a result of the reevaluation of the limiting Cycle 2 tran-sients.
If COLSS is out-of-service but at least one CEAC is operable, Specification 3.2.4c applies.
It states that the DNBR operating margin shall be maintained by compar-ing the DNBR indicated on any operable CPC channel with the allowable value from Figure 3.3-2.
In the proposed change, this figure is revised to account for the less favorable Cycle 2 core parameters and revised CPC software.
If COLSS is out-of-service and both CEAC's are inoperable, Specification 3.2.4d applies.
It states that the DNBR operating margin shall be maintained by com-paring the DNBR indicated on any operable CPC channel with the allowable value from Figure 3.2-3.
In the proposed change, this figure is revised to account for all the proposed changes to Specifications 3.2.4a, 3.2.4b and 3.2.4c, which are described above.
NS41172
The surveillance requirement described by Specification 4.2.4.4, which currently imposes a penalty as a function of burnup on the CPC calculated DNBR, will be deleted. This penalty is an allowance for rod bow and is incorporated in the proposed change to the DNBR Safety Limit and Reactor Trip Setpoint described in Technical Specifications 2.1.1.1 and 2.2.1 (see NPF-38-43).
Additionally, a typographical error is being corrected in ACTION statement (a) by replacing " linear heat rate" with "DNBR".
Safety Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:
1.
Will the operation of the facility in accordance with this proposed change significantly increase the probability or consequences of any accident previously evaluated?
Response: No.
The Cycle 2 safety analyses have shown that when COLSS is in service and at least one CEAC is operable, Specification 3.2.4a provides enough margin to DNB to accommodate the limiting A00 without violating the SAFDLs.
For the case when neither CEAC is operable but COLSS is in service, the CPCs assume a preset CEA configuration and can not obtain the required CEA position information to ensure the SAFDL on DNBR will not be violated during an A00. Thus, as a result of the reevaluation of the limiting A00s for Cycle 2, Specification 3.2.4b requires that core power be reduced to a value 13% less than the current COLSS calculated power operating limit. This ensures the limiting A00 will not result in a violation of SAFDLs. The proposed revision to Figure 3.2-2 accounts for the situation when COLSS is out-of-service but at least one CEAC is operable.
In this case, the Cycle 2 safety analysis has shown that by maintaining the CPC calculated DNBR above the value shown in the figure, the limiting A00 will not result in a violation of the SAFDLs. When COLSS is out-of-service and both CEACs are inoperable, there must be additional margin to DNB set aside in the CPCs to ensure they can mitigate the consequences of the lim-iting A00. A reevaluation of the limiting transients performed as part of the Cycle 2 safety analysis has shown that by maintaining the CPC calcu-lated DNBR above the limits shown in the proposed revision to Figure 3.2-3,
[
there is sufficient thermal margin to ensure that the limiting A00 will not result in a violation of the SAFDLs. Therefore, the proposed change will not significantly increase the probability or consequences of any accident previously evaluated.
2.
Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes are primarily a result of changes in the Cycle 2 core parameters and new approved analytical methods. There has been no physical change to the facility. All changes are either internal to the CPCs or are reflected as proposed revisions to the Technical Specifica-tions. Thus, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
Will operation of the facility in accordance with this proposed change involve a significant reduction in the margin of safety?
Response: No.
The intent of this Specification is to ensure that there is always suf-ficient margin to DNB such that the CPCs can mitigate the consequences of the most limiting A00 prior to a violation of the SAFDLs. Generally, this margin is continuously monitored by COLSS; however, if COLSS is out-of-service, the limitation on CPC calculated DNBR (as a function of ASI) shown in Figures 3.2-2 and 3.2-3 represents a conservative envelope of operating conditions consistent with the Cycle 2 safety analysis assumptions. This band of operating conditions has been analytically demonstrated to maintain an acceptable minimum DNBR throughout all A00s.
The penalty factor imposed by surveillance requirement 4.2.4.4 has been deleted because these penalties have been included in the low DNBR trip setpoint. Thus, the proposed changes will not result in a significant reduction in the margin of safety.
Safety and Significant Hazards Determination Based on the above Safety Analysis, it is concluded that (1) the proposed change does not constitute a significant hazards consideration as defined by 10 CFR 50.92; (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environ-ment as described in the NRC Final Environmental Statement.
t
4 NPF-38-44 ATTACHMENT A
l POWER DISTRIBUTION L1MITS 3/4.2.4 DNBR MARGIN LIMITING CONDITION FOR OPERATION 3.2.4 The DNBR margin shall be maintained by one of the following methods:
Maintaining COLSS calculated core power less than or equal to COLSS a.
calculated core power operating limit based on DNBR (when COLSS is in service, and either one or both CEACs are operable); or b.
Maintaining COLSS calculated core power less than or equal to COLSS calculated core power operating limit based on DNBR decreased by 19% RATED THERMAL POWER (when COLSS is in service and neither CEAC is operable); or Operating within the region of acceptable operation of Figure 3.2-2 c.
using any operable CPC channel (when COLSS is out of service and either one or both CEACs are operable); or d.
Operating within the region of acceptable operation of Figure 3.2-3 using any operable CPC channel (when COLSS is out of service and neither'CEAC is operable).
APPLICABILITY:
MODE 1 above 20% of RATED THERMAL POWER.
ACTION:
With the DNBR not being maintained:
1.
As indicated by COLSS calculated core power exceeding the appropriate COLSS calculated power operating limit; or 2.
With COLSS out of service, operation outside the region of acceptable operation of Figure 3.2-2 or 3.2-3, as applicable; within 15 minutes inititate corrective action to increase the DNBR to within the limits and either:
Restore the linear heat rate to within its limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or a.
b.
Be in at least HOT STAND 8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.4.1 The provisions of Specification ~4.0.4 are not applicable.
i 4.2.4.2 The DNBR shall be determined to be within its limits when THERMAL POWER is above 20% of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory System (COLSS) or, with the COLSS out of service, by verifying at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> that the DNBR, as indicated on any OPERABLE ONBR channels, is within the I
i limit shown on Figure 3.2-3.
4.2.4.3 At least once per 31 days, the COLSS Margin Alarm shall be verified to actuate at a THERMAL POWER level less than or equal to the core power operating limit based on DNBR.
WATERFORD - UNIT 3 3/4 2-6 AMENOMENT NO. 5
-,--,--,----,,--,,,----,w,-.
.,-,___,_-,--,an.
...w-m-
--v
_ POWER DISTRIBUTION LIMITS 1
SURVEILLANCE _ REQUIREMENTS (Continued) 4.2.4.4 the COLSS and CPC ON8R calculations at least once pe GWo BURNUP(MTU)
ON8R PENALTY (%)
0-10.0 0.50 10.0-20.0 1.00 20.0-30.0 2.00 30.0-40.0 3.50 40.0-50.0 s.50 l
i i
1 l
WATERFORD - UNIT 3 3/4 2-7 i
- + --
,,,,..,,y-
_v
,,y-,,ma,
_m,,_,y.-
..,w,%_,g-y
._.,_m,,_
5 i
COLSS OUT OF SERVICE DNBR LIMIT LINE 1.7 3
I s
ACCEPTABLE OPERATION (0.10, 1.62)
(0.24, 1.62) 1.6 Two ci m
2 Q
1.5 g
(-0.19,1.50) 3 UNACCEPTABLE OPERATION
- g....
d 1.4 l
l
\\
l e
t t
i 0.3
-0.2
-0.1-0.0-
+ 0.1
+ 0.2
+0.3 L
CORE AVERAGE ASI i
FIGURE 3.2-2 ONBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS l
(COLSS OUT OF SERVICE, CEACs OPERABLE)
~.
l WATERFORD - UNIT 3 3/4 2-8 AMEN 0 MENT NO. 5
-. - - - - - - ~ _ - -., -,.., -, _, - - -, - - - -
-r--+
-.m
COLSS OUT OF SERVICE DNBR LIMIT LINE 1.7 i
ACCEPTABLE OPERATION (0.10, 1.62) 1.6 ci m
2 O
1.5 g
(-0.19,1.50)
~
o UNACCEPTABLE 2
OP, ERA, TION 1.4
(
l 1.3'
-0.3
-0.2'
-0.1 0.0
+ 0.1
+ 0.2
+ 0.3 CORE AVERAGE ASI FIGURE 3.2-3 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE, CEACs INOPERABLE).
~.
WATERFORD - UNIT 3 3/4 2-9 AMEN 0 MENT NO. 5
.. _.. _ _ _. _ _ _ _ -, _, _.,... _ _.,.. _, ~ _ _,. -,,,, _ _. _ -. _,,. _, - -. _ -.. -. _ _ _ _ _.. _ - _ _ -,., _..,,,. -
POWER DISTRfBUTTON LfMITS BASES AZIMUTHAL POWER TILT - Tq (Continued)
P
/P is the ratio of the power at a core location in the presence tilt untilt of a tilt to the power at that location with no tilt.
3/4.2.4 DNBR MARGIN The limitation on DN8R as a function of AXIAL SHAPE INDEX represents a conservative envelope of operating conditions consistent with'the safety analysis assumptions and which have.been analytically demonstrated adequate to maintain an acceptable minimum DNBR throughout all anticipated operational occurrences, of which the loss of flow transient is the most limiting.
Operation of the core with a DN8R at or above this limit provides assurance that an acceptable minimum DN8R will be maintained in the event of a loss of flow transient.
Either of the two core power distributionr monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the DN8R channels in the Core Protection Calculators (CPCs), provides adequate monitoring of the core power distribution and is capable of verifying that the DN8R does not violate its limits.
The COLSS performs this function by continuously monitoring the core power distribution and calculating a core operating limit corresponding to the allowable minimum DN8R.
The COLSS calculation of core power operating limit based on the minimum DN8R limit includes appropriate penalty factors which provide a 95/95 probability / confidence level that the core power calculated by COLSS, based on the minimum DN8R limit, is conservative with respect to the actual core power limit.
These penalty factors are determined from the uncer-tainties associated with planar radial peaking measurements, engineering heat flux, state parameter measurement, software algorithm edelling, computer processing, rod bow, and core power measurement.
W Parameters required to maintain the margin to DN8 and total core power are also monitored by the CPCs.
Therefore, in the event that the COLSS is not being used, operation within the limits of Figure 3.2-3 can be maintained by utilizing a predetermined DNBR as a function of AXIAL SHAPE INDEX and by monitoring the CPC trip channels.
The above listed uncertainty and penalty factors plus those associated with startup test acceptance criteria are also included in the CPCs which assume a minimum core power of 20% of RATED THERMAL POWER.
The 20% RATED THERMAL POWER threshold is due to the neutron flux detector system being less accurate below 20% core power.
Core noise level at low power is too large to obtain usable detector readings.
WATERFORD - UNIT 3 8 3/4 2-3 AMEN 0 MENT NO. 5 i
i L -
I POWER DISTRIBUTION LIMITS I
BASES ON8R MARGIN (Continued)
J The DNBR penalty factors listed in Specification 4.2.4.4 are penalties used to ac'commodate the affects of rod bow.
The amount of rod bow in each assembly is dependent upon the average burnup experience'd'by that assembly.
Fuel assemblies that incur higher average burnup will experience a greater magnitude of rod bow.
Conversely, lower burnup assemblies will experience less rod bow.
The penalty for each batch required to compensate for rod bow is determined from a batch's maximum average assembly burnup applied to the batch's maximum integrated planar radial power peak.
A single not penalt9 for COLSS and CPC is then determined from the penalties associated with each batch, accounting for the offsetting margins due to the lower radial power peaks in the higher burnup batches.
3/4.2.5 RCS FLOW RATE This specification is provided to ensure that the actual RCS total flow rate is maintained at or above the minimum value used in the LOCA safety analyses, and that the DN8R is maintained within the safety limit for Anti-I cipated Operational Occurrences (A00).
3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE This specification is provided to ensure that the actual value of reactor coolant cold leg temperature is maintained within the range of values used in the safety analyses, with adjustment for instrument accuracy of 22*F, and that the peak linear heat generation rate and the moderator temperature coefficient effects are validated.
3/4.2.7 AXIAL SHAPE INDEX This specification is provided to ensure that the actual value of AXIAL SHAPE INDEX is maintained within the range of values used in the safety analyses, to ensure that the peak linear heat rate and DN8R remain within the safety limits for Anticipated Operational Occurrences (A00).
3/4.2.8 PRESSURIZER PRESSURE This specification is provided to ensure that the actual value of pressurizer pressure is maintained within the range of values used in the safety analyses.
The inputs to CPCs and COLSS are the most limiting.
The values are adjusted for an instrument accuracy of i 25 psi.. The sensitive events are SGTR, LOCA, FWLB and loss of condenser vacuum to initial high pressure, and MSLB to initial low pressure.
WATERFORD - UNIT 3 8 3/4 2-4
4-a m
9 9
NPF-38-44 ATTACHMENT B i
i i
POWER DISTRIBUTION LIMITS 3/4.2.4 DNBR MARGIN LIMITING CONDITION FOR OPERATION 3.2.4 The ONBR margin shall be maintained by one of the following methods:
Maintaining COLSS calculated core power less than or equal to COLSS a.
calculated core power operating limit based on DNBR (when COLSS is in service, and either one or both CEACs are operable); or b.
Maintaining COLSS calculated core power less than or equal to COLSS calculated core power operating limit based on DNBR decreased by 495 RATED THERMAL POWER (when COLSS is in service and neither CEAC j
13 I is operable); or
}
c.
Operating within the region of acceptable operation of Figure 3.2-2 using any operable CPC channel (when COLSS is out of service and either one or both CEACs are operable); or d.
Operating within the region of acceptable operation of Figure 3.2-3 using any operable CPC channel (when COLSS is out of service and neither'CEAC is operable).
APPLICA8ILITY:
MODE 1 above 20% of RATED THERMAL POWER.
ACTION:
With the DN8R not being maintained:
1.
As indicated by COLSS calculated core power exceeding the appropriate COLSS calculated power operating limit; or 2.
With COLSS out of service, operation outside the region of acceptable operation of Figure 3.2-2 or 3.2-3, as applicable; i
within 15 minutes inititate corrective action to increase the DNBR to within the limits and either:
hM6A a.
Restore the l' ::r 5::t r:t: to within its limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or i
b.
Be in at least H0T STAN08Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.4.1 The provisions of Specification 4.0.4 are not applicaole.
4.2.4.2 The ONBR shall be determined to be within its limits when THERMAL POWER is above 20% of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory System t
l (COLSS) or, with the COLSS out of service, by verifying at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> that the DNBR, as indicated on any OPERABLE ONBR channelf, is within the limit shown on Figure -2.2 3. 3.1-4 or Fgve 3.J 5.
4.2.4.3 At least once per 31 days, the COLSS Margin Alarm shall be verified to actuate at a THERMAL POWER level less than or equal to the core power operating limit based on DNBR.
