ML20081J623
| ML20081J623 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 03/24/1995 |
| From: | Johannes R WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20081J626 | List: |
| References | |
| CO-95-0001, CO-95-1, NUDOCS 9503280071 | |
| Download: ML20081J623 (19) | |
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W$LF CREEK NUCLEAR OPERATING CORPORATION Rchard N. Johannes Chief AdmnustrativeOffcer March 24, 1995 CO 95-0001 U.' S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-137 Washington, D.
C.
20555
Subject:
Docket No. 50-482: Revision to Technical Specification 3.5.1,
" Emergency Core Cooling Systems - Accumulators" Gentlemen:
This letter transmits an application for amendment to Facility Operating License No. NPF-42 for Wolf Creek Generating Station (WCGS).
This license amendment request proposes revising Technical Specification.3.5.1 and its associated Bases.
Specifically, it is being proposed that a new action statement be added to Technical Specification 3.5.1 which would provide a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed outage time for one accumulator being inoperable due to its boron concentration not meeting the 2300-2500 ppm band.
This proposed change is consistent with NUREG-1431,
" Standard Technical Specifications - Westinghouse Power Plants."
Also, the allowed outage time for the current action statement would be increased to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in lieu of the current allowed outage time of one hour.
In addition to the proposed changes described above, it is being proposed that Technical Specification Surveillance Requirements 4.5.1.1.a.1 and
- 4. 5.1.1.b be revised and Technical Specification Surveillance Requirement 4.5.1.2 be deleted.
These proposed changes to the surveillance requirements are compatible with plant operating experience and are consistent with the guidance of Generic letter 93-05.
Also, Technical Specification 3.5.1 Bases Section would be revised to discuss the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowed outage times for the action statements associated with Technical Specification 3.5.1.
Attachment I provides a description of the proposed change along with a Safety Evaluation.
Attachment II provides a No Significant Hazards Consideration Determination.
Attachment III provides the Environmental Impact Determination.
The' specific change to the technical specifications proposed by this request is provided as Attachment IV.
In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Kansas State of ficial.
This proposed revision to the WCGS technical specifications will be fully implemented within 30 days of formal Nuclear Regulatory Commission approval.
PO Box 411/ Burhngton, KS 66839 / Phone: (316) 364-8831 4
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9503280071 950324 l
l PDR ADOCK 05000482 An Equal Opportunity Employer WF/HCNET J
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I CO'95-0001 Page 2 of 2 If you have any questions concerning this matter, please contact me at (316) 364-8831, extension 4001, or Mr. Richard D.
Flannigan, at extension 4500.
Very truly yours, h
ll k "
o Richard N. Joh es RNJ/jra Attachments:
I - Safety Evaluation II - No Significant Hazards Consideration Determination III - Environmental Impact Determination IV - Proposed Technical Specification Change cc:
G.
W. Allen (KDHE), w/a L. J.
Callan (NRC), w/a D.
F.
Kirsch (NRC), w/a J.
F. Ringwald (NRC), w/a J.
C.
Stone (NRC), w/a
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STATE OF KANSAS
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ss COUNTY OF COFFEY-
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l Richard N. Johannes, of lawful age, being first duly sworn upon oath says that-he is Chief-Administrative Officer of Wolf Creek Nuclear: Operating
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'I Corporationi that he has read the foregoing document and 'knows the content thereof; that he has executed that same for and on behalf of said Corporation with full power and authority _to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and. belief.
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By i
Richard N. Johannes //
Chief Administrative Mfficer SUBSCRIBED and sworn to before me this
_ day of
/ M, 1995.
'I CAROLYN E. LONG, W
Notary Public '
f Notary Public - State of Kansas
,My Appt. Empires
/=S-99 j
Expiration Date
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Attachment I to CO 95-0001 Page 1 of 10 ATTACHMENT I SAFETY EVALUATION i
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Attachment I to CO 95-0001 i
Page 2'of 10 Safety Evaluation PROPOSED CHANGE
. This license' amendment request proposes revising Technical Specification 3.5.1,
" Emergency Core Cooling Systems - Accumulators," and'its associated Bases. A new' Action Statement - would be added to Technical. Specification. 3.5.1 which. would -
provide an allowed outage time (AOT) of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if one accumulator is' inoperable
'i due to its boron concentration not meeting the 2300-2500 ppm band.
