ML20034C577
| ML20034C577 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 04/26/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20034C575 | List: |
| References | |
| NUDOCS 9005040158 | |
| Download: ML20034C577 (87) | |
Text
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p asog'o UNITED STATES g
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NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D. C. 20666 j
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION I
RELATED 'TO AMENDMENT NO.125 l
i' TO FACILITY OPERATING LICENSE NO. DPR-61 CONNECTICUT YANKEE ATOMIC POWER COMPANY I
l l
HADDAM NECK PLANT l
I DOCKET NO. 50-213 l
PART 1 - Reviews the reformatting of all current Technical Specification i
sections except for Sections 3.6, 3.7, 3.12, 4.3 and 4.5.
PART 2 - Review of changes to the Technical Specifications-to reflect l
modifications implemented by the end of Cycle 15.
i PART 3 - Review of changes to the Technical Specifications to reflect installation of additional fire protection features.
PART 4 - Review of changes to,the Technical Specifications as proposed by Generic Letter 88-16 " Removal of Cycle-Specific Parameter Limits from Technical Specifications.
i PART 5 - ReviewofchangestotheTechnicalSpecificationsto1) incorporate-degraded grid voltage protection requirements; 2) incorporate emergency diesel generator requirements of Generic Letterc84-15
" Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability;" 3) incorporate industry improvements;. 4) change custom Technical Specification format to one that is similar to the Westinghouse Standard Technical Specification format; and
- 5) incorporate requirements for battery discharge testing as required by the Systematic Evaluation Program Topic VIII-3.A.
PART SA-Review of changes to the Technical Specifications _related to the
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electrical power systems and the degraded grid undervoltage setpoints.
PART 6-Review of changes to the Technical Specifications _to reflect installation of a new reactor protection system and nuclear instrumentation system.
t DATE: April 26,1990 9005040158 900426 PDR ADOCK 05000213 P
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PART 1 0F SAFETY EVALUATION i-RELATED TO AMENDMENT NO.125 l
1.0 INTRODUCTION
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By submittals dated October 26,_1988, March 6, June 2, June 23, July 28, and August 4, 1989, and supplemented by submittals on August November 22, 1989, Connecticut Yankee Atomic-Power Company (CYAPC021,1989,)
proposed to upgradetheircurrentcustomformatTechnicalSpecifications(TS)tothe Westinghouse Standard-format Technical Specifications (WSTS).
All sections of the current custom TS will be reformatted in this-proposed TS except for Sections 3.6, " Core Cooling Systems," 3.7, " Minimum Water Volume and Boron Concentration in the Refueling Water Storage Tank," 3.12 " Station Service Power," 4.3, " Core Cooling Systems-Periodic Testing" and 4.5, " Emergency Power System Periodic Testing".
Sections 3.6, 3.7 and 4.3 were reformatted by Amendment No. 121.
Sections 3.12 and 4.5 will be reformatted by amendment -
request dated August 29, 1988.
2.0 DISCUSSION l
As part of the Systematic Evaluation Program (SEP), CYAPC0 committed to convert their custom TS to the WSTS.
In a meeting on September 20, 1988,-CYAPC0 proposed i
to submit the TS conversion-packages over a three month period beginning-l October 1988. With the impending issuance of the revised WSTS (MERITS), the staff proposed that it would be advantageous for CYAPC0 to await the issuance of the revised WSTS before addressing the full WSTS conversion.
In the-interim, the staff' agreed that the custom TS format could be upgraded to the current WSTS format.
The staff concluded that this interim step:would:
1)providea substantially improved TS while facilitating the future conversion effort to the revised WSTS, 2) provide definitive LCO and Action statements for several safety relatedsystems,3)eliminatetheuseofadministrativeTS,4)providea mechanism to close prior TS commitments associated with NUREG-0737, SEP and variousotherGenericLetter(GL) recommendations,and5)eliminateambiguities inherent with the wording and format of the current TS.
Based on the above. the staff concluded that the revised TS would enhance public safety and therefore justified this interim step to improve the Haddam Neck TS.
The staff has-informed CYAPC0 several tilnes that this TS upgrade does not fulfill. CYAPCO's SEP commitment to convert to.the WSTS.
This. amendment is one of several that is part of the TS upgrade.
By letter dated September 22, 1987, the NRC provided Northeast Utilities with an accept-able revision of the WSTS. The TS upgrade will be using the provided WSTS revision as a guide for the format change while maintaining the current TS requirements.