WATERFORD - UNIT 3 3/4 2-6 AMENOMENT NO. 5 5
-,,---,_-e,.-.,,
_m.,
m,,
.m,_w
..-.,m_,m
--,cr----
ww e-ve w e r--
T
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) d. 0. " '
'5.; ';1 ? ;. ' ; O!""' ;;.;1 t 3 ';;;;c; ;t.;1 the COLSS and CPC ON8R calculations at least once per 31 days:t;.;r!;; t; t; ic, BUR MTU)
DN8R P ALTY (%)
-10.0 0.50 10.0-20.
1.00 20.0-
.0
- 2. 0 30.0-40.0 3.50
'0.0 50.0 This pge inlensionall left blank y
l l
l l
WATERTORD - UNIT 3 3/4 2-7
,m.,-
--,-----.-_-.m
_,,s-.
--7.
--,.r-
f wHb nea Ngum SAcan on ' ' '
L epkce COLSS OUT OF SERVICE DNBR LIMIT LINE 1.7 1
1 i
ACCEPTABLE OPERATION (0.10, 1.62)
(0.24, 1.62) 1.6 7
uJu nie 20 1.5 g
(-0.19,1.50) 3 NACC TABLE OP ATION _
E.. _ _
i 1.4
.3
-0.3 2
-0.1 00-
+ 0.
+ 0.2
+0.3 l
CORE AVERAGE ASI FIGURE 3.2-2 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE, CEACs OPERABLE)
WATERFORD - UNIT 3 3/4 2-8 AMENDMENT NO. 5 6
.,,---,_,.,-,,.----.,,.__e
. -, -, -, - - + -. - - - -
w-----
-y-as
COLSS OUT OF SERVICE ONBR LIMIT LINE 2.5 8
8 2.4 ACCEPTABLE OPERATION O em bl' Ininirnum f CEAC f
,=8 (0.0.2.20)
= 2.2
[
(.22,2.20)
D 2.1
~
UNACCEPTABLE OPERATION
(.17,2.07) 2.0 l
I i*
i i
e g,9 0.1 0.0 0.1 0.2 0.3 0.2
-0.3 CORE AVERAGE ASI FIGURE 3.2-2 DN8R MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE,_CEACs OPERABLE _)
l WATERFORD - UNIT 3 3/4 2-8
,__----__._,..e____.-,.<-_.m,
,y.
..-.7... -.
____,,.y.,_me.,._,._.-___%__,,,,,..,-----m...
(?ephce wil4 nea Sjure Sl COLSS OUT OF SERVICE DNBR LIMIT LINE 1.7 ACCEPTABLE OPERATION (0.10, 1.62)
(0.24, 1.62) 1.6 7
U ti m
2 O
1.5 3
(-0.19,1.50)
~
s AC PTABLE 2
P RATION 1.4 1.3
-0.3
-0.2
-0.1 0.0
+ 0.1
+ 0.2
+ 0.3 CORE AVERAGE ASI FIGURE 3.2-3 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE, CEACs INOPERABLE).
WATERFORD - UNIT 3 3/4 2-9 AMEN 0HENT NO. 5
\\
COLSS OUT OF SERVICE DNBR UMIT LINE 2.8 ACCEPTABLE 2.7 OPERATION CE/)Cs Laoferable (0.0,2.6) 2.6 (0.22,2.6) cc I
g 2.5 E
UNACCEPTABLE E
OPERATION u
U 2.4
~
(-0.17,2.4) l t
f f
l 2.3
-0.3
-0.2
-0.1
- 0. 0 0.1 0.2 0.3 CORE AVERAGE ASI FIGURE 3.2-3 DN8R MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULA (COLSS OUT OF SERVICE, CEACs IN0PERA8LE).
- WATERFORD - UNIT 3 3/4 2-9
POWER OfSTRIBUTION LlMITS BASES AZIMUTHAL POWER TILT - T, (Continued)
P
/P tilt untilt is the ratio of the power at a core location in the presence of a tilt to the power at that location with no tilt.
3/4.2.4 DN8R MARGIN The limitation on DNBR as a function of AXIAL SHAPE INDEX represents a conservative envelope of operating conditions consistent with*the safety analysis assumptions and which have.been analytically demonstrated adequate to maintain an acceptable minimum DN8R throughout all anticipated operational occurrences.
f.;hich th: ?;;; :f fir.: tnn;i::t i: th; :::t 'initing.
Operation I
of the core with a DN8R at or above this limit provides assurance that an acceptable minimum DN8R will be maintained.f-th: r;;;t f : 1 : = O f '? r.:
trri::t.
l Either of the two core power distributionrmonitoring systems, the Core Operating Limit Supervisory System (COLSS) and the DN8R channels in the Core Protection Calculators (CPCs), provides adequate monitoring of the core power distribution and is capable of verifying that the DN8R does not violate its i
limits.
The COLSS performs this function by continuously monitoring the core power distribution and calculating a core operating limit corresponding to the allowable minimum DN8R.
The COLSS calculation of core power operating limit based on the minisaum DN8R limit includes appropriate penalty factors which provide a 95/95 probability / confidence level that the core power calculated by COLSS, based on the minimum DN8R limit, is conservative with respect to the actual core power limit, These penalty factors are determined from the uncer-tainties associated with planar radial peaking measurements, : ;i= # :; but Mtne, state parameter measurement, software algorithm m? elling, computer d
processing, rod bow, and core power measurement.
G Parameters required to maintain the margin to ON8 and total core power i
are also monitored by the CPCs.
Therefore, in the event that the COLSS is not being used, operation within the limits of Figure 3.2-3 can be maintained by utilizing a predetermined DNBR as a function of IAL SHAPE INDEX and by f
monitoring the CPC trip channels.
The above lis ed uncertainty and penalty factors plus those associated with startup test acceptance criteria are also i
included in the CPCs which assume a minimum co e power of 20% of RATED THERMAL i
POWER.
The 20% RATED THERMAL POWER threshol is due to the neutron flux detector system being less accurate below 2 core power.
Core noise level at low power is too large to obtain usable dets tor readings.
)
- 3. 2. 2 or Fyn WATERFORD - UNIT 3 8 3/4 2-3 AMEN 0 MENT NO. 5 i
,....m.
._,.m.--._._..-._
POWER DISTRIBUTION LIMITS I
BASES
,/
ONBR MARGIN (Continued) has been included in M8 C0455 4"d g
CPL bN6R cakula+ ions he ONBR penalty f torg ?i;ted in 4;;ificeth a t.2.t.i cr; p;n:1tf::
-weed-to ac'commodate th effects of rod bow. The amount of rod bow in each assembly is dependen upon the average burnup experience'd 'by that assembly.
1 Fuel assemblies tha incur higher average burnup will experience a greater magnitude of rod w.
less rod bow.
Conversely, lower burnup assemblies will experience penalty for each batch required to compensate for rod bow is determined from a batch's maximum average assembly burnup applied to the l
batch's maximum integrated planar radial power peak.
A single net penalty for COLSS and CPC is then determined from the penalties associated with each batch, accounting for the offsetting margins due to the lower radial power peaks in the higher burnup batches.
3/4.2.5 RCS FLOW RATE This specification is provided to ensure that the actual RCS total flow rate is maintained at or above the minimum value used in the LOCA safety analyses, and that the DN8R is maintained within the safety limit for Anti-cipated Operational Occurrences (A00).
3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE This specification is provided to ensure that the actual value of reactor coolant cold leg temperature is maintained within the range of values used in i
the safety analyses, with adjustment for instrument accuracy of 22*F, and that the peak linear heat generation rate and the moderator temperature coefficient effects are validated.
3/4.2.7 AXIAL SHAPE INDEX This specification is provided to ensure that the actual value of AXIAL SHAPE INDEX is maintained within the range of values used in the safety analyses, to ensure that the peak linear heat rate and ON8R remain within the safety limits for Anticipated Operational Occurrences (A00).
3/4.2.8 PRESSURIZER PRESSURE This specification is provided to ensure that the actual value of pressurizer pressure is maintained within the range of values used in the safety analyses.
The inputs to CPCs and COLSS are the most limiting.
The values are adjusted for an instrument accuracy of 2 25 psi.. The sensitive events are SGTR, LOCA, FWLB and loss of condenser vacuum to initial high pressure, and MSLB to initial low pressure.
I WATERFORD - UNIT 3 8 3/4 2-4
_m d
NPF-38-45 l
l
DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-45 This is a request to revise Technical Specification 3.1.2.9, " Reactivity Control Systems, Boron Dilution", Surveillance Requirement 4.1.2.9.4, and the associated Bases section (3/4.1.2.9).
Existing Specification See Attachment A.
Proposed Specification See Attachment B.
Description The proposed change would revise Technical Specification 3.1.2.9 " Reactivity Control Systems, Boron Dilution", Surveillance Requirement 4.1.2.9.4 and the associated Bases section (3/4.1.2.9). The reasons for this change are:
(1) the Cycle 2 core will have higher enriched fuel and is therefore more reactive than the Cycle 1 core; (2) the Shutdown Margin for Cycle 2 is lower than it was for Cycle 1 (when all CEAs are inserted); and (3) it is desirable to have more than one charging pump operable when the reactor is in Mode 5 and the RCS is partially drained. Specifically, the proposed change will allow the use of 2 charging pumps when filling the RCS as long as the k-eff is maintained at a value less than 0.96.
Specification 3.1.2.9b currently requires removing power to two charging pumps when the reactor is in Mode 5 and the RCS is partially drained. The proposed change would replace this Specification with statements that allow more than one charging pump to be operable depending on the multiplication factor in the core.
That is, if the k-eff is between 0.94 and 0.96 it is permissible to have 2 charg-ing pumps operable or, if the k-eff is less than 0.94, it is permissible to have all 3 charging pumps operable.
In addition, Table 3.1-1 will be replaced with a series of Tables that provide the required boron sampling frequency as a function of the core multiplication factor that must be adhered to whenever the boron dilution alarm (s) is not operable.
By monitoring the boron concentration at these frequencies, the operators will have sufficient time to mitigate a boron dilution event prior to the loss of shutdown margin.
Safety Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:
1.
Will the operation of the facility in accordance with this proposed change significantly increase the probability or consequences of any accident previously evaluated?
Response: No.
NS41180
This Specification is provided to ensure the operators have sufficient time, from when they are first alerted to a potential boron dilution to take the' appropriate corrective action to mitigate the event. Normally, protection against this event is provided by two redundant alarms that actuate when the existing neutron flux doubles. With one or both of these alarms inoperable, the Cycle 2 safety analyses have shown that by monitoring the RCS boron concentration at the frequencies shown in-Tables i
3.1-1 through 3.1-5 the operators have sufficient time to take the actions necessary to mitigate the event. Since this Specification applies only to the Baron Dilution event, and the Cycle 2 Safety Analyses have shown that the consequences of this event are acceptable, the proposed change will not significantly increase the probability or consequences of any acci-dent previously evaluated.
4 l
2.
Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
)
Response: No.
l The proposed change is primarily a result of changes in the Cycle 2 core parameters and the desire to use more than one charging pump to fill the RCS following a refueling or following any maintenance that requires the the RCS to be partially drained. There has been no physical change to 4
the plant other than to allow an additional charging pump (s) to be oper-able if the core multiplication factor is low enough. The only accident that could be caused by an additional charging pump in operation is a boron dilution which has already been shown to have acceptable results.
Thus, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
.3.
Will operation of the facility in accordance with this proposed change involve a significant reduction in the margin of safety?
)
Response
No.
i.
The intent of this Specification is to prevent a boron dilution event or to prevent a loss of shutdown margin should a boron dilution event occur.
]
Normally, this event is precluded by isolating the primary makeup water or by the operability of the high neutron flux alarms which alert the oper-ator with sufficient time to take corrective action. The action state-ments of this Specification provide an alternate means to detect a boron dilution event by monitoring the RCS boron concentration to detect any changes. The frequencies specified in Table 3.1-1 through 3.1-5 provide the operator with sufficient time to recognize a decrease in the RCS baron concentration and take the appropriate corrective action prior to the loss of shutdown margin. More frequent checks of the RCS boron concentration 1
l are required when more charging pumps are operable or when there is a j
higher core multiplication factor because there is less time available for the operators to take corrective action. Thus, the proposed change does not result in a significant reduction in the margin of safety.
3 i
i 4
tk_._,______,_.-,__._,,_-,.._._.,______--__-.__-___--___-_-__.
Safety and Significant Hazards Determination Based on the above Safety Analysis, it is concluded that:
(1) the proposed change does not constitute a significant hazards consideration as defined by 10 CFR 50.92; (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.
4 i
-,-,.-,-----.w.-,--c------
,r,,-
- -. - - - - - - - - - -, - - - _ - _ -. - ~. -
v 4~-----
.e--.
NPF-38-45 ATTACHMENT A i
I i
1 i
k i
1
s l
-s INDEX LIST OF TABLES TABLE PAGE 1.1 FREQUENCY N0TATION.......................................
1-9
- 1. 2 OPERATIONAL M0 DES.~.......................................
1-10 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS.....................................
2-3 2.2-2 CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS.........
2-5 3.1-1 MONITORING FREQUENCIES FOR BORON DILUTION DETECTION.....
3/4 1-17 3.3-1 REACTOR PROTECTIVE INS TRUMENTATION......................
3/4 3-3 3.3-2 REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES.......
3/4 3-8 4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................
3/4 3-10 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION..........................................~3/4 3-14 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES.............................
3/4 3-19 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES...............
3/4 3-22 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEII. LANCE REQUIREMENTS...............
3/4 3-25 3.3-6 RADIATION MONITORING INSTRUMENTATION....................
3/4 3-29 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............................
3/4 3-32 3.3-7 SEISMIC MONITORING INSTRUMENTATION......................
3/4 3-36 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............................
3/4 3-37 3.3-8 HETEOR0 LOGICAL MONITORING INSTRUMENTATION...............
3/4 3-39 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION j
SURVEILLANCE REQUIREMENTS...............................
3/4 3-40 l
l 3.3-9 REMOTE SHUTDOWN INSTRUMENTATION.........................
3/4 3-42 WATERFORD - UNIT 3 XX
_ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~
_ _ _ _ _ _ _ _ ~. _ _ _ _ _, _ _...
l REACTIVITY CONTROL SYSTEMS
~
80RON DILUTION LINITING CONDITION FOR OPERATION i
l 3.1.2.9 Boron concentration shall be verified consistent with SHUT 00WN MARGIN requirements of Specifications 3.1.1.1, 3.1.1.2, and 3.9.1.
Boron dilution events shall be precluded by:
Either two boron dilution alarms (startup channel high neutron flux) a.
i shall be OPERA 8LE with the alams set in accordance with Specifica-tion 4.1.2.9.5 or the primary makeup water flow path to the Reactor Coolant System shall be isolated, and j
b.