The current' I
action statement would be followed if an accumulator is inoperable for any other reason.
This approach is consistent with NUREG-1431,
" Standard Technical-Specifications - Westinghouse Power Plants."
The AOT for the current action statement would be increased to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in lieu of the current AOT of one hour.
I Surveillance Requirements 4.5.1.1.a.1 and
- 4. 5.1.1.b would also be revised and Surveillance Requirement 4.5.1.2 would be removed in accordance with the guidance of NRC Generic Letter 93-05, "Line-Item Technical Specifications Improvements to Reduce Surveillance Requirements for. Testing During Power Operation," and NUREG-1366, " Improvements to Technical Specifications Surveillance Requirements."
- However, Surveillance Requirement 4.5.1.2 would be retained in Chapter 16 of the Updated Safety Analysis Report based on the recommendations of Generic Letter 93-05.
Technical Specification Bases Section 3/4.5.1 would also be revised to discuss the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AOTs for the action statements associated with Technical Specification 3. 5.1, the revision of Surveillance Requirements 4.5.1.1.a.1 and 4.5.1.1.b, and the deletion of Surveillance Requirement 4.5.1.2.
SYSTEM DESIGN The accumulators are pressure vessels partially filled with borated water and pressurized with nitrogen gas.
During normal operation, each accumulator is isolated from the Reactor Coolant System (RCS) by a motor-operated isolation valve (open.during power operation with power locked out) and two check valves in series.
Should the RCS pressure fall below the accumulator pressure, the check valves - open and borated water is forced into the RCS.
The accumulators are passive components, since no operator or control actions are required in order for them to perform their function.
One' accumulator is attached to each of the cold legs of the RCS.
Mechanical operation of the swing-disk check valves is the only action required to open the injection path from the accumulators to the reactor core via the cold legs.
Connections are provided to remotely adjust the level and boron concentration of the borated water in each accumulator during normal plant operation, as required.
The accumulator water level can be adjusted either by draining to the recycle holdup tank or by pumping borated water from the refueling water storage tank (RWST) to the accumulator.
Samples of the solution in the accumulators are taken periodically to verify proper boron concentration.
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4 Attachment I to CO 95-0001 Page 3 of 10 Accumulator pressure is provided by a supply of nitrogen gas, and can be adjusted, as required, during normal plant operation.
However, the accumulators are normally isolated from this nitrogen supply. The accumulators are also protected from pressures in excess of design pressure by gas relief valves.
The accumulator gas pressure is monitored by indicators and alarms and solenoid-operated vent valves are provided to depressurize the accumulators during emergency cold shutdown conditions.
EMERGENCY CORE COOLING SYSTEM ANALYSIS A Loss-of-Coolant Accident (LOCA) is defined as a rupture of the RCS piping or of any line connected to the system from which the break flow exceeds the flow capability of the normal makeup / charging system.
Ruptures of small cross-sections will cause expulsion of the reactor coolant at a rate which can be accommodated by the centrifugal charging pumps maintaining an operational water level in the pressurizer, permitting the operator to execute an orderly shutdown.
The maximum break size for which the normal makeup system can maintain the pressurizer level is obtained by comparing the calculated flow from the RCS through the postulated break against the centrifugal charging pump makeup flow at normal RCS pressure, i.e.,
2,250 psia.
A makeup flow rate from one centrifugal charging pump is adequate to sustain pressurizer level at 2,250 psia for a break through a 0.375 inch diameter hole.
This break results in a loss of approximately 17.5 lb/sec (127 gpm at 130 degrees Fahrenheit and 2,250 psia) of reactor coolant.
For the analyses reported in Chapter 15 of the Updated Safety Analysis Report, a major pipe break (large break) is defined as a rupture with a total cross-sectional area equal to or greater than 1.0 square foot (ft2).
This event is considered an ANS Condition IV event, a limiting fault, in that it is not expected to occur during the life of the plant but is postulated as a conservative design basis.
The Acceptance Criteria for the LOCA are described in 10 CFR 50.46 as follows:
A.
The calculated peak fuel element clad temperature shall not exceed the requirement of 2,200 degrees Fahrenheit.
B. The amount of the fuel element cladding that reacts chemically with water or steam, shall not exceed 1 percent of the total amount of Zircaloy in the fuel cladding.
C. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The localized cladding oxidation limit of 17 percent is not exceeded during or after quenching.
D.