Since this upgrade is primarily a format change. the staff did not pursue all deviations and omissions from the provided WSTS with the same-intensity as would have been done for a WSTS conversion.
Therefore, if the
2 proposed TS omitted portions of the requirements that appear in the provided WSTS revision and these same requirements did not already exist in the current TS, the review of these omissions will be deferred to the full WSTS conversion.
However, where new TS statements have been proposed (statements not previously found in the current TS) that deviate from the provided WSTS revision, a review of the deviation will be given.
The deviations will be reviewed in part,. based on three previously agreed upon criteria:
- 1) plant specific design, 2) previously approved hardware, structural or organizational changes, and 3) past operating experiences that can be shown to provide an equivalent degree of protection to that provided by the WSTS. Any deviations from the current custom TS will also be reviewed.
The format change and the additional restrictions resulting from this amendment make substantial improvements in the clarity and readability of the TS.
As a result, the staff considered this TS upgrade beneficial from both a public safety and an operational perspective.
3.0 EVALUATION The evaluation has been divided into two sections.
Section I will address proposed TS that are consistent with the provided WSTS and/or the current TS.
In addition, many of these TS sections add restrictions to the current TS.
Section II will address proposed TS that relax restrictions from either the current TS or the provided WSTS revision.
As noted earlier, the staff did not perform a " completeness" review to ensure that all sections of the WSTS were included in this format change. Therefore, this review will exclude the review of complete omissions of WSTS sections that did not already exist in the current TS.
Each of the deviations will be addressed individually.
If a GL or a SEP issue has been addressed by the proposed TS change then it will also be noted.
3.1 Section I Previously, the NRC staff provided a version of the WSTS to CYAPC0 and excluding plant specific alterations, stated the provided WSTS would be an acceptable guidance for a STS conversion. Although this amendment is not a STS conversion, the amendment does follow the guidance of this WSTS revision.
The logic for this TS upgrade has been stated in the Discussion section of this Safety l
Evaluation.
The staff review has determined that all sections of the propsed TS except for those discussed in Section 3.2 of this Safety Evaluation are consistent with the current TS and/or the WSTS, impose added restrictions to the current TS, and/or add restrictions that do not currently exist. Therefore, the proposed TS sections except for those delineated in Section 3.2, are administra.
tive in nature (format change) or provide additional limitations, restrictions, or, controls not previously included in the Haddam Neck TS.
In addition, the NRC staff has provided Table 1 which provides a list of all sections of the current TSs and where those TS sections (TS sections from the custom TS) now exist in the proposed TS.
This was done to verify that all sections and requirements of the current TS are incorporated in the proposed TS or that justification for deletion or modification of a current TS is provided.
The staff has concluded that the safety significant requirements of the current TS have been maintained in the proposed TS.
Based on the above, the staff concluded that the proposed TS are acceptable and provide an_ equivalent and in some areas an enhanced set of TS to the current custom TS.
3.2_Section II The TSs reviewed in this section will be addressed by number and subsection as it appears in the proposed TS. As noted earlier the WSTS refers to the WSTS revision provided to CYAPC0 by letter dated September 22, 1987.
A)
October 26, 1988 Submittal
- 1) Section 1, Definition, Table 1.1, Frequency Notation The definition of "S" in Table 1.1 has been changed from "at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />" to "at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />." The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> limit while consistent.with WSTS is a relaxation from the current-TS. CYAPC0 states that the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> frequency does not provide any latitude within an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift in which to perform surveillances that are required once per shift.
That is, the once per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift checks would have to be performed at exactly the same time interval or less within each shift.
CYAPC0 maintains that the surveillances notated with an "S" will be performed each shift, with a shift being 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> time limit will provide latitude within the shift to allow for scheduling and operational perturbations which could affect the timing of certain activities.
The staff believes the intent of the TS is to require a check once per shift and this requirement will be maintained.
Based on the above, the staff concludes that the proposed TS is acceptable.
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- 2) Sections 3.2.3.1.1, 3.2.3.2 and 3.2.4 The existing TSs contain only trip setpoints. The proposed TSs contain both trip and allowable setpoints.
The trip setpoints of the existing TS are equivalent to the allowable setpoints of the proposed TS.
The proposed trip setpoint is now 72% instead of 74% as the setpoint was written in the custom TS. The proposed trip setpoint has been set 2% lower to account for
' instrument drift, expected to be a maximum of 25. This ensures that the allowable value (74%) is not violated at any time between calibrations.