Removing power to at least two charging pumps in MODE 4 with reactor coolant loops not filled.
APPLICA81LITY: ' MODES 3, 4, 5, and 6.
)
ACTION:
With the boron concentration not consistent with required SHUT 00WN a.
MARGIN, initiate emergency boration.
b.
With one boron dilution alarm inoperable and the primary makeup water flow path to the Reactor Coolant System not isolated, determine Reactor Coolant System boron concentration within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least at the monitoring frequency specified in Table 3.1-1.
With both boron dilution alarms inoperable and the primary makeup c.
water flow path to the Reactor Coolant System not isolated, determine the Reactor Coolant System boron concentration by two independent
.i l
means within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least at the monitoring frequency specified 1
in Table 3.1-1; otherwise, immediately suspend all operations involving positive reactivity changes or CORE ALTERATIONS (if applicable).
d.
With power applied to more than one charging pump in MODE 5 with the reactor coolant loops not filled, immediately remove power from charging pumps to comply with the above requirement or isolate the primary makeup water flow path to the Reactor Coolant System.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
e.
SURVEILLANCE REQUIREMENTS 4.1.2.9.1 The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 from MODE 2.
WATERFORD - UNIT 3 3/4 1-15
.,,,,-.,,mwwe_m._
_,,,,,,e-..-_w,,,,_
--.n
1 REACTIVITY CONTROL SYSTEMS 9
SURVEILLANCE REQUIREMENTS (Continued) 4.1.2.9.2 Each required boron dilution alare shall be demonstrated OPERA 8LE by the performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a CHANNEL FUNCTIONAL TEST at least once per 31 days, and a CHANNEL CALIBRATION at least once per 18 months.
4.1.2.9.3 The required primary makeup water flow path to the Reactor Coolant System shall be verified to be isolated by either locked closed manual valves, deactivated automatic valves secured in the isolation position, or by power being removed from all charging pumps, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.1.2.9.4 At least two charging pumps shall be verified to have power removed when required in MODE 5 with the reactor coolant loops drained at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.1.2.9.5 Each required boron dilution alare setpoint shall be adjusted to less than or equal to twice (2x) the existing neutron flux (cps) at the follow-ing frequencies:
s a.
At least once per S hours if the reactor has been shut do' n less w
than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />; b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the reactor has been shut down greater than or equal to 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> but less than 7 days; c.
At least once per 7 days if the reactor has been shut down greater than or equal to 7 days.
1 l
WATERFORD - UNIT 3 3/4 1-16
TABLE 3.1-1 MONITORING FREQUENCIES FOR BORON DILUTION DETECTION OPERATIONAL NUM8ER OF OPERABLE CHARGING PUMPS
- MODE O
1 2
3 3
24 hr 10 hr 4 hr 3 hr 4
24 hr 8 hr 4 hr 2 hr 5
8 hr 3 hr 1 hr 0.5 hr 5
8 hr 1 hr Operation not' Operation not (System drained allowed **
allowed **
for repairs) 6 24 hr 4 hr 2 hr I hr
- Charging pump OPERA 8ILITY for any period of time shall constitute OPERA 8ILITY for the entire monitoring frequency.
- In MODE 5 with the system drained for repairs, at least two charging pumps shall be verified to be inoperable by racking out their motor circuit breakers.
l i
WATERFORD - UNIT 3 3/4 1-17 l
t i
l REACTIVITY CONTROL SYSTEMS t-
'l BASES 80 RATION SYSTEMS (Continued)
The contained water volume limits include allowance for water not availaole because of discharge line location, instrument tolerances, and other physical characteristics.
The OPERA 8ILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.
The lower limit on the contained water volume, the specified baron concen-tration, and the physical size (approximately 600,000 gallons) of the RWSP also ensure a pH value of between 7.0 and 11.0 for the solution recirculated within l
containment after a LOCA.
This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosien on mechanical systems and components.
i The maximum limit on the RWSP temperature ensures that the assumptions used in the containment pressure analysis under design base accident condi-tions remain valid and avoids the possibility of containment overpressure.
The minimum limit on the RWSP temperature is required to prevent freezing and/
or baron precipitation in the RWSP.
3/4.1.2.9 BORON DILUTION This specification is provided to. prevent a boron dilution event, and to prevent a loss of SHUTDOWN MARGIN should an inadvertent boron dilution event occur.
Due to boron concentration requirements for the RWSP and boric acid l
makeup tanks, the only possible baron dilution that would remain undetected by the operator occurs from the primary makeup water through the CVCS system.
Isolating this potential dilution path or the OPERA 8ILITY of the startup t
channel high neutron flux alarms, which alert the operator with sufficient time available to take corrective action, ensures that no loss of SHUT 00WN MARGIN and unanticipated criticality occur.
The requirement to remove power on two charging pumps in M00E 5 with the reactor coolant loops drained is necessary because there is insufficient time for operator response in the i
event two or three charging pumps inject unborated water into the RCS in this condition.
The ACTION requirements specified in the event startup channel high neutron flux alarms are inoperable provide an alternate means to detect boron dilution by monitoring the RCS boron concentration to detect any changes.
The l
frequencies specified in Table 3.1-1 provide the operator sufficient time to recognize a decrease in boron concentration and take appropriate corrective l
action without loss of SHUTDOWN MARGIN.
More frequent checks are required with more charging pumps in operation due to the higher potential boron dilu-tion rate.
The surveillance requirements specified provide assurance that the startup i
channel high neutron flux alarms remain OPERA 8LE and that required valve and
]
electrical lineups remain in effect.
3/4.1.3 MOVA8LE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is WATERFORD - UNIT 3 8 3/4 1-3
-m-6..a.a A
NPF-38 45 ATTACHMENT B t
I
INDEX LIST OF TABLES TABLE PAGE 1.1 FREQUENCY N0TATION.......................................
1-9
- 1. 2 OPERATIONAL M00ES........................................
1-10 g
2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS...................................................
2-3 D
2.2-2 CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS.~........
2-5
.11 MONITORING FREQUENCIES FOR BORON DILUTION DETECTION.....
3/0 1 17
}
3.3-1 REACTOR PROTECTIVE INSTRUMENTATION......................
3/4 3-3 3.3-2 REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES.......
3/4 3-8 4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................
3/4 3-10 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.........................................
3/4 3-14 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES.............................
3/4 3-19 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES...............
3/4 3-22 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............
3/4 3-25 3.3-6 RADIATION MONITORING INSTRUMENTATION....................
3/4 3-29 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............................
3/4 3-32 3.3-7 SEISMIC MONITORING INSTRUMENTATION......................
3/4 3-36 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............................
3/4 3-37 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION...............
3/4 3-39 l
l 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION l
SURVEILLANCE REQUIREMENTS...............................
3/4 3-40 3.3-9 REMOTE SHUTDOWN INSTRUMENTATION.........................
3/4 3-42 WATERFORD - UNIT 3 XX
INSERT a
1 I
.i 3
3.1-1 K, g g > 0. 98......................................... 3 /4 1-17
]'
3.1-2
- 0. 98 2 K,g g > 0. 9 7.................................. 3 /4 1-17a 3.1-3 0.97 2 Kegg > 0.96.'.................................
3/4 1-17b 3.1-4 0.96 2 Ke f f > 0. 9 5.................................. 3 /4 1 - 17 c 3.1-5 K g g 5 0. 95.........................................
3 /4 1 -17 d e
4 i.
4 e
I e
4 T
J 4
s e
e
.m.
.m.
REACTIVITY CONTROL SYSTEMS BORON DILUTION LIMITING CONDITION FOR OPERATION 3.1.2.9 Baron concentration shall be verified consistent with SHUTDOWN MARGIN requirements of Specifications 3.1.1.1, 3.1.1.2, and 3.9.1.
Boron dilution events shall be precluded by:
Either two baron dilution alarms (startup channel high neutron flux) a.
shall be OPERA 8LE with the alarms set in accordance with Specifica-tion 4.1.2.9.5 or the primary makeup water flow path to the Reactor Coolant System shall be isolated, and 5.
i n.' ; ;:n r t: :t !:::t in chcr;f ; ;7: *
"^^L5 efth :::ter hNSE 2" "1 22?; 220 2i APPLICA8ILITY: ' MODES 3, 4, 5, and 6.
ACTION:
With the boron concentration not consistent with required SHUTDOWN a.
MARGIN, initiate emergency boration.
b.
With one boron dilution alarm inoperable and the primary makeup water flow path to the Reactor Coolant System not isolated, determine Reactor Coolant System boron concentration within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ar.d at least at the monitoring frequency specified in Tables 3.1-OM
- 3. I-S With both boron dilution alarms inoperable and the primary makeup c.
. water flow path to the Reactor Coolant System not isolated, determine i
the Reactor Coolant System boron concentration by two independent l
g_
means within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least at the monitoring frequency specified
.fka in Table 53.1-Il otherwise, immediately suspend all operations involving i
3y positive rea_ctivity changes _qr CORE ALTERATIONS (if app _licabi (fiere With ;:nr :;guirensents of SWfse%s 3. A59b end's.I 2 9c not S* i'S*dr d.
,li:d t: nr: th:2 x: :hcr;in; ;- ; '
"a": 5.Jth th:
r:::t:r x i s t '::;; x t fill d, immediately remove power from charging pumps to comply with the above requirement or isolate the primary makeup water flow path to the Reactor Coolant System.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
e.
SURVEILLANCE REQUIREMENTS t
l 4.1.2.9.1 The provisions of Specification 4.0.4 are not applicable for entry l
into MODE 3 from MODE 2.
l WATERFORD - UNIT 3 3/4 1-15
INSERT b.
1.
When in Mode 5 with reactor coolant loops not filled and 0.98 tKeggt 0.96 isolate and remove power to at least two charging pumps; or 2.
When in Mode 5 with reactor coolant loops not filled and 0.96 2Kefft 0.94 isolate and remove power to at least one charging pump; or 3.
When in Mode 5 with reactor coolant loops not filled and three charging pumps are operable, maintain K gg <0.94; and e
c.
When in Mode.6, isolate and remove power to at least two charging Pumps i
i
,----n-----
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.1.2.9.2 Each required boron dilution alarm shall be demonstrated OPERABLE by the performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a CHANNEL FUNCTIONAL TEST at least once per 31 days, and a CHANNEL CALIBRATION at least once per 18 months.
4.1.2.9.3 The required primary makeup water flow path to the Reactor Coolant System shall be verified to be isolated by either locked closed manual valves, deactivated automatic valves secured in the isolation position, or by power being removed from all charging pumps, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.1.2.9.4
^t 1:::t t= c5cr;2q ; m : :5:!' 5: ;;r:d t: 5:;f ;n;;r 7:::;:d 2
ch:r 7:
'r:d '
."^^5 5
ith th: r:::t:r :::?;nt 1::;; dr: 3:d :t ?:::t ::::
- r 2' 5
- :r:.
4.1.2.9.5 Each required boron dilution alarm setpoint shall be adjusted to less than or equal to twice (2x) the existing neutron flux (cps) at the follow-ing frequencies:
a.
At least once per 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> if the reactor has been shut down less than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />; b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the reactor has been shut down greater than or equal to 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> but less than 7 days; c.
At least once per 7 days if the reactor has been shut down greater than or equal to 7 days.
The ceguirements of 5p' eiRcahon 3.1. 2. 9b an d 3.1..?.9c shall he veri fed af least once fer.29 hows when in Mode S pHh Me reacio< coolen t loops not Alled or when in rHode fo.
~
l
(
WATERFORD - UNIT 3 3/4 1-16
TABLE 3.1-1 MONITORING FREQUENCIES FOR BORON DILUTION DETECTION OPERATIONAL NUMBER OF OPERABLE CHARGING PUMP MODE O
1 2
3 3
hr 10 hr 4
3 hr 4
24 hr 8 hr 4 hr 2 hr 5
8 hr 1 hr 0.5 hr 5
8 hr 1 hr peration not
- Operation not (System drained a
d**
allowed **
for repairs) 6 24 hr 4 hr 2 hr 1 hr
- Charging p OPERABILITY for any period of time shall constitute ERA 8ILITY for the tire monitoring frequency.
- I DE 5 with the system drained for repairs, at least two charging pumps hall be verified to be inoperable by racking out their motor circuit breaker gepke # s T a le ugh +he foIlooing s T4les WATERFORD - UNIT 3 3/4 1-17
TABLE 3,1-1 REQUIRE 0 NONITORING FREQUENCIES FOR SACKUP BORON DILUTION DETECTION A5 A FUNCTION OF OPERATING CHARGING PUNPS AND PLANT OPERATIONAL 20E5 FOR K,ff GREATERrTHAN 0.98 Ker; 7 0.92 l
OPERATIONAL Number of Operatine Chareine Pumos#
MODE 0
1 2
3 i
3 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 0.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> Operation not allowed "
4 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 0.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> Operation not allowed 5 RCS filled 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> Operation not allowed **
5 RCS partially 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Operation not allowed **
drained 6
Operation not allowed *
- l
- ' Charging pump OPERABILITY for any period of time shall constitute OPERABILITY for the entire monitoring frequency.
l ke refaired Charging fumfS Skll be Vefi0ed $O be inofemble by Ackbij out theie mofor circuif bmders.
'~
WAIE/foAD - t/N/T 3
7 2
t f
TABLE 3.1-2 REQUIRED MONITORING FREQUENCIES FOR BACKUP 80R0N DILUTION DETECTION AS A FUNCTION OF OPERATING CHARGING PUMPS AND PLANT OPERATIONAL MODES FOR K,ff GREATER THAN 0.97 AND LESS THAN OR EQUAL 70 0.98 O. 98 2 Keff > 0. 9 7 OPERATIONAL Number of Doeratine Chareine Punos*
MODE 0
1 2
3 3
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 0.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> Operation' not allowed **
l 4
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 0.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> l
5 RCS filled 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 0.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 5 RCS partially 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Operation not allowed **
drained l
t 6
Operation not allowed *"
- Charging pump OPERABILITY for any period of time shall constitute OPERA 8ILITY for the entire monitoring frequency.
reguired chcrying fumps shell be verihed k be Inoperable by racWoy Tlte Cuf 1 heir n104ce circur+ bren gerS 5;ht-I /?g tJA n?RFoo an n 3
- w
y TABLE 3.1-3 REQUIRED MONITORING FREQUENCIES FOR BACKUP OORON DILUTION DETECTION A5 A FUNCTION 0F OPERATING CHARGING PUMPS AND PLANT OPERATIONAL MODES FQR K,ff_ GREATER THAN 0.96 AND LESS THAN OR EQUAL TO 0.97 0 97 2 Ke n > O. 94, OPERATIONAL Number of Operating Chareine Pumps
- MODE O'
1 2
3 3
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 1.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> 0.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> 4
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 3.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 1.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> 0.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> 5 RCS filled 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 0.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> 5 RCS partially 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> Operation not allowed "
l drained l
l 6
Operation not allowed **
l
\\
" Charging pump OPERABILITY for any period of time shall constitute OPERABILITY for the entire monitoring frequency.