The core reaains amenable to cooling during and after the break.
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' Af.tachmentcI to CO 95-0001 Page 4 of 10 E. The core. temperature is reduced - and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining'in the core.
'Should a pipe break occur, depressurization of the RCS results in a pressure r
decrease in the pressurizer.
A reactor trip occurs and the safety injection system is actuated when=their respective pressurizer low pressure trip setpoints.,
j are reached.
Reactor trip'and safety injection system actuation may be provided by a high containment pressure signal, depending on the actual break size. These countermeasures limit the consequences of the accident in two ways:-
A. The reactor trip and borated water injection provide additional negative reactivity insertion to supplement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.
However, no credit is taken during the LOCA blowdown for negative reactivity'due to the boron content of the injection water.
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B.
Injection of borated water ensures sufficient flooding of the core to l
prevent excessive clad temperatures.
During blowdown, heat from fission product decay, hot internals, and the vessel continues to be transferred to the reactor coolant.
At the beginning of the l
blowdown phase, the entire RCS contains subcooled liquid which transfers. heat i
from the core by forced convection -with some fully developed nucleate boiling.
Thereafter, the core heat transfer is based on local conditions with transition boiling, film boiling, and forced convection to steam as the major heat transfer j
mechanisms.
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When the RCS depressurizes to 600 paia, the accumulators begin to inject borated
[
water into the reactor coolant loops.
The blowdown phase of the transient ends when the RCS pressure (initially assumed at 2,300 psia) falls to' a value approaching that of the containment atmosphere.
Prior to, or at the end-of the blowdown, the mechanisms that are responsible for the bypassing'of emergency core cooling water injected into the RCS are no longer in effect.
At : this - time (called end-of-bypass), refill of the reactor vessel lower plenum begins. Refill is complete when emergency core cooling water has filled the lower plenum of the reactor vessel, which is bounded by the bottom of the fuel rods.(called bottom of i
core recovery time).
The reflood phase of the transient is defined as the time period lasting from the i
end-of-refill until the rector vessel has been filled with water to the extent i
that the core temperature rise has been terminated.
From the later stage of blowdown and then the beginning-of-reflood, the safety injection accumulator tanks rapidly discharged borated cooling water into the RCS, contributing to the filling of the reactor vessel downcomer.
The downcomer water elevation head f
provides the driving force required for the reflooding of the reactor core.
The f
centrifugal charging, safety injection, and residual heat removal pumps aid in the filling of the downcomer and subsequently supply water to maintain a full downcomer and complete the reflooding process.
Attachment I to CO 95-0001 Page 5 of 10 The accumulators are assumed operable in both the large and small break LOCA analyses at full power.
In performing the LOCA calculations, conservative assumptions are made concerning the availability of ECCS flow.
In the early stages of a LOCA, with or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS.
The assumption of loss of offsite power conservatively imposes a delay wherein the emergency core cooling system IECCS) pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go through their timed loading sequence.
In cold leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break.
The limiting large break LOCA is a double ended guillotine break at the discharge of the reactor coolant pump.
During this event, the accumulators discharge to the RCS as soon as RCS pressure decreases to below accumulator pressure.
As a conservative estimate, no credit is taken for ECCS pump flow until an effective delay has elapsed.
This delay accounts for the emergency diesel generators starting and the pumps being loaded and delivering full flow.
The delay time is conservatively set with an additional 2 seconds to account for a safety injection signal generation.
During this time, the accumulators are analyzed as providing the sole source of emergency core cooling.
No operator action is assumed during the blowdown stage of a large break LOCA.
The worst case small break LOCA analyses also assume a time delay before pumped flow reaches the core.
For the larger range of small breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then providing continued cooling. As break size decreases, the accumulators and centrifugal charging pumps both play a part in terminating the rise in clad temperature. As break size continues to decrease the accumulators are no longer required and the centrifugal charging pumps become solely responsible for terminating the temperature increase.
For both the large and small break LOCA analyses, a nominal contained accumulator water volume is used.
The contained water volume is the same as the deliverable volume for the accumulators, since the accumulators are emptied, once discharged.
For small breaks, an increase in water volume is a peak clad temperature penalty.
For large breaks, an increase in water volume can be either a peak clad temperature penalty or benefit, depending on downcomer filling and subsequent spill through the break during the core reflooding portion of the transient.
The analysis makes a conservative assumption with respect to ignoring or taking credit for line water volume from the accumulator to the check valves.