Based on the above, the staff concludes that the proposed TS is acceptable.
- 3) TS Table 5.7.1 Table 5.7.1 provides a list of reactor vessel design transients and the
- maximum permissible number of design cycles.
The list of transients is-different than that provided by the WSTS.
CYAPC0 states that their list provides a list of all transients which have been analyzed for cyclic design restrictions. As. modifications and analysis are revised and-updated, Table 5.7.1 will be revised to reflect the latest analysis.
l This table does not currently exist in the current TS.
Based on the above, the staff has concluded that the proposed TS meets the intent of the WSTS and represents all currently analyzed component cyclic or transient limits.
The staff concludes that the proposed TS is acceptable.
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. 4) TS'3.9.11
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TS 3.9.11 required a minimum of 20 feet of water be maintained over the l
top of the irradiated fuel, seated in the storage racks.. WSTS recommends 23 feet. CYAPC0 states 21 feet-is the maximum possible due to the design of the spent fuel pool. While it is possible to fill the pool to provide-21 feet of water, this would expose certain equipment and components to' water / boric acid and could cause equipment / component failures.
The proposed level of 20 feet would limit water / boric acid exposure to various j
equipment, especially the carbon s uel sleeve gate operator.
CYAPC0 has calculatedthedecontaminationfanor(DF)for20feetofwaterasapproxi-mately 250. This is conservative compared to the DF of 100 for iodine assumed in the fuel handling accident and the DF-of 133 recommended in Regulatory Guide 1.25, Revision 2.
While this is a deviation, the 20 feet of water provides an adequate degree of protection for any fuel handling accident. Based on the above the staff concludes that this TS is acceptable.,
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- 5) Section 4.9.6.2 This surveillance requirement specifies a load test weight to be 125%
of the weight of the load to be lifted be performed. The-WSTS requires a load-test of a fixed weight.
CYAPC0 states this TS provides.some flexibil-ity in the loads to be lifted. CYAPC0 states that the load test weight is consistent with the guidelines of ANSI B30.2 and will not exceed the rated
'i load capacity of the hoist.
Based on the above, the-staff concludes that the surveillance provides an equivalent degree of protection to the WSTS and therefore the TS is acceptable.
TS 5.3.1 The proposed TS allows the fuel assemblies.to consist of 1) fuel rods clad with Type 304 stainless steel, 2) filler rods fabricated from Type 309 stainless steel or 3) vacancies as justified by the cycle-specific reload analysis. The current TS requires that the fuel' assemblies consist only of fuel rods' clad with Type 304 stainless. steel. The proposed change provides flexibility to deviate from a fixed number of fuel rods per assembly. This is desirable because it permits timely removal of fuel rods that are found to be leaking during a refueling outage or are determined to be probable sources of future leakage.
Approval,of the proposed. change will-allow improvement in the licensee's fuel performance, which will provide for reductions in future occupational radiation exposure and plant radiological releases.
Under-the proposed change, limitations on fuel rod substitution or omissions'and limitations regarding core locations are those implicit in the justifying analyses required to be performed by the licensee for each fuel cycle using NRC-approved :.ethodology to demonstrate that existing design limits and safety analyses continue to be met.
The term "NRC-approved methodology" includes those methodologies acknowl-edged in the Final Safety Analysis Report and applied in support of issuance of the original operating license for the Haddam Neck Plant.
Additionally, it includes those subsequent methodologies that have been submitted to and accepted by the staff as amendments to the operating license.
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c The requirement for special reporting is consistent with existing TS 6.9.2 and is:necessary to keep NRC informed in the event a significant deviation from past fuel performances should be observed during a refueling outage.
The' licensee has proposed changes to Specification 5.3.A that are consis-L tent with the guidance provided in Generic Letter 90-02, " Alternative-L Requirements for Fuel Assemblies in the Design Features Section of 1
l Technical Specifications." Therefore, the staff has deferred approval of this request to the resolution of GL'90-02.