Tite (e7uired chcqiny fumfS S),e11 he VeriRed 40 be inofercble by facKiny
+w Clif their psolcr Clrraif brea ters.
3/4 I~ lYb tJA7ERFORb LIMIT 3
1:
l' TA8LE 3.1-4 REQUIRED MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION AS A FUNCTION 0F OPERATING CHARGING PUMP 5 AND PLANT OPERATIONAL MODES F0R K,ff GREATER THAN 0.95 AND LESS THAN OR EQUAL TO 0.I6 c.94? Grf>a95 I
OPERATIONAL Number of Operatine Chargine Punos*
MODE O
1 2
3 3
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 1.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 4
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 1.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 5 RCS filled 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 1.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 5 RCS partially 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Operation not allowed **
drained 6
Operation not allowed **
- Charging pump OPERABILITY for any period of time shall constitute OPERABILITY for the entire monitoring frequency.
'[he quired clurging famps Sinl/ be verified -lo be inoperah/e by tacky Duf f/seir snohr Circuif bfEck'erS Y I~ l h)t) TE*2f0RD - uN17-2
e TABLE 3.1-5 REQUIRED MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION A5 A FUNCTION OF OPERATING CHARGING PUMP 5 AND PLANT OPERATIONAL MODES FOR
~
K,ff LESS THAN OR EQUAL TO 0.95 Meu & o.qs-Number of Operatine Charoing Pumps #
OPERATIONAL M0DE o
1 2
3 3
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 1.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 4
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 5.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 5 RCS filled 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 5 RC5 partially 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Operation not allowed **
drained 6
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Operation not allowed "
- Charging pump OPERABILITY for any period of time shall constitute OPERA 8ILITY for the entire monitoring frequency.
Ne required C cging famps S/all be verified k be inop-reble by McKing h
ou0 +},cir n,o+c r circaH-bre9 Vers.
Qt}Tl'ARO - WJIT 3 jl4 ;. j7g
REACTIVITY CONTROL SYSTEMS
+..
~
BASES 80 RATION SYSTEMS (Continued)
The contained water volume limits include allowance for water not available j
because of discharge line location, instrument tolerances, and other physical characteristics.
The OPERASILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.
The lower limit on the contained water volume, the specified boron concen-tration, and the physical size (approximately 600,000 gallons) of the RWSP also ensure a pH value of between 7.0 and 11.0 for the solution recirculated within containment after a LOCA.
This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosi5n on mechanical systems and components.
i The maximum limit on the RWSP temperature ensures that the assumptions used in the containment pressure analysis under design base accident condi-tions remain valid and avoids the possibility of containment overpressure.
The minimum limit on the RWSP temperature is required to prevent freezing and/
or baron precipitation in the RWSP.
3/4.1.2.9 BORON DILUTION This specification is provided to. prevent a boron dilution event, and to prevent a loss of SHUTDOWN MARGIN should an inadvertent boron dilution event occur.
Due to boron concentration requirements for the RWSP and boric acid makeup tanks, the only possible boron dilution that would remain undetected by the operator occurs from the primary makeup water through the CVCS system.
Isolating this potential dilution path or the OPERA 8ILITY of the startup channel high neutron flux alarms, which alert the operator with sufficient time available to take corrective action, ensures that no loss of SHUTDOWN MARGIN and unanticipated criticality occur.
'h: r:;;'r;;;nt t: r;;r;; ;:r:r en tr: :h:rging ;- ;; '- "005 5 _ith th: rc::t:r :: !:rt !: ;: dr:* :d i:
n::::: ry b::: :: th:r: i; in;;ffici:nt ti;; f;r :per:t:r r;;;;n :
th; 2-
- nt tu: Or thr : chargin; ;- 7; 'nj :t ;nt:r;t:d ::t:r '7t: th: "CS thi:
- nditi:r The ACTION requirements specified in the event startup channel high neutron flux alarms are inoperable provide an alternate means to detect boron dilution by monitoring the RCS boron concentration to detect any changes.
The frequencies specified in Table 3.1-Ikprovide the operator sufficient time to h
recognize a decrease in boron concentration and take appropriate corrective L4h action without loss of SHUTDOWN MARGIN.
More frequent checks are required g
Sy with more charging pumps in operation due to the higher potential boron dilu-tion rate.
The surveillance requirements specified provide assurance that the startup channel high neutron flux alarms remain OPERA 8LE and that required valve and electrical lineups remain in effect.
3/4.1.3 MOVABLE CONTROL ASSEMBLIES i
The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is WATERFORD - UNIT 3 8 3/4 1-3
y
+-
NPF-38-46
DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-46 This is a request to revise Technical Specification 3.2.1, " Power Distribution Limits, Linear Heat Rate", and the associated Bases for this Specification.
Existing Specification See Attachment A.
Proposed Specification See Attachment B.
Description The proposed change would revise Technical Specification 3.2.1, " Power Distribu-tion Limits, Linear Heat Rate", and the associated Bases for this Specification.
The reasons for this change are: (1) the removal of certain transient-related uncertainties in the CPCs to be consistent with the Cycle 2 safety analysis; (2) elimination of the heat flux augmentation factors; and (3) correction of minor typographical errors.
When COLSS is in service, the Linear Heat Rate (LHR) is maintained at an accept-able level by ensuring that the COLSS calculated core power is less than the i
COLSS calculated Power Operating Limit (POL) based on linear heat rate. Main-taining the core power below the LHR-based POL ensures that, in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200 F.
When COLSS is out of service, the LHR is maintained at an acceptable level by ensuring that the LHR, as indicated on any operable CPC channel, is maintained below the maximum allowable value shown on Figure 3.2.la.
The proposed revision will modify Figure 3.2.la to be consistent with the CPC analyses performed for Cycle 2.
Specifically, conservatisms on the monitoring of linear heat rate with the CPCs, which were credited when Figure 3.2.la was generated, have been reduced for Cycle 2; thus, the limit line will be shifted.
In addition, the penalty which is automatically applied by the CPCs when both CEACs are inoperable is no longer large enough to maintain adequate LHR margin for Cycle 2.
Therefore, Specification 3.2.1c, which allows the CPCs to automatically monitor the LHR, will be deleted. Specification 3.2.lb and the associated Figure 3.2.la have been modified such that they now apply whether or not any CEACs are operable.
Safety Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:
1.
Will the operation of the facility in accordance with this proposed change significantly increase the probability or consequences of any accident previously evaluated?
Response: No.
The Cycle 2 safety analyses have shown that when COLSS is in service, maintaining the core power below the COLSS calculated LHR-based POL, as required by Specification 3.2.la, ensures there is sufficient LHR margin to maintain the clad surface temperature below 2200 F during a LOCA.
Similarly, for the case when COLSS is out of service, operation within the region of acceptable operation of the proposed revision to Figure 3.2.la, as required by Specification 3.2.lb, also ensures that the clad surface temperature will remain below 2200*F during a LOCA. Since the clad surface temperature during a LOCA is the limiting event with respect to LHR margin, the proposed change will not significantly increase the probability or consequences of any accident previously evaluated.
2.
Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes are primarily a result of changes in the Cycle 2 core parameters as well as changes to the CPC software. There has been no physical change to plant structures, systems or components. All changes are either internal to the CPCs or are reflected as proposed revisions to the Technical Specifications. Thus, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
Will operation of the facility in accordance with the proposed change involve a significant reduction of the margin of safety?
Response: No.
The intent of this Specification is to limit the linear heat rate in the core to ensure that, in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200 F.
Generally, the LHR is continuously monitored by COLSS; however, if COLSS is out of service, the limitation on CPC calculated LHR (as a function of cold leg temperature) shown in Figure 3.2.1 represents a conservative envelope of operating conditions consistent with the Cycle 2 safety analysis assumptions. This band of operating conditions has been analytically demonstrated adequate to maintain the clad surface temperature below 2200 F during a LOCA.
Therefore, the proposed change will not result in a significant reduction in the margin of safety.
Safety and Significant Hazards Determination Based on the above Safety Analysis, it is concluded that (1) the proposed change does not constitute a significant hazards consideration as defined by 10 CFR 50.92; (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.
NPF-38-46 ATTACHMENT A
3/4.2 POWER DISTRIBUTION LIMITS 3/4 2.1 LINEAR HEAT RATE LIMITING CONDITION FOR OPERATION 3.2.1 The linear heat rate limit (of Figure 3.2.1) shall be maintained by one of the following methods as applicable:
Maintaining COLSS calculated core power less than or equal to COLSS a.
calculated core power operating limit based on linear heat rate (when COLSS is in service); or b.
Operating within the region of acceptable operation of Figure 3.2-la using any operable CPC channel (when COLSS is out of service and
,either one or both CEACs is operable).
Automatically by CPC (when COLSS is out of service and neither CEAC c.
i is operable).
APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER.
ACTION:
With the linear heat rate limit not being maintained as indicated by:
1.
COLSS calculated core power exceeding COLSS calculated core power operating limit based in linear heat rate; or 2.
When COLSS is out of service, operation outside the region of accep-table operation in Figure 3.2-la; within 15 minutes initiate corrective action to reduce the linear heat rate to within the limits and either:
l a.
Restore the linear heat rate to within its limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or b.
Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.1.1 The provisions of Specification 4.0.4 are not applicable.
4.2.1.2 The linear heat rate shall be determined to be within its limits'when THERMAL POWER is above 20% of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory System (COLSS) or, with the COLSS out of service, by verifying at least once per.
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> that t$e linear heat rate, as indicated on any OPERA 8LE Local Power I
Density channd s, is within the limits shown on Figure 3.2-1.
I 4.2.1.3 At least once per 31 days, the COLSS Margin Alarm shall be verified to actuate at a THERMAL POWER level less than or equal to the core power operating limit based on kW/ft.
s l
WATERFORD - UNIT 3 3/4 2-1 AMEN 0 MENT NO. 5 l
.,_,,,____.m_
..,__~._.,._,__-,,,_4._-,.-.f-____,,_,,._m..
~.,.,,c.
-m
~
l
> k 14.7 a 557.5'F w
Kw/FT s.
14 7 I
I a3 UNACCEPTABLE 3g,'
OPERATION
!hzi w4
>Z
<u 14.6 l
Eo 2 n.
CU Pu!
4m e
E<-y 14.5 m.
7 a
a.o 14 4 Kw/FT
$>Z:
a 520*F ACCEPTABLE I 4
\\
OPERATION m2 34,4
..h
=
a
- m:4 m
ch 14.3 3
510 520 530 540 550 560 Tc INITIAL CORE COOLANT INLET TEMPERATURE, 'F FIGURE 3.2-la ALLOWA8LE PEAK LINEAR HEAT RATE VS Tc FOR COLSS OUT OF SERVICE WATERFORD - UNIT 3 3/4 2-2 AMENOMENT NO.5
.=---,-,-----,,n-
3/4.2 POWER DISTRIBUTION LIMITS BASES 9
3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200*F.
Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the Local Power Density channels in the Core Protection Calculators (CPCs), provides adequate monitoring of the core power distribution and is capable of verifying that the linear heat rate does not exceed its limits.
The COLSS performs this function by continuously monitoring the core power distribution and calculating a core power operating ifmit corresponding to the allowable peak linear heat rate.
Reactor operation at or below this calculated power level assures that the limit of Figure 3.2-1 is not exceeded.
The COLSS calculated core power and the COLSS calculated core power
~
operating limits based on linear heat rate are continuously monitared and l
displayed to the operator.
A COLSS alarm is annunciated in the event that the core power exceeds the core power operating limit.
This provides adequate margin to the linear heat rate operating limit for normal steady-state operation.
Normal reactor power transients or equipment failures which do not require a reactor trip may result in this core power operating limit being exceeded.
In the event this occurs, COLSS alarms will be annunciated.
If the event which causes the COLSS limit to be exceeded results in conditions which approach the core safety limits, a reactor trip will be initiated by the Reactor Protective Instrumentation.
The COLSS calculation of the linear heat rate limit includes appropriate uncertainty and penalty factors necessary to provide a 95/95 confidence level that the maximum ifnear heat rate calculated by COLSS is greater than or equal to that existing in the core.
To ensure that the design margin to safety is maintained, the COLSS computer program includes an Fxy measurement uncertainty factor of 1.053, an engineering uncertainty factor of 1.03, a THERMAL POWER measurement uncertainty factor of.1.02 and appropriate uncertainty and penalty factors for flux peaking augmentation and rod bow.
Parameters required to maintain the operating limit power level based on linear heat rate, margin to DNB and total core power are also monitored by the CPCs (assuming minimum core power of 20% of RATED THERMAL POWER).
The 20%
i RATED THERMAL POWER threshold is due to the neutron flux detector system being less accurate below 205 core power.
Cure noise level at low power is too large to 1 obtain usable detector readings.
Therefore, in the event that the COLSS is not being used, operation within the limits of Figure 3.2-la can be maintained by l
utilizing a predetermined local power density margin and a total core power limit in the CPC trip channels.
The above listed uncertainty and penalty factors are also included in the CPCs.
These penalty fact' ors are determined from uncertainties associated with planar radial peaking measurements, engineering heat flux uncertainty, axial densification, software algorithm modelling, computer processing, rod bow, and core power measurement.
o WATERFORD - UNIT 3 8 3/4 2-1 AMEN 0 MENT N0. 5
BASES i.
~
The additional uncertainty tenns included in the CPC's for transient protection are credited in Figure 3.2-la since this curve is intended to monitor the LC0 only during ste~ady state operation.
In addition, when COLSS is out of service and both CEAC's are inoperable, the 57% penalty applied automatically in CPC can be credited in the CPC linear heat rate calculation since it is required only for transient protection.
In this case, Figure 3.2-1 is automatically maintained by the CPC trip limit.
WATERFORD - UNIT 3 8 3/4 2-la AMENOMENT NO.
5
NPF-38-46 ATTACHMENT B
3/4.2 POWER DISTRIBUTION LINITS 3/4 2.1 LINEAR HEAT RATE -
A LIMITING CONDITION FOR OPERATION 3.2.1 The linear heat rate limit (:' "?;;;r: 2.2.21 shall be maintained by one f
of the following methods as applicable:
a.
Maintaining COLSS calculated core power less than or equal to COLSS calculated core power operating limit based on linear heat rate (when COLSS is in service); or b.
Operating within the region of acceptable operation of Figure 3.2-1/
using any operable CPC channel (when COLSS is out of service)end.
l
, ith:r ::: ;r 5:th C".".C: i: :;;rd h).