Attachment I to CO 95-0001 Page 6 of 10 The minimum boron concentration setpoint is used in the post-LOCA boron concentration calculation.
The calculation is performed to assure reactor subcriticality in a post-LOCA environment.
No credit is taken for control rod assembly insertion for a large break LOCA.
A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sump concentration for post-LOCA shutdown and an increase in the maximum sump pH.
The maximum boron concentration is used in determining the cold leg to hot leg recirculation switchover time and minimum sump pH, EVALUAT10H The proposed change to Technical Specification 4.5.1.1.a.1 to remove the reference to verifying operability "by the absence of alarms" is consistent with the guidance of Generic Letter 93-05 and NUREG-1366.
These guidance documents also recommend the relocation of Technical Specification 4.5.1.2, which requires the performance of a channel calibration of each accumulator water level and pressure channel once per 18 months, to the LW tted Safety Analysis Report.
These recommendations were based on the re :gnization that accumulator instrumentation operability is not directly related to the capability of the accumulators to perform their safety function.
These proposed changes are also compatible with the operating experience at Wolf Creek Generating Station (WCGS).
The proposed change to Technical Specification 4.5.1.1.b to add a note clarifying when the surveillance is not required is also consistent with Generic Letter 93-05 and NUREG-1431.
This generic letter and NUREG concluded that it should not be necessary to verify boron concentration of accumulator inventory after a volume increase of 1 percent or more if the makeup water is from the RWST and the minimum concentration of boron in the RWST is greater than or equal to the minimum boron concentration in the accumulator, the recent RWST sample was within specifications, and the RWST has not been diluted.
These proposed changes are also compatible with the operating experience at WCGS.
The proposed addition of a new action statement to Technical Specification 3.5.1 which would provide a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed outage time if one accumulator is inoperable due to its boron concentration not meeting the 2300 - 2500 ppm band is conaistent with NUREG-1431, " Standard Technical Specifications Westinghouse Plants."
With an accumulator declared inoperable due to its boron concentration being out of specification, the ability to maintain subcriticality or minimum boron precipitation time may be reduced.
The boron in the accumulators contributes to the assumption that the combined ECCS water in the partially recovered core during the early reflooding phase of a large break LOCA is sufficient to keep that portion of the core suberitical.
One accumulator below the minimum boron concentration limit, however, will have no effect on available ECCS water and an insignificant effect on core subcriticality during reflood.
Therefore, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is appropriate to return the boron concentration to within limits.
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Attachment I to CO 95-0001 Page 7 of 10 In support of the proposed license amendment to increase the allowed outage time (AOT) of the accumulator from one hour to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the WCGS Individual Plant Examination (IPE) was reviewed.
The IPE was performed by conducting a Probabilistic Safety Assessment (Level 1), including an analysis of internal flooding, and a Containment Performance Analysia (Level 2) of the plant.
In performing the IPE, standard probabilistic safety assessment (PSA) systems analysis practices such as those outlined in the Probabilistic Risk Assessment Procedures Guide (NUREG/CR-2300) were used.
The total core damage frequency (CDF) reported in the WCGS IPE is 4.2E-5 per reactor year.
This total includes internal flooding events.
The PSA has been reexamined to determine the consequences of an increased allowed outage time of the ECCS accumulators.
I The major topics of discussion in this section are:
equipment and system considerations in the initial development of the o
accumulator fault tree, changes to the accumulator fault tree based upon additional basic events, e
changes in the Event Trees to eliminate several initial conservative e
decisions, requantification of the accumulator fault trees and the resultant impact on e
core damage frequency.
Accumulator Fault Tree Model The ECCS accumulator injection fault tree in the WCGS IPE did not contain a basic event for Test and Maintenance of the accumulators.
The ECCS accumulwar injection fault tree in the IPE only contained basic events for the diocharge check valves and common cause failure of the same valves.
- However, discharge isolation motor-operated valve mispositioning, flow diversion, and loss of nitrogen pressure from the accumulator were considered in the development of the accumulator fault tree.
The spurious closure of the accumulator discharge isolation valves EPHV8808A, B,
C, and D is not considered as a credible fault due to the fact that these valves are normally open with their motor control center circuit breakers padlocked in the power disconnect position.
These valves are not required to change state during an accidenc and are treated as locally locked open valves.