B)
March 6, 1989 Submittal 1)TS3.1.2.2and4.1.2.2.d The proposed TS requirement differs from the WSTS in-that the required three flow paths are from the boric acid tanks (BAT) rather than one path from the BAT tank and two paths from the-RWST and that the flow test surveillance does not specify a flow l.
rate for the BAT flow paths. The boration system ensures that negative reactivity control is available during each Mode of normal operation and-for abnormal operational: occurrences. At the Haddam Neck Plant, the boric acid concentration in the RWST is significantly lower than that in the BAT. As a result, the limiting case for operation is when.the metering pump.is l
l used to inject borated water. 'The metering pump'cannot inject I
sufficient boric acid into the RCS from the RUST to provide the required shutdown margin.
Because of-post-LOCA chemistry requiremeni.s the boric acid concentration in the RWST is.. bounded in the TS. Therefore, CYAPC0 cannot use the RWST as a required water source for reactivity control; and the boration capability l
to ensure the shutdown margin in all Modes provided by the proposed TS 3/4.1.2.2 can only be provided by the' BAT.
Ac-cordingly'TS 3.1.2.2, Flow Paths-Operating, only references the three flow paths from the BAT to-the charging / metering pumps.
Although the RWST flow path to the charging / metering pumps is not credited for reactivity control, the RWST flow path-to the charging pumps is required to be available by TS 3/4.5.1, ECCS Subsystem-Tavg Greater Than Or Equal To 350' F and TS 3/4.5.2, ECCS Subsystems-Tavg Less Than Or Equal: To 350* F.
The licensee also states that no flow instrumentation exist in-the BAT lines to determine flow. The licensee states that they.-
will demonstrate that the BAT lines to the charging pump suction are unobstructed. As allowed by the ground rules of the TS upgrade one of the basis for deviation is plant specific design.
Based on the above the staff concludes that-the proposed TS deviations are.a result of plant specific design and to obtain the WSTS format would require modification to the plant.. In addition, the proposed applicability and surveillance require-l ments are more restrictive than the current TS and the Action statement did not previously exist. Based on the above, the.
staff concludes the the TS meets the intent of the WSTS and provides at least-an equivalent degree of protection as the
- current TS and therefore is acceptable.
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- 2) TS 3.1.2.6
~The proposed TS differs from the WSTS because the RWST is not included.
As noted in the discussion of TS 3.1.2.2, the RWST is not a required water source for reactivity control consideration at the Haddam Neck Plant.
In-addition, the equivalent requirements (LCO, applicability, action and surveillance requirements) for the RWST exist in the Emergency Core
' Cooling Systems section of the proposed TS.. Based on the above, the staff concludes that the TS is acceptable, b
C)
June 2, 1989 Submittal
- 1) TS 3.7.1.2 This TS is for the auxiliary feedwater system.
The proposed TS is equiva-lent to or more conservative than the current TS and therefore by the groundrules of the conversion is acceptable.
However, this TS is also i
part of the GL 83-37 "NUREG-0737 Technical Specifications." The NRC staff i
has. concluded that the-proposed TS does not meet the intent of the GL 83-37 CYAPC0 and the staff have agreed-that this issue will be resolved in a future license amendment.
- 2) TS 3.6.1.5 and 4.6.1.5-The proposed TS does not include the specific-locations of where the i
temperature readings are to be made as specified in the WSTS. The locations and methodology for calculating containment average temperature was reviewed in Inspeccion Report 88-23.
The report concluded-that the dispersion of the resistance temperature detectors -(RTDs) adequately represents containment temperature.
However, during containment integrated leak rate test an additional RTD is necessary in the dome above the polar crane. While IR 88-23 has concluded that the calculated temperature adequately represents the containment, the-inspectors are still reviewing the RTD placements which will assure that the RTDs will provide a representative temperature of containment.
Based on the above, the staff concludes that the exact location of the RTDs need not be specified in the TS as the RTD placement-will be confirmed by future inspections.
l 3)
TS Table 3.3-3, Footnote for Items 4a, 4b, and 4c Table 3.3-3, Footnote for 4a, 4b and 4c states that the davice must :hange state within.95-1.05 seconds when the input voltage to the device goes i
from normal to zero volts instantaneously. The proposed change requires l
that the relays actuate when the input voltage decreases instantaneously from normal to 50 percent of the tap setting voltage.
By requiring the device to change state within one second 15 percent, when the input voltagetothedevicereducesfromnormalto50-percentoftapsetting i
voltage instantaneously, the relay is being challenged to operate in a real degraded voltage situation.
If the' input voltage were allowed to drop to zero, the time-voltage characteristics of the induction coil in the degraded voltage range would not fully be testeri.
A loss of all voltage would simply cause the relay to return to its de-energized state.