" t: :th:ll C0'.!! f: ::t ;f :;r;!:: ::d ::!;h;r 0:^,0 i; :;;rch)y 5; ""O Ol::
c.
APPLICA8ILITY:
MODE 1 above 20% of RATED THERMAL POWER.
ACTION:
With the linear heat rate limit not being maintained as indicated by:
1 1.
COLSS calculated core power exceeding COLSS calculated core power operating limit b linear heat rate; or i
2.
When COLSS is out of service, operation outside the region of accep-table operation in Figure 3.2-14; I
l within 15 minutes initiate corrective action to reduce the linear heat rate to l
within the limits and either:
a.
Restore the linear heat rate to within its limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or b.
Be in at least H0T STAN08Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.1.1 The provisions of Specification 4.0.4 are not applicable.
4.2.1.2 The linear heat rate shall be determined to be within its limits'when THERMAL POWER is above 20% of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory System (COLSS) or, with the COLSS out of service, by verifying at least once per.
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> that the linear heat rate, as indicated on any 0PERABLE Local Power l
Densitychannel/,iswithinthelimitsshownonFigure3.2-1.
I 4.2.1.3 At least once per 31 days, the COLSS Margin Alarm shall be verified to actuate at a THERMAL POWER level less than or equal to the core power operating limit based on kW/ft.
x l
WATERFORD - UNIT 3 3/4 2-1 AMEN 0 MENT N0. 5 l
l
/$.E t
44+a 557.54 Kw/FT -
1.37
,g.
l l
03 UNACCEPTABLE 3y OPERATION Shal u
w4Z
- 13. 7
>0
-M.
EUa z.
0 0 m
ya
$e$
13.4 h
n.:
z=t
/
o.o
-1+.+ Kw/ FT
$>2 a $20T ACCEPTABLE 4
{
.I
- 3.5 OPERATION E $
-M.'
r
._p 2
a w
h
- 13. 4 n.
y 510 520 530 540 550 560 Tc l
l INITIAL CORE COOLANT INLET TEMPERATURE, V t
I FIGURE 3.2-1/
ALLOWA8LE PEAK LINEAR HEAT RATE VS Tc FOR COLSS OUT OF SERVICE l
WATERFORD - UNIT 3 3/4 2-2.
AMEN 0 MENT NO. 5 y
t.
3/4.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200*F.
Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the Local Power Density channels in the Core Protection Calculators (CPCs), provides adequate monitoring of the core power distribution and is capable of verifying that the linear heat rate does not exceed its limits.
The COLSS performs this function by continuously monitoring the core power distribution and calculating a core power operating limit corresponding to the allowable peak linear heat rate.
Reactor operation at or below this calculated power level assures that the limit of Fig _ure 0.21 I
is not exceeded.
@jg
'The COLSS calculated core power and the COLSS calculated core powei-operating limits based on linear heat rate are continuously monitored and displayed to the operator. A COLSS alarm is annunciated in the event that the core' power exceeds the core power operating ifmit.
This provides adequate margin to the linear heat rate operating limit for normal steady-state operation.
Normal reactor power transients or equipment failures which do not require a reactor trip may result in this core power operating limit being exceeded.
In the event this occurs, COLSS alarms will be annunciated.
If the event which j
causes the COLSS limit to be exceeded results in conditions which approach the core safety limits, a reactor trip will be initiated by the Reactor Protective Instrumentation.
The COLSS calculation of the linear heat rate limit includes appropriate uncertainty and penalty factors necessary to provide a 95/95 confidence level that the maximum linear heat rate calculated by COLSS is greater than or equal to that existing in the core.
To ensure that the design margin to safety is maintained, the COLSS computer program includes an Fxy measurement uncertainty factor of 1.053, an engineering uncertainty factor of l
1.03, a THERMAL POWER measurement uncertainty factor of 1.02 and appropriate r:r-t '-ty : d penalty factors for *1= ;::9 ;; r; ::t ti:: e% rod bow.
Parameters required to maintain the operating limit power level based on linear heat rate, margin to DNB and total core power are also monitored by the CPCs (assuming minimum core power of 20% of RATED THERMAL POWER).
The 20%
RATED THERMAL POWER threshold is due to the neutron flux detector system being less accurate below 20% core power.
Core noise level at low power is too large to 1 obtain usable detector readings. Therefore, in the event that the COLSS is not being used, operation within the limits of Figure 3.2-1g can be maintained by ll utilizing a predetermined local power density margin and a total core power limit in the CPC trip channels. The above listed uncertainty and penalty factors are also included in the CPCs.
These penalty fact' ors are determined from uncertainties associated with
~
planar radial peaking measurements, engineering heat flux uncertainty, axial densification, sof tware algorithm modelling, computer processing, rod bow, i
and core power measurement.
WATERFORD - UNIT 3 8 3/4 2-1 AMEN 0 MENT N0. 5
BASES The additional uncertainty terms included in the CPC's for transient protection are credited in Figure 3.2-1g since this curve is intended to I
monitor the LCO only during steady state operation.
.r.
22,.,__
raie.c 4....__... _..:.
_2
.m _ _
1 m _. m.r..e a r.. ___,_____.mi.
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2
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- r..o r.
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- _. m. _
r..or.
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- ... 4....,.....
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3.,_,_ 4... __.__..,_.i.,,..
_.t....
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m.
f 1
l WATERFORD - UNIT 3 B 3/4 2-la AMEN 0 MENT NO.
5 e
H 9
ne f
a E
o'e E
13.4 5 557.5V k
UNACCEPTABLE Kw/F1V U
kNS_ _Y_ ?_D g
5 13.3
/
wwo 11 2 p.
g T
- 13. I Km/ FT a:
5 520*F
/
4
\\
Al:cEPTanLE N
w 13.1 g
( 20 R f
^^^^{h oR 5
l E
13.0 l
SiO 520 530 540
!50 560 Tc INITIAL CORE COOLANT INLET TEMPERATURE,Y.
Fmute e L ?-t
+HiWIE-ari>+
ALLOWA8LE PEAK LINEAR HEAT RATE VS Tc 6 3/4 2.ib WATERFORD - UNIT 3
- /' :-:
4 NPF-38-47 r
f
DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-47 This is a request to revise Table 3.3-2 of Technical Specification 3.3.1,
" Reactor Protective Instrumentation Response Times".
Existing Specification See Attachment A.
Proposed Specification See Attachment B.
Description The proposed change would revise Table 3.3-2 of Technical Specification 3.3.1,
" Reactor Protective Instrumentation Response Times". The reason for this change is to ensure that the response times associated with the CPC calculation of a low DNBR, high Local Power Density (LPD) or low reactor coolant pump shaft speed reactor trip are consistent with the values used in the Cycle 2 safety analysis.
The current values of the subject response times represent the conservatively long delay times that had to be assumed for the Cycle 1 safety analysis since plant specific measurements were not available. They were adequate to satisfy the acceptance criteria for low DNBR and high LPD during the limiting transients that were analyzed for Cycle 1.
The proposed CPC related response times for Cycle 2 would be shorter (i.e.,
faster) than those used for Cycle 1.
These values have been justified by the Cycle 2 safety analysis and are well within the actual response times measured prior to and during Cycle 1.
Although these changes are not required for all the instruments shown in the proposed change to Table 3.3-2, they are being submitted at this. time to eliminate unnecessary conservatisms in anticipation of potential-ly less favorable core parameters in future cycles.
The proposed increase in the hot and cold leg RTD response times is being done to account for the potential degradation of the RTD response times that has been observed at other plants. This change will increase the likelihood that field measurements made on RTD response times will satisfy the Technical Specifications for Cycle 2 and future cycles.
In addition, the footnote labeled "**" which follows Table 3.3-2 has been revised. The actual response time testing of these instruments in the field includes the time delay associated with the opening of the reactor trip breakers; hence, the footnote has been revised to reflect this.
The Cycle 2 safety analysis was performed with the proposed response times dis-cussed above, and the less favorable core parameters that result from an 18 month fuel cycle, and resulted in all applicable acceptance criteria being satisfied.
NS20578
Safety Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:
1.
Will the operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?
Response: No.
The anticipated operational occurrences (A00s) and accidents that require the CPCs to respond with shorter (faster) response times than those listed in the current Technical Specification 3.3.1 (Table 3.3-2) are:
(1) the loss of load from one steam generator (i.e., spurious closure of a main steam isolation valve), which is dependent upon the maximum response time of the cold leg temperature instrumentation and internal processing of that signal by the CPCs; (2) the CEA withdrawal and steam line break, which are dependent upon the maximum response times of the neutron flux power input from the ex-core detectors; and (3) the loss of flow, which is dependent upon the maximum response time of the reactor coolant pump shaft speed signals in the CPCs. These events were analyzed using response times consistent with the proposed change to Table 3.3-2 and the less favorable Cycle 2 core parameters during the Cycle 2 safety analysis. The results of these analyses are presented in Section 7 of the Reload Analysis Report and show that the events are either bounded by the Cycle 1 safety analysis or they are within the acceptance criteria specified by the appropriate sections of the NRC Standard Review Plan. Thus, the proposed change will not significantly increase the probability or consequences of any accident previously evaluated.
2.
Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change is primarily a result of changes in the Cycle 2 core parameters and the subsequent reevaluation of the limiting A00s. There has been no physical change to plant systems, structures or components.
lhe only change to plant procedures will be a tightening of the acceptance criteria that must be satisfied when performing surveillance testing to verify the response times associated with internal CPC processing. Thus, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
Will operation of the facility in accordance with these proposed changes involve a significant reduction in a margin of safety?
Response: No.
The intent of Table 3.3-2 is to ensure that the appropriate Engineered Safety Feature Actuation and/or reactor trip, which is associated with each protection system channel, is completed within the time limit assumed on the safety analysis.
In this case, the response times assumed for the Cycle 2 safety analysis were somewhat faster than those used in the Cycle 1 analysis. Thus, in order to ensure the Cycle 2 analyses are valid, the actual response times must be less than or eoual to the proposed change to Table 3.3-2.
Since the Cycle 2 safety analyses have shown acceptable re-sults using the response times listed in the proposed change to Table 3.3-2, there will not be a significant reduction in the margin of safety.
The Commission has provided guidance concerning the application of standards for determining whether a significant hazards consideration exists by providing cer-tain examples (48 FR 14870) of amendments that are considered not likely to involve significant hazards considerations.
Example (ii) relates to a change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications, (e.g., a more stringent surveillance requirement).
With the exception of increasing the RTD response time, the proposed change is similar to Example (ii) in that the change to Table 3.3-2 provides additional restrictions on the allowable CPC response times that are not presently included en the Technical Specifications.
Safety and Significant Hazards Determination Based on the above Safety Analysis, it is concluded that:
(1) the proposed change does not constitute a significant hazards consideration as defined in 10 CFR 50.92; (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station the environment as described in the NRC Final Environmental Statement.
1 NPF-38-47 ATTACHMENT A
TABLE 3.3-2 d
REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES 8
- =
7 FUNCTIONAL UNIT RESPONSE TIME E
1.
Manual Reactor Trip Not Applicable w
2.
Linear Power Level - High 5 0.40 second*
3.
Logarithmic Power Level - High 5 0.40 second*
4.
Pressurizer Pressure - High 5 0.90 second 5.
Pressurizer Pressure - Low
< 0.90 second 6.
Containment Pressure - High 5 1.70 seconds
{
7.
Steam Generator Pressure - Low
$ 0.90 second
[
8.
Steam Generator Level - Low
$ 0.90 second i
j 9.
Local Power Density - High a.
Neutron Flux Power from Excore Neutron Detectors 5 0.634 second*
b.
CEA Positions
< 0.645 second**
)
c.
CEA Positions:
CEAC Penalty Factor 50.429second 10.
DNBR - Low a.
Neutron Flux Power from Excore Neutron Detectors
< 0.634 second*
i b.
CEA Positions 2 0.645 second**
c.
Cold Leg Temperature 30.634second#
d.
Hot Leg Temperature 5 0.634 second#
1 e.
Primary Coolant Pump Shaft Speed
< 0.487 second**
l f.
Reactor Coolant Pressure from Pressurizer 7 0.634 second##
g.
CEA Positions:
CEAC Penalty Factor 30.429second
C.'.
sc TABLE 3.3-2 (Continued)
D gg -
REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES 2
B FUNCTIONAL UNIT RESPONSE TIME f
11.
Steam Generator Level - High Not Applicable 12.
Reactor Protection System Logic Not Applicable 13.
Reactor Trip Breakers Not Applicable 14.
Core Protection Calculators Not Applicable 15.
CEA Calculators Not Applicable 16.
Reactor Coolant Flow - Low 0.70 second R.
.n
- Neutron detectors are exempt from response time testing.
Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
- Response time shall be measured from the time the CPC/CEAC receives an input signal until the system outputs a trip signal.
- Response time shall be measured from the output of the sensor.
RTD response time for all the RTDs shall be measured at least once per 18 months.
The measured P of the slowest RTD shall be less than 7
or equal to 6 seconds (P assumed in the safety analysis).
t
- Response time shall be measured from the output of the pressure transmitter.
The transmitter response time shall be less than or equal to 0.70 second.
NPF-38-47 ATTACHMENT B
TABLE 3.3-2 Dg REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES 8
=
]
7 FUNCTIONAL UNIT RESPONSE TIME E
1.
Manual Reactor Trip Not Applicable w
2.
Linear Power Level - High
_ 0.40 second*
3.
Logarithmic Power Level - High
< 0.40 second*
4.
Pressurizer Pressure - High
< 0.90 second 5.
Pressurizer Pressure - Low
< 0.90 second 6.
Containment Pressure - High
< 1.70 seconds
{
7.
Steam Generator Pressure - Low
$ 0.90 second
[
8.
Steam Generator Level - Low
$ 0.90 second 9.
Local Power Density - High I
a.
Neutron Flux Power from Excore Neutron Detectors second*
O.4,29
~
4 b.
CEA Positions i
second==
a4,24 c.
CEA Positions:
CEAC Penalty Factor 3
second C 379 i
10.
DNBR - Low 1
a.
Neutron Flux Power from Excore Neutron Detectors
.63 second*
i o.4A9 b.
CEA Positions 2.
.6 second**
'O.4J4 c.
Cold Leg Temperature 2 0.6 4lsecond#
- o.,2n d.
Hot Leg Temperature 7 0. 34 second#
- ##9 i
e.
Primary Coolant Pump Shaft Speed 21047 second**
- ## 7 f.
Reactor Coolant Pressure from Pressurizer 7
. 63 ' second##
O ##7 g.
CEA Positions:
CEAC Penalty Factor 3
_. 42 second 6JW
C. '.
i j
gg TABLE 3.3-2 (Continued)
-4 j
gy -
REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES i
a EI FUNCTIONAL UNIT RESPONSE TIME C
2j 11.