Potential flow diversion of the water in the accumulator tanks via the fill lines from the safety injection pump, the accumulator tank sample lines, or the valve test lines are not modeled.
The failure probability of these diversion paths, given the manual valves and check valves in the flow paths, is not significant relative to other faults.
Loss of nitrogen pressure from the accumulators to the atmosphere through the normally closed vent valves or l
through the nitrogen supply source via the normally closed air-operated valves EPHV8875A, B,
C, and D, in series with check valve EPV0046 and normally closed air-operated EPHV8880 is not treated as a potential fault of the system because the probability of such faults is considered insignificant relative to other system faults.
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Attachment I to CO 95-0001 Page 8 of 10 Each accumulator has two independent level and pressure transmitters.
These transmitters provide indication and alarm to the operators on the level and pressure in the accumulators. Miscalibration of the transmitters could result in an improper amount of borated water at the wrong pressure for injection into the RCS.
With the level taps only 16 inches apart, this does not result in a great effect.
The accumulator volume at the lower tap (equal to 0%
level) is 5916 gallons. Although physically impossible, if it were postulated that the level transmitter were grossly miscalibrated such that the 100% span was at the bottom instrument tap, the accumulator volume would be off by a maximum of 850 gallons. This is 206 gallons below the Technical Specification minimum.
The operators, however, maintain the level in the accumulators between the low (31%) and high (69%) level alarms.
This alarm band is between 6180 and 6505 gallons.
If we assume the level miscalibration described above, this will result in the accumulator volume being 795 gallons, or 13% below the Technical Specification minimum.
MAAP' cases performed during the IPE and in support of this proposed license amendment request used a significantly smaller volume of water (i.e.,
one accumulator) and were still successful in preventing core damage.
Some of these MAAP cases used a reduced accumulator pressure of 614.7 psig and were also successful in preventing core damage.
Since the pressure in the MAAP runs was higher than the minimum Technical Specification pressure, a basic event for the miscalibration of the pressure transmitter has been applied to the accumulator tree, using a calibration error frequency of 3.0E-4 found in the Surry PSA (NSAC-152, Vol. 3).
The MAAP runs discussed here are not intended to be in conflict with the deterministic LOCA analyses in Section 15.6.5 of the USAR.
Event Tree Models In the WCGS IPE, Event Trees for Large, Medium and Small LOCA contain a node for the ECCS accumulators (IPE Report Figures 3.1-2, 3.1-3, and 3.1-4 respectively). These were conservative assumptions established early in the development of the IPE.
It should be noted that the accumulator top node placement in the Medium LOCA Event Tree is the same as for a Large LOCA (Initiating event occurs; accumulators fail; early core damage).
Credit for high pressure injection is taken only after successful accumulator injection.
Successful MAAP runs in support of this proposed increase in accumulator AOT indicate that the accumulators are not required in the Medium LOCA Event Tree to prevent core damage.
In spite of this, results of the increased AOT effect on the Medium LOCA core damage sequence frequency will be shown.
MAAP is a computer code that quantitatively predicts the evolution of a severe accident sequence starting from full power conditions given a set of system faults and initiating events through events such as core melt, reactor vessel failure, and containment failure.
Its use is described in Section 4.2 and 4.3.4 of the WCGS IPE.
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A'ttachment I to CO 95-0001 Page 9 of 10 I
l Impact on Core Damage Frequency The unavailability of the accumulators in the IPE was assumed to be 4.28E-06.
J To demonstrate the minimum impact of the accumulator AOT change to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, it was decided to calculate the effect on the WCGS CDF using an accumulator test and maintenance unavailability of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> per year (1.14E-02).
The results of the calculations of the accumulator injection fault tree are summarized in Table 1.
The Base Case is from the WCGS IPE.
Case ACCAOT1 added a test and maintenance unavailability of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.
Case ACCAOT2 i
eliminated one accumulator from the fault tree and changed the success criteria to two-out-of-two accumulators inject into two-out-of-two RCS cold legs.
(Note: One accumulator is assumed to spill its contents out the broken loop and a second is out of service for one year.)
Case ACCAOT3 was the same as the Base Case and added a basic event for the miscalibration of accumulator pressure transmitters. Case ACCAOT4 was the same as the ACCAOT3 plus the accumulator test and maintenance unavailability of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.
It is seen that a miscalibration of the accumulator pressure transmitters may have a much larger impact on the accumulator unavailability than the allowed outage time.