Since the proposed testing requirements will challenge the device in a degraded
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condition, the proposed change represents a more conservative test.
Furthernsore, the test is consistent with the plant's standard method of testing undervoltage relays of this type.
Based on the above, the staff concludes that the proposed TS change is acceptable.
l 4)
CYAPC0 added "to be repaired" to the TS.
Currently, the staff requires l
l that repairing of tubes requires a TS amendment.
The amendment would e
include the approval of a-sleeve specifically for use at the Haddam Neck l
Plant. The TS upgrade did not provide this information and therefore the staff does not. find this change acceptable.
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D)
June 23, 1989 Submittal 1)
Ti.ble3.3-2(3b)
The proposed action statement for the auxiliary feedwater system requires that with one less than the minimum channels operable restore the channel to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or reduce the thermal power to below 10%
of rated thermal power within the following hour. The current TS-would i
imply a shutdown on a loss of one channel with no specified time frame.
The WSTS would allow u) to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> with one less than the minimum channels operable but require.tle plant to shutdown if the channel cannot be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
The WSTS is applicable for MODES 1 and 2 while for Haddam-Neck the applicable mode is MODE 1. greater than'10% power.
CYAPCO states that below 10% power the plant operators would have more than adequate time to manually initiate the auxiliary feedwater pumps since the decay heat loads below 10% power are small.
In accordance with the FSAR, the auxiliary feedwater initiation system is defeated below 10% power.
With one channel inoperable the plant would have 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to repair the l
channel or reduce power to less than 10% where the auxiliary feedwater.
initiation system is defeated and the action statement would no longer be-applicable.
This action is similar'to the WSTS which provides a fixed time frame to restore the channel or place the plant in a condition for which the action statement is not applicable.
Due to hardware design, the inoperable channel cannot be placed in a tripped position.
Therefore, for a maximum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the plant would be without automatic auxiliary feedwater initiation from the trip of all main feedwater pumps.
This.is partially compensated for by the fact that automatic auxiliary feedwater initiation is still provided by low' steam generator water level.
CYAPCO's proposed TS provides a reasonable-compromise between the plant configuration and the WSTS.
Based on the above, the staff concludes the proposed TS is acceptable.
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- 2) TS 3.3.3.2 Actin :
The pro osed action =tatement for the movable incore de^ector. system would a low continued use of the system with less than the minimum number 4
of detector thimbles required if penalty factors are applied to the linear heat generation rate or quadrant power tilt; or during recalibra-tion of the system.
The staff currently requires that penalty factors be approved before they can be applied in such cases. Therefore, the staff denies this proposed action statement.
. 3) Proposed Deletion of Various Current TS Requirements
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a) Current TS 3.9.C CurrentTSrequiresthatneutronmonitorsineachrange(source, intermediate and power) shall be in continuous operation until at least one decade of reliable indication.is verified on the next range of l
instrumentation.
CYAPC0 has recently replaced their nuclear <nstrumenta.
tionsystem(NIS).
The new power range instrumentation cov rs the entire L
l rangeoftheoriginalequipment(from2005powerto-1X10"g% power).
The l
new source range and power range instruments are provided data.from the-l same detectors.
Therefore, there is no need to verify the decade overlap L
as the entire range is provided by the power instrumentation.
The staff agrees that this requirement can be deleted, b) Current TS 3.11.E Current TS 3.11.E requires the containment spray system to be operable whenever the reactor is critical. The containment spray system is an
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auxiliary system that is not credited for in any safety analysis.
Containment heat removal is provided by two 100% Containment Air Recirculation fan systems. The staff agrees that this requirement can be i
deleted from the TS.
c) Current TS 3.13.A Current TS 3.13.A requires radiation' levels in the containment and fuel storage building to be monitored continuously during refueling.- Radiation Monitoring of the containment and spent fuel building are part of the Refueling Procedures.
In addition, radiation monitoring is required;to be maintained in each area in which such licensed special nuclear material is handled, used, or stored by 10.CFR 70.24.
The staff agrees that this requirement can be deleted from the TS.
d) Current TS 3.13 F I
Current TS 3.13.F requires that whenever new fuel is added to the reactor L
core, a 1/m plot be maintained to verify the suberiticality of the core.
This requirement is not in the WSTS, and it does not have any corresponding limiting condition for operation. The 1/M surveillance is part of CYAPCO's Refueling Procedure and will be maintained there.