Steam Generator Level - High Not Applicable 12.
Reactor Protection System Logic Not Applicable
'a 13.
Reactor Trip Breakers Not Applicable 14.
Core Protection Calculators Not Applicable t
i 15.
CEA Calculators Not Applicable 4
16.
Reactor Coolant Flow - Low 0.70 second I
us 32
- Neutron detectors are exempt from response time testing.
Response time of the neutron flux signal portion i
of the channel shall be measured from detector output or input of first electronic component in channel.
l
- Response time shall be measured from the time the CPC/CEAC receives an input signal until the ;y:trr
- tput; ; trip 1 7.11. elecifical foo>er is in ferrupted to the CEA drive niec)ninism.
9
- Response time shall be measured from the output of the sensor.
RTD response time for all the RTDs l
shall be measured at least once per 18 months.
The measured P f the slowest RTD shall be less than orequaltoj[ seconds (P assumed in the safety analysis).
l r
- Response time shall be measured from the output of the pressure transmitter.
The transmitter response time shall be less than or equal to 0.70 second.
I 4
l
NPF-38-48 t
DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-48 This is a request to revise Technical Specification B3/4.1.3 and 3.1.3.6,
" Movable Control Assemblies" and " Regulating CEA Insertion Limits", respec-tively.
Existing Specification See Attachment A.
Proposed Specification See Attachment B.
Description Technical Specification 3.1.3.6 imposes limits on the allowable position of the regulating CEA groups and on the allowable duration within a given position range. The CEA insertion limits are given by Figure 3.1-2.
The proposed change will increase the maximum allowable insertion of Group 6 by 7.5 inches during long term steady state operation.
It would also reduce the allowable insertion during transient operation for Groups 5 and 6 above 20% power. Between zero and 20% power, insertion of Group 4 would be limited to 60 inches.
Technical Specification 3.1.3.6 provides protection against a CEA ejection event. During such an event, limits are generally placed on how far CEAs may be inserted into the core in order to limit the event consequences - i.e., the greater the insertion, the larger the positive reactivity addition is during the rod ejection. Because of the slightly higher fuel enrichments for Cycle 2, a slight increase in ejected rod reactivity would be expected in comparison to Cycle 1.
The proposed changes to Figure 3.1-2 will modify the insertion limits with respect to Cycle 1 to ensure that the consequences of the Cycle 2 CEA ejection are bounded by the Cycle 1 analysis presented in the FSAR.
Axial Shape Index (ASI) control is necessary during low power operation. ASI control can be provided by maneuvering Groups 5 and 6 essentially tip-to-tip within the limits established by the proposed change to Figure 3.1-2.
An additional statement is proposed to be included in the Bases, B3/4.1.3 to clarify that this operation is permissible, and that proper protection against sequence error is provided for by the Core Protection Calculators.
Safety Analysis The proposed changes described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:
NS41182
a 1.
. Will the operation of the facility in accordance with these proposed changes involve a significant increase in the probability or conse-quences of any accident previously evaluated?
Response: No.
The proposed change to the Bases is a statement of clarification for operation under analyzed conditions. The proposed change to Figure 3.1-2 allows less flexibility during operation at and during approach to steady state full power operation. The Cycle 1 Safety Analysis remains the bounding analysis for the CEA ejection event. Therefore, there will be no significant increase in the probability or consequences of any accident previously evaluated.
2.
Will operation of the facility in accordance with these proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The statement in the Bases only provides clarification for operation under analyzed conditions. The proposed change to the operating restric-tions on CEA insertion limits does not create any new fault or accident path and remains bounded by the Cycle 1 analysis.. As such, it cannot cause the creation of a new or different kind of accident than any pre-viously evaluated.
3.
Will operation of the facility in accordance with these proposed changes involve a significant reduction in a margin of safety?
Response: No.
As clarification, the statement in the Bases has no impact on any safety margins. The proposed change of Figure 3.1-2 ensures that the Cycle 2 i
CEA ejection event is less limiting than the Cycle 1 event. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The Commission has provided guidance concerning the application of standards for determining whether a significant hazards consideration exists by providing cer-tain examples (48 CFR 14870) of amendments that are considered not likely to involve significant hazards considerations.
Example (iii) relates to a change resulting from a nuclear reactor core reloading, if no fuel assemblies signifi-cantly different from those found previously acceptable to the NRC for a pre-vious core at the facility in question are involved. This assumes that no sig-nificant changes are made to the acceptance criteria for the Technical Specifi-cations, that the analytical methods used to demonstrate conformance with the Technical Specifications and regulations are not significantly changed, and that the NRC has previously found such methods acceptable.
The proposed change to the graph is similar to Example (iii) since this change directly results from the Waterford 3 Cycle 2 reload, and is bounded by the l
Cycle 1 CEA Ejection Analysis.
l i
--n-._-.,,
Safety and Significant Hazards Determination Based on the above Safety Analysis, it is concluded that:
(1) the proposed change does not constitute a significant hazards consideration as defined in 10 CFR 50.92; (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station the environment as described in the NRC Final Environmental Statement.
d 1
i I
c i
i i
9 NPF-38-48 ATTACHMENT A
~
REACTIVITY CONTROL SYSTEMS REGULATING CEA INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The regulating CEA groups shall be limited to the withdrawal sequence and to the insertio~n limits
- shown on Figure 3.1-2** with CEA insertion between the Long Term Steady State Insertion Limits and the Transient Insertion Limits restricted to:
Less than or equal to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval, a.
b.
Less than or equal to 5 Effective Full Power Days per 30 Effective Full Power Day interval, and Less than or equal to 14 Effective Full Power Days per calendar c.
year.
APPLICA8ILITY:
MODES 1*** and 2*** #.
ACTION:
With the regulating CEA groups inserted beyond the Transient Insertion a.
Limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, or Reactor Power Cutback, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:
1.
Restore the regulating CEA groups to within the limits, or 2.
Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the CEA group position using the above figure.
b.
With the regulating CEA groups inserted between the Long Term Steady State Insertion Limits and the Transient Insertion Limits for intervals greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per 24-hour interval, operation may proceed provided either:
1.
The Short Term Steady State Insertion Limits of Figure 3.1-2 are not exceeded, or 2.
Any subsequent increase in THERMAL POWER is restricted to less than or equal to 5% of RATED THERMAL POWER per hour.
- Following a reactor power cutback in which (1) Regulating Groups 5 and/or 6 are dropped or (2) Regulating Groups 5 and/or 6 are dropped and the remaining Regulating Groups (Groups 1, 2, 3, and 4) are sequentially inserted, the Transient Insertion Limit of Figure 3.1-2 can be exceeded for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
- CEAs are fully withdrawn in accordance with Figure 3.1-2 when withdrawn to at least 145 inches.
- See Special Test Exceptions 3.10.2 and 3.10.4.
- With X,77 greater than or equal to 1.0.
WATERFORD - UNIT 3 3/4 1-25 i
..m
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
ACTION: (Continued)
With the regulating CEA groups inserted between the Long Term Steady c.
State Insertion Limits and the Transient Insertion Limits for intervals greater than 5 EFPD per.30 EFPD interval or greater than 14 EFPD per calendar year, either:
1.
Restore the regulating groups to within the Long Tern Steady State Insertion Limits within two hours, or 2.
Be in at least HOT STAN08Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each regulating CEA group shall be determined to be within the Transient Insertion Limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the POIL Auctioneer Alarm Circuit is inoperable, then verify the individual CEA positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
The accumulated times during which the regulating CEA groups are inserted beyond the Long Term Steady State Insertion Limits but within the Transient Insertion Limits shall i
be determined at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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REACTIVITY CONTROL SYSTEMS BASES MOVA8LE CONTROL ASSEMBLIES (Continued) a continued operations when the positions of CEAs with inoperable position indicators can be verified by the " Full In" or " Full Out" limits.
CEA positions and OPERA 8ILITY of the CEA position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent j
verifications required if an automatic monitoring channel is inoperable.
l These verification frequencies are adequate for assuring that the applicable i
LCO's are satisfied.
l The maximum CEA drop time restriction is consistent with the assumed CEA j
drop time used in the safety analyses.
Measurement with T,yg greater than or i
equal to 520*F and with all reactor coolant pumps operating ensures that the l
measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.
The establishment of LSSS and LCOs requires that the expected long and short-term behavior of the radial peaking factors be determined.
The long term behavior relates to the variation of the steady-state radial peaking factors with core burnup and is affected by the amount of CEA insertion assumed, the portion of a burnup cycle over which such insertion is assumed, and the expected power level variation throughout the cycle.
The short term behavior relates to transient perturbations to the steady-state radial peaks due to radial xenon redistribution.
The magnitudes of such perturbations depend upon the expected use of the CEAs during anticipated power reductions and load maneuvering.
Analyses are performed based on the expected mode of operation of the NSSS (base loaded, or load maneuvering) and from these 4
1 analyses CEA insertions are determined and a consistent set of radial peaking factors defined.
The Long Tern Steady State and Short Tern Insertion Limits are determined based upon the assumed mode of operation used in the analyses i
and provide a means of preserving the assumptions on CEA insertions used.
The limits specified serve to limit the behavior of the radial peaking factors within the bounds determined from analysis.
The actions specified serve to limit the extent of radial xenon redistribution effects to those accommodated l
in the analyses.
The Long and Short Term Insertion Limits of Specifica-l tion 3.1.3.6 are specified for the plant which has been designed for primarily base loaded operation but which has the ability to accommodate a limited amount j
of load maneuvering.
The Transient Insertion Limits of Specification 3.1.3.6 and the Shutdown j
CEA Insertion Limits of Specification 3.1.3.5 ensure that (1) the minimum i
SHUTDOWN MARGIN is maintained, and (2) the potential effects of a CEA ejection accident are limited to acceptable levels.
Long-term operation at the Transient Insertion Limits is not permitted since such operation could have effects on j
the core power distribution which could invalidate assumptions used to determine
]
the behavior of the radial peaking factors.
The Part Length CEA Insertion Limits of Specification 3.1.3.7 ensure that adverse power shapes and rapid local power changes which affect radial peaking factors and DN8 considerations do not occur as a result of a part-length CEA group covering the same axial segment of the fuel assemblies for an extended period of time during operation.
f WATERFORD - UNIT 3 8 3/4 1-5 l
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REACTIVITY CONTROL SYSTEMS BASES MOVA8LE CONTROL ASSEDSLIES (Continued) continued operations when the positions of CEAs with inoperable position indicators can be verified by the " Full In" or " Full Out" limits.
CEA positions and OPERABILITY of the CEA position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is inoperable.
These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.
The maximum CEA drop time restriction is consistent with the assumed CEA drop time used in the safety analyses.
Measurement with T,,g greater than or equal to 520*F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.
The establishment of LSSS and LCOs requires that the expected long and short-term behavior of the radial peaking factors be determined.
The long term behavior relates to the variation of the steady-state radial peaking factors with core burnup* and is affected by the amount of CEA insertion assumed, the portion of a burnup cycle over which such insertion is assumed, and the expected power level variation throughout the cycle.
The short ters
- i behavior relates to transient perturbations to the steady-state radial peaks l
due to radial xenon redistribution.
The magnitudes of such perturbations depend upon the expected use of the CEAs during anticipated power reductions and load maneuvering.
Analyses are performed based on the expected mode of operation of the NSSS (base loaded, or load maneuvering) and from these analyses CEA insertions are determined and a consistent set of radial peaking
{
factors defined.
The Long Ters Steady State and Short Tern Insertion Limits 1
are determined based upon the assumed mode of operation used in the analyses l
and provide a means of preserving the assumptions on CEA insertions used.
The i
limits specified serve to limit the behavior of the radial peaking factors within the bounds determined from analysis.
The actions specified serve to limit the extent of radial xenon redistribution effects *to those accommodated in the analyses.
The Long and Short Tern Insertion Limits of Specifica-tion 3.1.3.6 are specified for the plant which has been designed for primarily base loaded operation but which has the ability to accommodate a limited amount of load maneuvering.
The Transient Insertion Limits of Specification 3.1.3.6 and the Shutdown CEA Insertion Limits of Specification 3.1.3.5 ensure that (1) the minimum SHUTDOWN MARGIN is maintained, and (2) the potential effects of a CEA ejection accident are limited to acceptable levels.
Long-term operation at the Transient Insertion Limits is not permitted since such operation could have effects on the core power distribution which could invalidate assumptions used to determine the behavior of the radial feakingfactors.7_ _,
)
The Part length CEA Insertion Limits of Specification 3.1.3.7 ensure that adverse power shapes and rapid local power changes which affect radial peaking factors and DN8 considerations do not occur as a result of a part-length CEA, i
group covering the same axial segment of the fuel assemblies for an extended period of time during operation.
34 of Reg. Cvoses 5 and G is permdfed i
to hs. essutially %-to-tip JtSin & limits by W 9 m
WATERFORD - UNIT 3 8 3/4 1-5 %s mekod of uns@ation W 6 W W.
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NPF-38-49
DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-49 This is a request to revise ACTION statement "b" to Technical Specification 3.3.3.10, " Radioactive Effluent Monitoring Instrumentation", and ACTION statements 28, 29 and 30 to Table 3.3-12 of that same Specification.
Existing Specification See Attachment A.
Proposed Specification See Attachment B.
Description The proposed change would revise ACTION statement "b" to Technical Specification 3.3.3.10, " Radioactive Effluent Monitoring Instrumentation", and ACTION statements 28, 29 and 30 to Table 3.3-12 of that same Specification.
The reason for this change is to clarify the wording of the ACTION statements such that there is no confusion as to what actions must be taken when the minimum channels OPERABLE requirements of Table 3.3-12 are not satisfied.
ACTION statement "b" will be clarified by providing a 30 day time period in which to restore any required monitering instrumentation (listed in Table 3.3-12) to OPERABLE status or, if unsuccessful, to explain in the next Semi-annual Effluent Release Report why the instrument was not restored in the specified time.
Additional clarification will be added by including a statement that allows releases to continue as long as the specified ACTIONS of Table 3.3-12 are continued.
This change is consistent with the proposed Revision 3 to NUREG-0472 (Standard Radiological Effluent Technical Specifications for PWRs) in which the intent is to eliminate the need for a Licensee Event Report (LER) simply because the required instrumentation could not be restored to OPERABLE status within the specified time.
The concentration of radionuclides released will remain within the limits specified in Technical Specification 3.11.1 and 10 CFR 20 Appendix B by continuing to perform the ACTIONS required by Table 3.3-12 prior to and during any liquid effluent releases.
This will ensure that the levels of radioactive materials in bodies of water in unrestricted areas will not result in exposures to any member of the general public in excess of 3 millirems to the total body or 10 millirems to any organ which is in accordance with 10 CFR 50 Appendix I Section II.A.