TABLE 1 Accumulator Injection Fault Tree Results Accumulator Unavailability General Description Injection Fault per Year Trees IPE Fault Tree ACC Same ACCAOT1 8.84E-06 Base Case +100 Hour Test & Maint.
ACCAOT2 4.01E-04 Base Case +1 Year Test & Maint.
ACCAOT3 3.04E-04 Base Case + Pressure Trans.
ACCAOT4 3.09E-04 Base Case + Pressure Trans. + 100 Hour Test and Maint.
The requantification of the core damage sequences associated with the ECCS accumulators are summarized in Table 2.
The results indicate that the increase in the core damage with a 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> allowed outage time of one accumulator is not significant.
There is a more notable difference when one accumulator is assumed to be out of service for one year.
However, this is far beyond the request of the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AOT.
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Ahtachment I to CO 95-0001 Page 10 of 10 TABLE 2 Core Damage Sequence Requentification Results Initiating Event CDF From IPE Event Tree Seq. #
Accumulator AOT Accumulator AOT 5
Frequency per
(% of Total (CDF) of 100 Hours of 1 Year year CDF)
Section 3.1 (ACCAOT1)
(ACCAOT2).
Section 3.1.1 Table 3.4-2 (Note 1]
CDF CDF
=
(Note 1]
[ Note 1]
(% Increase)
-(% Increase)
I*
1.37E-6 LLO #4 2.14E-09 4.42E-09 2.00E-07 (3.28%)
(<0.0054%)
(0.47%)
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1.85E-6 MLO #8 4.71E-09 1.10E-03 (0.012%)
(1.04%)
g g)
(Note 2]
[ Note 2) 6.67E-7 SLO #10
<1.28E-13 2.50E-03 (1.59%)
Notes:
1)
The Section references are to the WCGS IPE Report.
2)
CDF calculation shown for information only. MAAP runs indicated that the accumulators are not necessary to prevent core damage i
given a Medium LOCA.
i 3)
Small LOCA sequences were not requantified since their. base CDF was 4 orders of magnitude lower than Large LOCA.
Any increase in i
CDF at this level is insignificant.
l The above PSA discussion does not eliminate the actions required when one accumulator is inoperable; however, it does indicate that the consequences of extending the AOT to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are acceptable.
Based on the above discussions and the considerations presented in Attachment II, the proposed change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report; or create'a possibility for an accident or malfunction of a different type that any previously evaluated in the safety analysis report; or reduce the margin of safety as defined in the basis for any technical specification.
Therefore, the proposed change does not adversely affect or endanger the health or safety of the general public or involve a significant safety hazard.
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A'tachment II to CO 95-0001 t
Page 1 of 4 ATTACHMENT II NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION s
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'A'ttachment II to CO 95-0001 Page 2 of 4 No Significant Hazards Consideration Determination This license amendment request proposes revising Technical Specification 3.5.1 and its associated Bases.
A new Action Statement would be added to Technical Specification 3.5.1 that would provide an allowed outage time (AOT) of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if one accumulator is inoperable due to its boron concentration not meeting the 2300-2500 ppm band.
The current action statement would be followed if an-accumulator is inoperable for any other reason.
The AOT for the current action statement would be increased to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in lieu of the current AOT of one hour.
Surveillance 4'.5.1.1.a.1 and 4.5.1.1.b would be revised and Surveillance 4.5.1.2 would be deleted in accordance with the guidance of NRC Generic Letter 93-05 and NUREG-1366.
Technical Specification Bases Section 3.5.1 would also be revised to discuss the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AOTs for the action statements associated with Technical Specification 3.5.1.
Standard I - Involves a Significant Increase in the Probability or Consequences I
of an Accident Previously Evaluated The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The overall' protection system performance will remain within the bounds of the accident analyses documented in Chapter 15 of the Updated Safety Analysis Report, WCAP-10961-P, and WCAP-11883 I
since no hardware changes are proposed.
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The safety injection accumulators are credited in Section 15.6.5 of the Updated Safety Analysis Report for large and small break LOCA.
There will be no effect
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on these analyses, or any other accident analysis, since the analysis assumptions t
are unaffected and remain the same as discussed in Section 15.6.5.
Design basis accidents are not assumed to occur during allowed outage times covered by the Technical Specifications.