The staff agrees that this requirement can be deleted from the TS.
e) Current TS 3.13.H f
Current TS 3.13.H forbids the movement of spent fuel cask above the fuel pool or its edge until the NRC has received and approved the spent fuel cask drop evaluation.
In a letter dated June 28, 1985 GL 85-11, the NRC staff indicated that all licensees have completed the requirement to perform a review and submit a Phase I and Phase II report regarding NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants." The GL further stated that based on the improvements in heavy loads handling
a
.g-obtained from implementation of Phase I of NUREG-0612, further action is i
not required to reduce the risk associated with the handling of heavy I
loads (PhaseIIofNUREG-0612). Therefore, the staff concluded that a detailed Phase II review of heavy loads is not necessary and Phase II of i
NUREG-0612 is complete.
In that GL the staff recommended each licensee tosubmitalicenseamendmenttodeleteanyrequirementsrelatedtoheavy loads from the TS citing this GL as the basis.
CYAPC0 has stated that the only TS related to heavy loads is TS Section-3.13.H.
Based on the above, the staff agrees that TS 3.13.H can be deleted. However, the staff recommended that any actions identified by the licensee in regard to Phase II of NUREG-0612 should be implemented.
Therefore, all open items identified in CYAPCO's letter dated July 21, 1983 relating to Phase II, should be completed prior to the handling of spent fuel casks in the fuel-handling building.
f) Current TS 3.22, A.2, A.3, B.3, C.3, E.2.b and 6.3 The above TS sections require Special Reports be made to the NRC whenever the associated system of the TS is declared inoperable. CYAPC0 will 4
review all reportable events in accordance with the requirements of 10 CFR 50.72 as proposed in the upgraded TS Section 6.6.1.
The staff agrees that this section can be deleted and the reportability be provided under 10 CFR 50.72.
g) Current TS Table 4.2-1, Item 13 This item requires Charging Flow Indication be calibrated each refueling.
While this requirement is not in the WSTS, the staff does not believe sufficient bases has been provided to remove this TS re This surveillance will be maintained in TS Section 4.5.lf(4)quirement.
h) TS Table 4.2-1, Item 20 l
l This TS item requires calibration of the boric acid control system each l-refueling.
This system is used during normal operation of the plant-for 1
boric acid control and is not credited for in any design basis analysis.
When and if it becomes necessary to make a rapid addition of boric acid -
to the RCS, this flow element is bypassed as boric acid from the boric acid mix tank flows through a pump directly to the charging pump suction.
This system is calibrated routinely by procedure.
The staff concludes that this TS item can be deleted.
1 i) Current TS Table 4.2-2, Item 10 i
l This TS item requires Refueling System Interlocks to have a function check each refueling.
The testing of these interlocks is performed as part of the Refueling Procedures and there is no credit taken for these interlocks in any design basis analysis.
There are 13 interlocks to control motion of such things as the crane, bridge, fuel upender and the gripper tube..
The staff concludes that this item can be deleted from the TS.
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E)
July 28, 1989 Submittal L
- 1) TS 3.0.4, 4.0.3, 4.0.4 and associated Bases l
These statements deviate from WSTS and do not exist in this form in the l
current TS.
The proposed TSs reflect NRC guidance as recommended in GL l
l 87-09 for. improved wording and clarity.
The proposed wording recommended j
by the GL were incorporated in verbatim by the proposed TS. These changes l
represent part of the improved TS effort as encouraged by the staff and..
therefore are found to be acceptable.
1 1
- 2) TS 4.0.2 and associated Bases.
This statement deviates from the WSTS and does not exist in this form in the current TS.
The proposed TS reflect NRC guidance as recommended in GL 89-14 for improved wording and clarity.
The proposed wording recommended by the GL were incorporated in verbatim by the proposed TS.~
s Experience has shown that the 18-month surveillance interval, with the provision to extend it by 25 percent, is usually sufficient to accommodate normal variations in the length of a fuel cycle. However, the NRC. staff has routinely' granted requests for one-time exceptions to the 3.25 limit on extending refueling surveillances because the risk to safety is low in contrast to the alternative of a forced shutdown to perform these surveil-lances.
Therefore, the 3.25 limitation on extending surveillances has not been a practical limit on the use of the 25-percent allowance for extending surveillances that are performed on a refueling outage-basis.
The use of the allowance to extend surveillance intervals by 25 percent can also result in a significant safety benefit for.surveillances that are performed on a routine basis during plant operation.