ACTION statements 28, 29 and 30 to Table 3.3-12 will be clarified by removing any reference to a specified time period and replacing it with a provision that best efforts be made to repair the required instrumentation.
In addition, the last statement of ACTION statement 28 will be deleted since it is already quite clear that releases through a specific pathway may not occur if the two condi-tions of this ACTION statement can not be satisfied.
NS41158
This change is being' proposed in order to make the ACTION statements of Table 3.3-12 consistent with the proposed change to ACTION statement "b".
That is, since the proposed' change to ACTION statement "b" will define the time limits and reporting requirements that must be adhered to whenever the minimum number of channels OPERABLE is less than that required by Table 3.3-12, there is no need to repeat them.
i All changes meet the intent of General Design Criteria 60, 63 and 64 which require a means to control and monitor all radiological storage areas and releases to the environment during normal operation, including anticipated operational occurrences and' postulated accidents.
Safety Analysis i
The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:
1.
Will the operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?
Response
No.
Theproposedchangestothir,TechnicalSpecificationbreadministrative and do not affect the mann<.r in which the plant ~is operated.
The changes are meant to define the time limits and reporting requirements that must be adhered to whenever the minimum channels OPERABLE requirements of Table 3.3-12 are not satisfied and therefore, do not have an effect on any of the accident analyses.
Thus, the proposed change will not involve a significant increase in the probability or consequences of any accident previously evaluated.
2.
Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of a,ccident from any accident previously evaluated?
Response
No.
The proposed changes are meant to clarify the ACTION statements of Tech-nical Specification 3.3.3.10 and do not make any changes to the facility or to the operating procedures.
Since the liquid radiological effluents have no new pathways to the environment, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
Will operation of the facility in accordance with this proposed change involve a significant reduction in the margin of safety?
Response
No.
The intent of this Technical Specification is to comply with General Design Criteria 60, 63 and 64 by requiring a means to monitor and control radiological storage areas and releases to the environment.
Normally this is accomplished using on-line instrumentation; however, in the event that the required instruments are not in service, compliance with these criteria is ensured by following the appropriate ACTION statements.
Since the changes made to the ACTION statements were administrative, the proposed change does not involve a significant reduction in the margin of safety.
Safety and Significant Hazards Determination Based on the above Safety Analysis, it is concluded that:
(1) the proposed change does not constitute a significant hazards consideration as defined by 10 CFR 50.92(c); (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.
b NPF-38-49 ATTACHMENT A
INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 Tne radioact1ve liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded.
The alare/
trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (00CM).
APPLICABILITY:
At all times.
ACTION:
With radioactive liquid effluent monitoring instrumentation channel a.
alarm / trip setpoint less conservative than required by the above Specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable.
b.
With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or, explain in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.8, why this inoperability was not corrected within the time specified.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
c.
SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERA 8LE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-8.
l t
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WATERFORD - UNIT 3 3/4 3-55 I
I TABLE 3.3-12 RADYo&CTIVR LICWID EFFLUBrF NDIFTTotIWS TMBTEtBERTATION I
IWafatafENT CIW arELs s
OPBtABLE ACTION 1.
. RADIOACTIVITY MONITORS FROFIDIWG At. ARM AND AUTONATIC TERNIMATION OF RELEASE 5
s 5
Boric Acid Coodenaste Discharge a.
1 to b.
Vaste, Umsta Condamente and Laondry Discharge 5
w 1
23 c.
Dry Cooling Tower Suure 1/ sump 29 i
4 Turbine Building Industrial Waste Sway 29 1
Circulating Water. Dischargs (31oudoun and Bleudeum usat y
e.
Exchanger and Auxiliary Component Cooling Water Pumps)#
r I
- .;i u
2.
CONTINUOUS COMPOSITE SAMPf.RR8 29 g
1 m
~
C y
a.
Steam Generato'r 31oudous Effluent Line
_n g
1 29 3.
FLOW RATE HEASUREMENT DEVICES i
n i
o 1
m Doric Acid Coodeneste Discharge a.
i 1
30 b.
Waste.. Waste Condensate and Laundry Discharge t
go a
Turbisie Ruilding Industrial Weste Sump c.
N.A.
3,A, j
d.
Dry Coo' ling Tower Suspe N.A.
- N.A.
'Ji e.
Circulating Weter Discharge (Alowdown and Slowdown Heat Exchanger and Availlery Component Cooling Water Pumps) l 2)
N.A.
N. A..
N f
- Alf70HATI.
r etMINAT]ON DIP AI.OUDOWN DISCHARc6 (WFY
7 l
TABLE 3.3-12 (Continued)
TA8LE NOTATIONS
- Pump performance curves generated in place shall be used to estimate flow.
ACTION STATEMENTS ACTION 28 -
With the number of channels OPERA 8LE less than required by the Minimum Channels OPERA 8LE requirement, effluent releases via this pathway may continue for up to 14 days provided that prior to initiating a release:
At least two independent samples are analyzed in accordance a.
with Specification 4.11.1.1.1, and b.
At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving; Otherwis'e, suspend release of radioactive effluents via this pathway.
ACTION 29 -
With the number of channels OPERABLE less than required by the Minimum Channels OPERA 8LE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are analyzed for radioactivity at a lower limit of detection of at least 10-7 microcurie /mL.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activity of a.
the secondary coolant is greater than 0.01 microcurie / gram DOSE EQUIVALENT I-131.
b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of the secondary coolant is less than or equal to 0.01 micro-curie / gram DOSE EQUIVALENT I-131.
ACTION 30 -
With the number of channels OPERA 8LE less than required by the Minimum Channels OPERA 8LE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases.
Pump performance curves generated in place may be used to estimate flow.
WATERFORD - UNIT 3 3/4 3-57
TABLE 4.3 8 RADIOACTIVS LIQUID BFFLUENT NONITORIWS INSTRUMBrFATION SURVIILLANCE REQOIRBSENTS 1
CRAINEL CEAlttIL SOURCE CRANNEL FUNCTICIIAL TNSTRUMBff CafECK MPrr CAI,IBRATION TEST g
8 4
4 1.
RAD 10ACTIVIIT HONITORS PROVIDING l
g ALARMS AND AUTCHATIC TERMINATI0tf m
0F RELEASE Q
.e ta W
j a.
Boric Acid Condensate Discherse P
P R(3)
Q(1)
_e a
b.
Vaste, Vaste Coedamente and Laundry P
P R(3)
Q(1) 1 Discharge c.
Dry Cooling Tower Suspe D
N R(3)
Q(5) s d.
Turbine Building Industrial Weste Swap D
M R(3)
Q(5)
N r-7 g
i s.
Circulating Water Discharge (Blowdown and Blowdows Neat Exchanger 2,
j u
)
and Auxi,11ery Ceaposent Cooling Water Pumpe)#
0 M
R(3) q(3) p l
y I
g 2.
' CONTINUOUS COMPOSITE SAMPLERS n
i o
l a.
Steen' Generator Blowdova Effluent Lima D(6)
N.A.
R q
3.
FLOW RATE HEASUREMENT' DEVICES Soric., Acid Condebeste Discharge D(4)
N.A.
R q
a.
Weste, Weete Con'ensate and Laundry D(4)
N.A.
O q
l 5
b.
d Discharge g
L l
c.
Turbina Building Indesatrial Weste Samp N.A.
N.A.
N.A.
N.A.
e l
d.
Dry Cooling Tower Swaps M.A.
N.A.
N.A.
N.A.
U l
e.
Circulating Water Dischstgo j
(Blovdown end Slowdown Nest Exchenr. ore l
j, and Auxiltnry Componcut Cooling Wator
MAY 16 ' 95 18:15 QATEL SERVICES CORP.
PAGE.04
?
TABLE 4.3-8 (continued) r TA8tE NOTATION (1) The CHANNEL FUNCTIONAL TEST ghtli also demonstrate of this pathway ar.d Centrvi riios alarm annunciation occur if any following conditions exists:
1.
Instrument indleates measured levels above the alarm / trip set 2.
Circuit failure.
3.
Instrument indicates a downscale failure.
(2) The CHANNEL. FUNCTIONAL TEST sheti also demonstra alars annunciation occurs if any of the following conditions exists:
1.
Instrtment indicates measured levels above the alam setpoint.
2.
Circuit failure.
, (3)
The initial CHANNEL CALIBRATICII shall b'e perfomed using one or more the reference standards certified by the National Bureau of Standards or esing standards that have been obtained free sigpliers that participate in seasurement assurance activities with 25.These standards shall permit calibrating the system for over its intended range of energy and measurement range.
been related to the initial-calibration shall be used.For subsequent C c
(4). CHAMEL CHECK shall consist of verifying indication of flow during p t
l of release.
days on which continuous, periodic, or batch releases are ma l
The CHAME5. FUNCTIONAL TEST shall also demonst (5) of this pathway occurs if the instrument indicates esasured levels above the alars/ trip setpoint and that control room alars annunciation occlirs if any of the following conditions exists:
1.
Instrument indicates measured levels above the alara set.
2.
Circuit failure.
3.
Instrument controls not. set in operate mode.
(6) CWuGEL CHECK shall be ande at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous releases are ande to the Circulating Water System or l
Waterford 3 waste poed.
6 l
l WaTERFORD - UNIT 3 3/4 3-59 1
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NPF-38-49 ATTACHMENT B
INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT-MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive liquia effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded.
The alare/
trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (00CM).
APPLICA8ILITY:
At all times.
ACTION:
With radioactive liquid effluent monitoring instrumentation channel a.
alarm / trip setpoint less conservative than required by the above Specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable.
b.
With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERA 8LE, take the ACTION shown in Table 3.3-12. Restore the inoperable instrumentation to OPERABLE 30 days or, status within th; ti : :;;;f'f:P th: ".CTI^" Or, explain in the l
if unsuccessful, next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.8, why this inoperability was not corrected within the time specified. #eleases need rioe he te -ong fed a //ee 3o days prov;ded the spa,/,ed ACTIONS are confin4*d-c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
l SURVEILLANCE REQUIREMENTS I
4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the l
frequencies shown in Table 4.3-8.
WATERFORD - UNIT 3 3/4 3-55
TABLE 3.3-12 (Continued)
TABLE NOTATIONS
" Pump performance curves generated in place shall be used to estimate flow.
rosdd Q e[ Arts ACTION STATEMENTS are mde k refer y/,, insframent and s u _-
._ d ACTION 28 -
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue ':r x; t: M d y: pr:rfd:d that prior to initiating a release:
At least two independent samples are analyzed in accordance a.
with Specification 4.11.1.1.1, and b.
At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving; Oth: rf:, :::;: d re?:::: c' r:d!:::t! : : "? ::t: vf: thi
- t
- :;.
ACTION 29 -
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue ':r up t: 20 dry: pr:rfd:d grab samples l
are analyzed for radioactivity at a lower limit of detection of
.g g gp at least 10-7 microcurie /mL.
e *de d# "f'ir d' instrumet M
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activity of a.
~
the secondary coolant is greater than 0.01 microcurie / gram DOSE EQUIVALENT I-131, or b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of the secondary coolant is less than or equal to 0.01 micro-l curie / gram DOSE EQUIVALENT I-131.
1 ACTION 30 -
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue ':r ; t: 'O dry: pr:r!d:d the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases.
Pump performance curves generated in place may be used to estimate flow.
prowded besi efforl5 are made to ref* 4 inhwQ"f WATERFORD - UNIT 3 3/4 3-57
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i l
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E DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-50 This is a request to revise ACTION statements "a" and "b" to Technical Specifi-cation 3.3.3.11, " Radioactive Gaseous Effluent Monitoring Instrumentation", and ACTION statements 35, 36, 37, 38, 39 and 40 to Table 3.3-13 of that same Specification.
Existing Specification See Attachment A.
Proposed Specification See Attachment B.
Description The proposed change would revise ACTION statements "a" and "b" to Technical Specification 3.3.3.11, " Radioactive Gaseous Effluent Monitoring Instrumentation",
and ACTION statements 35, 36, 37, 38, 39 and 40 to Table 3.3-13 of that same
' Specification.
The reason for changing ACTION statement "a" is to add an option that would return the plant to compliance with the Limiting Condition for Operation (LCO).
The reason for changing ACTION statement "b" and all the ACTION statements to Table 3.3-13 is to clarify the wording such that there is no confusion as to what actions must be taken when the minimum channels OPERABLE requirements of Table 3.3-13 can not be satisfied.
ACTION statement "a" will be modified by adding the option to change the alarm /
trip setpoint of a monitoring channel to an acceptably conservative value when it has been determined the LC0 has not been satisfied.
Currently this ACTION statement requires that releases via the affected channel be suspended or that the channel be declared INOPERABLE.
Adding the option to change the setpoint to an acceptable value would allow that channel to be returned to OPERABLE status and hence, to comply with the LCO.
ACTION statement "b" will be clarified by providing a 30 day time period in which to restore any required monitoring instrumentation (listed in Table 3.3-13) to OPERABLE status or, if unsuccessful, to explain in the next Semi-annual Effluent Release Report why the instrument was not restored in the specified time.
Additional clarification will be added by including a statement that allows releases to continue as long a the specified ACTIONS of Table 3.3-13 are continued.
This change is consistent with the proposed Revision 3 to NUREG-0472 (Standard Radiological Effluent Technical Specifications for PWRs) in which the intent is to eliminate the need for a Licensee Event Report (LER) simply because the required instrumentation could not be restored to OPERABLE status within the specified time.
By continuing to comply with the ACTION statements of Table 3.3-13, the radioactive gaseous effluents released will not result in the exposure of a member of the public to an average annual concentration exceeding the limits specified in Appendix B of 10 CFR 20.
In addition, this change is NS41161
4 consistent with the requirements of Appendix I to 10 CFR 50 by keeping the annual dose from gaseous effluents to individuals in unrestricted areas less than 10 millirads for gamma radiation and 20 millirads for beta radiation.
ACTION statements 35, 36, 37, 38, 39 and 40 to Table 3.3-13 will be clarified by removing any reference to a specific time period and replacing it with a provi-sion that best efforts be made to repair the required instrumentation.
In addition, the last sentence of ACTION statement 35 will be deleted since it is already quite clear that releases through a specific pathway may not occur if the two conditions of this ACTION statement can not be satisfied.
This change is being proposed in order to make the ACTION statements of Table 3.3-13 consistent with the proposed change to ACTION statement "b".
That is, 4
since the proposed change to ACTION statement "b" will define the time limits and reporting requirements that must be adhered to whenever the minimum number of channels OPERABLE is less than that required by Table 3.3-13, there is no need to repeat them.
All changes meet the intent of General Design Criteria 60, 63 and 64 which require a means to control and monitor all radiological storage areas and releases to the environment during normal operation, including anticipated operational occurrences, and postulated accidents.