As such, the ECCS Evaluation Model equipment availability assumptions made in Section 15.6.5 remain valid.
The safety injection accumulators will continue to function in a manner consistent with the above analysis assumptions and the plant design basis.
As such, there will be no degradation in the performance of nor an increase in the i
number of challenges to equipment assumed to function during an accident situation.
The proposed technical specifications changes do not involve any hardware changes nor do they affect the probability of any event initiators.
There will be no change to normal plant operating parameters, ESF actuation setpoints, accident mitigation capabilities, accident analysis assumptions or inputs.
Therefore, these changes will not increase the probability of an accident or malfunction.
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Attachment II to CO 95-0001 Page 3 of 4 The corresponding increase in CDF due to the proposed change to increase the AOT of the accumulators from one hour to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is insignificant.
Pursuant to the guidance in Section 3.5 of NSAC-125, the proposed increase in AOT does not
" degrade below the design basis the performance of a safety system assumed to function in the accident analysis," nor does it " increase challenges to safety systems assumed to function in the accident analysis such that safety system performance is degraded below the design basis without compensating effects."
Therefore, it is concluded that these changes do not increase the probability of occurrence of a malfunction of equipment important to safety.
Standard II - Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated i
The proposed change does not create the possibility of a new or different-kind of accident from any accident previously evaluated.
This change is administrative in nature and does not involve any change to the installed plant systems or the overall operating philosophy of WCGS.
No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of these proposed changes.
There will l
be no adverse effect or challenges imposed on any safety-related system as a result of these changes.
Therefore, the possibility of a new or different type of accident is not created.
There are no changes which would cause the malfunction of safety-related equipment, assumed to be operable in the accident analyses, as a result of the proposed technical specification changes.
No new mode failure has been created y
and no new equipment performance burdens are imposed. Therefore, the possibility of a new or different malfunction of safety-related equipment is not created.
Standard III - Involve a Significant Reduction in the Margin of Safety The proposed change does not involve a significant reduction in a margin of safety.
There will be no change to the Departure from Nucleate Boiling Ratio (DNBR) Correlation Limit, the design DNBR limits, or the safety analysis DNBR limits discussed in Bases Section 2.1.1.
l As discussed previously, the performance of the accumulators will-remain within the assumptions used in the large and small break LOCA analyses, as presented in l
USAR Section 15.6.5.
Also, there will be no effect on the manner in which safety limits or limiting safety system settings are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions.
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4 nttachment II to CO 95-0001 Page 4 of 4 Based on the above discussions it has been determined that the requested technical specification revision does not involve a significant increase in the probability or consequences of an accident or other adverse condition over previous evaluations; or create the possibility of a new or different kind of accident or condition over previous evaluation; or involve a significant reduction in a margin of safety.
The requested license amendment does not involve a significant hazards consideration.
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Attachment III to CO 95-0001 Page 1 of 2 ATTACHMENT III ENVIRONMENTAL IMPACT DETERMINATION i
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l Attachment III to CO 95-0001 Page 2 of 2
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l Environmental Impact Determination l'
10 CFR 51. 22 (b) specifies the criteria for categorical exclusions from the requirements for a specific environmental assessment per 10 CFR 51.21.
This amendment request meets the criteria specified in 10 CFR 51.22 (c) (9).
The specific criteria contained in this section are discussed below.
(i) the amendment involves no significant hazards consideration l
As demonstrated in the No Significant Hazards Consideration Determination in I
Attachment II, the requested license amendment does not involve any significant hazards consideration.
(ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite i
The requested license amendment involves no change to the facility and does not involve any change in the manner of operation of any plant systems involving the generation, collection or processing of radioactive materials or other types of effluents.
Therefore, no increase in the amounts of effluents or new types of effluents would be created.
(iii) there is no significant increase in individual or cumulative occupational radiation exposure The requested license amendment involves no change to the facility and does not involve any change in the manner of operation of any plant systems involving the generation, collection or processing of radioactive materials or other types of effluents.
Furthermore, implementation of this proposed change will not involve work activities which could contribute to occupational radiation exposure.
Therefore, there will be no increase in individual or cumulative occupational radiation exposure associated with this proposed change.
Based on the above it is concluded that there will be no impact on the environment resulting from this change.
The change meets the criteria specified in 10 CFR 51.22 for a categorical exclusion from the requirements of 10 CFR 51.21 relative to specific environmental assessment by the Commission.
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