This safety benefit is incurred when a surveillance interval is extended at a time that conditions are not suitable for performing the surveillance.
Examples of this-include transient plant operating conditions or conditions in which safety systems are out of service because of ongoing surveillance or maintenance activities.
In such cases, the safety benefit of allowing the use of the 25-percent allowance to extend a surveillance interval would outweigh any benefit derived by limiting three consecutive surveillance intervals to the 3.25 limit. Also there is'the administrative burden-associated with tracing the use of,the 25-percent allowance to ensure compliance with the 3.25 limit.
On the basis of these considerations, the staff concluded that remova? of the 3.25 limit will have an overall positive impact on safety.
This alternative to the requirements of Specification 4.0.2 will remove an unnecessary restriction on extending surveillance requirements.and wil1~
result in a-benefit to safety when plant conditions are not conducive to the safe conduct of surveillance requirements.
The removal of the 3.25 limit will provide greater flexibility in the use of the provision for extending surveillance intervals, reduce the administrative burden asso-ciated with its use, and have a positive effective on safety. Therefore, the staff concludes the proposed TS is acceptable.
0 e
. 4.0
SUMMARY
The staff has reviewed the proposed TS and as stated in Section 3.1 has determined that all of the safety significant current TS requirements will be maintained by the proposed TS.
Furthermore, the proposed amendment is an improved format over the current TS and incorporates numerous new TS limitations, restrictions or controls to plant operation.
Based on the considerations discussed in the above evaluation, the staff concluded that the proposed amendment will make overall improvements in the operational safety of the plant while maintaining the current safety analysis.
Therefore, the staff finds the proposed amendment to be acceptable.
5.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact have been prepared and published in the Federal Register on February 23,1990(55FR6563). Accordingly, based upon the environmental assessment, we have determined that the issuance of the amendment will not have a significant effect on the quality of the human environment.
5.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1)there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
A. Wang G. Garten l
,,e,
~
e w
Table 1 - Current T.S. # With Corresponding Proposed T.S. #
Existing T.S. #
Description P,roposed RTS-11. 0 Definition 1.1 Defined Terms 1.0 1.2 Thermal Power 1.32 1.3 Rated Thermal Power 1.25 1.4 Operation Mode 1.18 1.5 Not Used 1.6 Operability 1.17 1.7 Reportable Event 1.26.
1.8 Containment Integrity-1.6 1.9 Channel Calibration 1.4 1.10 Channel Check
- 1. 5 -
1.11 Channel Functional Test 1.2 1.12 Core Alteration 1.8 1.13 Shutdown Margin 1.28 1.14 Identified Leakage
-1.13 1.15 Unidentified Leakage 1.34 1.16 Pressure Boundary Leakage 1.20 1.17 Controlled Leakage 1.7 1.18 Quadrant Power. Tilt Ratio 1.22 1.19 Not Used 1.20 Not Used 1.21 Frequency Notation 1.12 Table 1.1 1.22 Not Used 1.23 Not Used 1.24 Axial Offset 1.3 1.25 Low Power Physics Test 1.19 1.26 Action 1.1 1.27 Channel Calibration 1.4 1.28 Channel Check 1.5 1.29 Channel Functional Test 1.2 1.30 Dose Equivalent I-131 1.9 1.31 Member (s)ofthePublic 1.15 1,32 Operable 1.17 1.33 Purge - Purging 1.21 1.34 Radioactive Waste Treatment Systems l'.23 1.35 Radiological Effluent Monitoring 1.24 and Off-site Dose Calculation Manual 1.36*
Site Boundary 1.29 1.37 Source Check 1.30 1.38 Unrestricted Area Same as Exclusion-area 1.39 Venting 1.35 Table 1.1 Operational Modes Table 1.2 Table 1.2 Frequency Notation Table 1.1
2-1 Existing T.S. #
Description Proposed RTS 2.0 Safety Limits and Maximum Safety e
Settings.
2.1 Introduction 2.1 Bases Section i
2.2 Safety Limits 2.2.1 Reactor Core 2.1.1 2.3 Reactor Coolant System Pressure 2.1.2 2.4 Maximum _ Safety Settings Protective.