Safety Analysis The proposed change described above shall be deemed to involve a significant l
hazards consideration if there is a positive finding in any of the following areas:
i 1.
Will the operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?
Response
No.
t The proposed changes to this Technical Specification are administrative and do not affect the manner in which the plant is operated.
The changes are meant only to clarify what actions are to be taken when the LC0 can not be satisfied.
Since there is no change to the intent of this Specifi-cation or on any of the Safety Analyses, the proposed change will not involve a significant increase in the probability or consequences of any accident previously evaluated.
2.
Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response
No.
The proposed changes are meant to clarify the ACTION statements of i
Technical Specification 3.3.3.11 and the associated Table 3.3-13.
There are no changes being made to the facility or to any of the operating pro-cedures.
Since the gaseous radwaste effluents have no new pathways to the environment, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
Will operation of the facility in accordance with this proposed change involve a significant reduction in the margin of safety?
Response
No.
The intent of this Technical Specification is to comply with General Design Criteria 60, 63 and 64 by requiring a means to monitor and control radioactive storage areas and gaseous releases to the environment.
Normally this is accomplished by using on-line instrumentation; however, in the event that the required instruments are not in service, compliance with these criteria.is ensured by following the appropriate ACTION state-ments.
Since the changes made to the ACTION statements were only admin-istrative, the proposed change does not involve a significant reduction in the margin of safety.
Safety and Significant Hazards Determination Based on the above Safety Analysis, it is concluded that:
(1) the proposed change does not constitute a significant hazards consideration as defined by 10 CFR 50.92; (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.
O 4
NPF-38-50 ATTACHMENT A
INSTRUMENTATION RADIOACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.11 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded.
The alarm /
trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and param(ters in the 00CM.
APPLICABILITY:
As shown in Table 3.3-13.
ACTION:
a.
With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above Specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel,, inoperable.
b.
With less than the minimum number of radioactive gaseous ef fluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13.
Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or, explain in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.8, why this inoperability was not corrected within the time specified.
c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.11 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERA 8LE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-9.
WATERFORO - UNIT 3 3/4 3-60
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TABLE 3.3-13 s-h RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 4
o MINIMUM CHANNELS E
INSTRUMENT OPERA 8LE APPLICA8ILITY ACTION g
1.
WASTE GAS HOLDUP SYSTEM
[
a.
Noble Gas Activity Monitor -
Providing Alarm and Automatic Termination of Release 1
3S j
b.
Effluent System Flow Rate Measuring Device 1
36 2.
WASTE GAS HOLDUP SYSTEM EXPLOSIVE GAS l
MONITORING SYSTEM a.
Hydrogen Monitor 1
38 l
[
b.
Oxygen Monitors 2
40 1
3.
MAIN CONDENSER EVACUATION AND TURBINE GLAND SEALING SYSTEM i
a.
Noble Gas Activity Monitor 1
37 b.
Iodine Sampler #
1 39 i
- l c.
Particulate Sampler #
1 39 i
d.
Sampler Flow Rate Monitor 1
36 i.
l
- If a primary to secondary leak exists or if the noble gas monitors in the main condenser evacuation and j
turbine gland sealing system or if the steam generator blowdown monitor indicates the presence of radioactivity in the secondary system, the flow from this release point shall be diverted immediately to the plant stack.
i These instruments are in the plant stack and sampling for radiciodines and particulates shall occur at the i
plant vent when this occurs.
I i
j i
b i
TABLE 3.3-13 (Continued)
^
I i
g RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION I
E j
g MININUM CHANNELS i
O INSTRUMENT OPERA 8LE APPLICA8ILITY ACTION e
i E
4.
REACTOR AUXILIARY SUILDING Q
VENTILATION SYSTEM (PLANT STACK) w Noble Gas Activity' Monitor -
i a.
Providing Alare and Automatic Termination of Released 1
37 4
b.
Iodine Sampler 1
39 i
l c.
Particulate Sampler 1
39 w
i
)
d.
Flow Rate Monitor 1
36 w
d a
e.
Sampler Flow Rate Monitor 1
36 i
n 5.
FUEL HANDLING BUILDING j
VENTILATION SYSTEM (NDRMAL)
I l
a.
Noble Gas Activity Monitor 1
37 i
b.
Iodine Sampler 1
39
)
c.
Particulate Sampler 1
39 b
d.
Flow Rate Monitor 1
36 e.
Sampler Flow Rate Monitor 1
36 i
l
- Automatic termination of containment purge only.
TABLE 3.3-13 (Continued)
TA8LE NOTATIONS
- At all times.
- During WASTE GAS HOLDUP SYSTEM operation.
- With irradiated fuel in the storage pool.
ACTION STATEMENTS ACTION 35 -
With the number of channels OPERA 8LE less than required by the Minimum Channels OPERA 8LE requirement, the contents of the tank (s) may be released to the environment for up to 14 days provided that prior to initiating the release:
a.
At least two independent samples of the tank's contents are analyzed, and b.
At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve Ifneup; otherwise, suspend release of radioactive effluents via this pathway.
ACTION 36 -
With the number of channels OPERA 8LE less than required by the Minimum Channels OPERA 8LE requirement, affluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
For the waste gas holdup tank this action item is applicable only during periods of release.
For the main condenser evacuation and turbine gland sealing systems, this action item applies only during 4
l release via the discharge silencer and only during turbine gland sealing operations and/or vacuum pump operation.
ACTION 37 -
With the number of channels OPERA 8LE less than required by the Minimum Channels OPERA 8LE requirement, affluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- However, containment purging of radioactive affluents must be immediately suspended during this condition for the plant stack only.
ACTION 38 -
With the number of channels OPERA 8LE less than required by the Minimum Channels OPERA 8LE requirement, operation of the WASTE GAS HOLOUP SYSTEM may continue for up to 14 days provided grab samples are collected at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and analyzed within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the onservice gas decay tank.
WATERFORD - UNIT 3 3/4 3-63
F l
TABLE 3.3-13 (Continued)
ACTION STATEMENTS ACTION 39 -
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided ss=ples are continuously collected with auxiliary sampling equipment as required in Table 4.11-2.
ACTION 40 -
With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, operation of the WASTE GAS HOLDUP SYSTEM may continue provided that the system is sampled by either the remaining monitor or by a grab sample once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and the oxygen concentration remains less than 2%.
Such operation may continue for up to 14 days.
If there are no monitors OPERABLE, WASTE GAS HOLDUP SYSTEM operation may continue provided a grab sample is taken and analyzed from the onservice gas decay tank once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and the oxygen concentration remains less than 1%.
With oxygen concentration exceeding 1%, reduce the oxygen concentration to less than 1%
within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
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WATERFORD - UNIT 3 3/4 3-64
1 TABLE 4.3-9 e
i
_RADI0 ACTIVE GASEGUS EFFLUENT MDNITORING INSTRUMENTATION SURVEILLANCE REQUIRENENTS
.i g
o CHANNEL MODES IN WHICH B
CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION TEST IS REQUIRED 5
1.
WASTE GAS HOLDUP SYSTEM
'd a.
Noble Gas Activity Monitor -
Providing Alare and Automatic l
Termination of Release P
P R(3)
Q(1) 1 b.
Effluent System Flow Rate 4
Measuring Device P
N.A.
R Q
l 2.
WASTE GAS HDLDUP SYSTEM EXPLOSIVE GAS MDNITORING SYSTEM Y
a.
Hydrogen Monitor D
M.A.
Q(4)
M aa Y
b.
Oxygen Monitors D
N.A.
Q(5)
M 3.
MAIN CONDENSER EVACUATION AND TUR81NE GLAND SEALING SYSTEM 1
j a.
Noble Gas Activity Monitor D
M R(3)
Q(2) a b.
Iodine Sampler W
N.A.
N.A.
N.A.
c.
Particulate Sampler W
N.A.
N.A.
N.A.
a 1
d.
Sampler Flow Rate Monitor D
N.A.
R Q
a l
I l
1
TA8LE 4.3-9 (Continued)
$g RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRlstENTATION SURVEILLANCE REQUIREMENTS 4
E CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRtBENT CHECK CHECK CALIBRATION TEST IS REQUIRED i
e E
4.
REACTOR AUXILIARY SUILDING VENTILATION SYSTEM (PLANT STACK) a.
Noble Gas Activity Monitor -
Providing Alarm and Automatic Termination of Release #
D M
R(3)
Q(6) j b.
Iodine Sampler W
N.A.
N.A.
N.A.
c.
Particulate Sampler W
N.A.
N.A.
N.A.
j l
N d.
Flow Rate Monitor D
N.A.
R Q
i E
e.
Sampler Flow Rate Monitor D
N.A.
R Q
5.
FUEL HANDLING BUILDING j
VENTILATION SYSTEM (NORMAL) a.
Noble Gas Activity Monitor D
M R(3)
Q(2)
{
b.
Iodine Sampler W
N.A.
N.A.
N.A.
i l
c.
Particulate Sampler W
N.A.
N.A.
N.A.
i d.
Flow Rate Monitor D
N.A.
R Q
i i
e.
Sampler Flow Rate Monitor D
N.A.
R Q
l
- Automatic termination of containment purge only.
1
, TABLE 4.3-9 (Centinued)
TABLE NOTATIONS
- At all times.
- 0uring WASTE GAS HOLDUP SYSTEN operation.
- When irradiated fuel is in the spent fuel pool.
(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolat of this pathway and control room alare annunciation occurs if any of the following conditions exists:
1.
Instrument indicates measured levels above the alare/ trip setpoint.
2.
Circuit failure.
3.
Instrument indicates a downscale failure.
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room
~
alarm annunciation occurs if any of the following conditions exists:
1.
Instrument indicates measured levels above the alarm setpoint.
2.
Circuit failure.
(3) The initial CHANNEL CALIBRATION shall be perforised using one or more of the reference standards certified by the National Bureau of Standards (N85) or using standards that have been obtained from suppliers that participate in measurement assurance activities with N85.
These standards shall permit calibrating the system over its intended range of energy and measurement range.
For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.
(4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
1.
Zero volume percent hydrogen, balance nitrogen, and 2.
Four volume percent hydrogen, balance nitrogen.
(5) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
1.
Zero volume
Four volume percent oxygen, balance nitrogen.
(6) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway occurs if the instrument indicates measured levels above the alarm / trip setpoint and that control room alarm annunciation occurs if any of the following conditions exists:
1.
Instrument indicates measured levels above the alarm set.
2.
Circuit failure.
3.
Instrument controls not set in operate mode.
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l WATERFORD - UNIT 3 3/4 3-67
f NPF-38-50 ATTACHMENT B
r INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.11 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded.
The alarm /
trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.
APPLICABILITY:
As shown in Table 3.3-13.
ACTION:
a.
With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above Specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel, inoperable, or cheye +he selpint so it is accytebly conserw+Ne.
b.
With less than the minimum number of radioactive gaseous effluent
~
monitoring instrumentation channels OPERABLE, take the ACTION shown
_ _ _ T
~
30 J<ys or, if s in Table 3.3-13.
Restore the inoperable instrumentation to OPERA 8LE Jstatuswithinth: :f:: :;;;i " f:d'"- th: "CT:CM ;r, explain in the
-""8"##883#"h next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.8, why this inoperability was not corrected within the time spec 1fied. Selesses need no+ pa }.msins+ed efter 30 hys provided 4he specifed Actions Are em nue c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.11 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-9.
WATERFORO - UNIT 3 3/4 3-60
e TA8LE 3.3-13 (Continued)
TA8LE NOTATIONS
- At all times.
- During WASTE GAS HOLDUP SYSTEM operation.
- With irradiated fuel in the storage pec1.
l ACTION STATEMENTS l
ACTION 35 -
With the number of channels OPERA 8LE less than required by the L
Minimum Channels OPERA 8LE requirement, the contents of the tank (s) may be released to the environment ':7 pr: td:d that prior to initiating the release: /; " M d:y:
l y e g p e fforfs g
[re made la refair d' insfru,,ient ed A,f j At least two independent samples of the tank's contents,
a.
are analyzed, and i
i b.
At least two technically qualified members of the facility staff fredependently verify the release rate calculations and discharge valve lineup;
^t.pf::,:::;:dr:!:::: :" r:dir::t!:: : "?:::t: ;i ei; ACTION 36 -
With the number of channels OPERA 8LE less than required by the I
Minimum Channels OPERA 8LE requirement effluent releases via jg T.nis pathway may continue ':r ; t 20*d:y: pr: td: the flow h g, g gejs g
je g,,
rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
For the waste gas Mput,
gg holdup tank this action item is applicable only during periods of release.
For the main condenser evacuation and turbine gland sealing systems, this action item applies only during release via the discharge silencer and only during turbine gland sealing operations and/or vacuum pump operation.
ACTION 37_-
With the number of channels OPERA 8LE less than required by the (yg g,gorf33 Minimum Channels OPERA 8LE requirement, effluent releases via
,,j, f, 9 ;r A s this pathway may continue ";r ; O 20"d y: ;;;;fd:d grab Ms/rument and 1/wt samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- However, containment purging of radioactive affluents must be immediately suspended during this condition for the plant stack only.
ACTION 38 -
With the number of channels OPERA 8LE less than required by the Minimum Channels OPERA 8LE requirement. operation of the WASTE t
Jed eQeh GAS HOLDUP SYSTEM may continue f:r ; t: Md:y: ;r: td:d grab IdSYS*'4E "['" N f' "d* f* "
samples are collected at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and analyzed l
l within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the onservice gas decay tank.
l 1
WATERFORD - UNIT 3 3/4 3-63 l
i
TABLE 3.3-13 (Continued) grovided est e/ Torts are made to repair the
~
ACTION STATEMENTS ins /mment and W,,/
ACTION 39 -
WiththenumberofchannelsOPERABLElessthaIrequiredbythe Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue f:r up t: 30'd:y: pr:.td:d samples are continuously collected with auxiliary sasoling l
equipment as required in Table 4.11-2.
ACTION 40 -
With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, operation of the WASTE GAS HOLDUP SYSTEM may continue pr:.id:d SP. t the system is sampled by either the remaining monitor or by a grab sample et least jonce per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and the oxygen concentration remains less than 2%.
Ex:5 ;;;r:ti:n ;;; ::ntinu; f;r ;; t: li d;y:.
If there are no monitors OPERABLE, WASTE GAS HOLDUP SYSTEM operation may continue provideEja grab sample is taken and analyzed from the onservice gas aecay tank once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and the oxygen concentration remains less than 1%.
With oxygen concentration
- j g exceeding 1%, reduce the oxygen concentration to less than 1%
within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
hed efforts are pude lo (L"lurn 46 leqSt one ch<nnel b OPERA 6LE sinius and fbGI t
I e
t i
WATERFORD - UNIT 3 3/4 3-64
_.