2.2.1 Instrumentation i
Specifications Trip Setpoints l'
Item 1 Pressurizer Pressure Table 2.2 l Item 5 l
Item 2 Pressure Level Table 2.2-1 Item 6 Item 3 Variable Low Pressure
. Table 2.2-1 Item 4 Item 4 Nuclear Overpower Table 2.2-1 Item 2 Item 5 Low Coolant Flow Table 2.2-1 Item 7 Item 6 Reactor Coolant Loop Valve Section 4.4.1.7.1
- Temperature Interlock l
Item 7 High Steam Flow Table 2.2-1 Item 8 Item 8 High Start-up Rate Table 2.2-1 Item 3 3.0 Limiting Conditions for Operation 3,01 3.1 Introduction
- 3.01 3.2 Reactor Coolant System Activity 3/4.4.8 3.3.1.1 Start-up & Power Operation 3.4.1.1 a.
4 Loops 3.4.1.1.a b.
3 loop 3.4.1.1.b Applicability 3.4.1.'1 Action 3.4.1.1 Surveillance (1 4.4.1.1.1 Surveillance (2 4.4.1.1.2 i
g--v w er-e u-e--
3-Existing T.S. #
Description.
Proposed RTS 3.3.1.2 Hot Standby 3.4.1.2 a.
Reactor Trip Breakers 3.4.1.2.a Closed b.
Reactor Trip Breakers.
3.4.1.2.b J
Open Applicability 3.4.1.2 Action 3.4.1.2 Surveillance a 4.4.1.2.1 Surveillance b 4.4.1.2.2 Surveillance c 4.4.1.2.3 Surveillance d 4.4.1.2.4-3.3.1.3 Hot Shutdown 3.4.1.3 &-1.2.7
^
Applicability 3.4.1.3 Action 3.4.1.3 Surveillance (a 4.4.1.3.1' Surveillance (b 4.4.1.3.2 Surveillance c 4.4.1.3.3 Surveillance d 4.4.1.3.4 3.3.1.4.1 Cold Shutdown - Loops Filled 3.4.1.4.1 a.
RHR Loop 3.4.1.4.la b.
SG Water Levels 3.4.1.4.lb Applicability 3.4.1.4.1 Action 3.4.1.4.1 4
Surveillance a 4;4.1.4.1.1 Surveillance b 4.4.1.4.1.2 Surveillance c 4.4.1.4.1.3 Surveillance d 4.4.1.4.1.4 3.3.1.4.2 Cold Shutdown - Loops Not Filled 3.4.1.4^.2 Applicability 3.4.1.4.2 Action 3.4.1.4.2 Surveillance a 4.4.1.4.2.1 Surveillance b 4.4.1.4.2.2 Surveillance c 4;4.1.4.2.3 3.3.1.5 Isolated Loops 3.4.1.5 & 3.4.1.6 Applicability 3.4.1.5 & 3.4.1.6-Action 3.4.1.5 & 3.4.1.6 Surveillance 3.4.1.4 & 3.4.1.6 3.3.1.6 Isolation Loop Start-up 3.4.1.7 l
Applicability 3.4.1.7 Action 3.4.1.7 Surveillance a 4.4.1.7.1 l
Surveillance b 4.4.1.7.3 Surveillance c 4.4.1.4.3 l
l l
i L -
/
+
Existing T.S. #
Description Proposed RTS p"
l' 3.3.1.7 Idled Loop
'3.4.1.8 & 3.4.1.9 Applicability 3.4.1.8 & 3.4.1.9 Action 3.4.1.8 & 3.4.1.9 Surveillance a 4.4.1.8.1 & 4.4.1.9.1 Surveillance b 4.4.1.8.2 & 4.4.1.9.1 3.3.1.8 Idled Loop Start-up 3.4.1.10 & 3.4.1.11 Applicability 3.4.1.10 & 3.4;1.11 Action 3.4.1.10 & 3.4.1.11
-Surveillance a 4.4.1.10 & 4.4.10.11.1 L
Surveillance b 4.4.1.10.2 L
. Surveillance c 4.4.1.10.3 & 4.4.1.11.2 3.3.2.1 Safety Valves-Shutdown 3.4.2.1 Applicability 3.4.2.1 Action 3.4.2.1 Surveillance 4.4.2.1 3.3.2.2 Safety Valves - Operation 3.4.2.2 Applicability 3.4.2.2 Action 3.4.2.2 Surveillance 4.4.2.2 3.3.3 Pressurizer 3.4.3 Applicability 3.4.3 i
Action 3.4.3 I
Surveillance (a) 4.4.3.1 l
Surveillance (b) 4.4.3.2 3.3.4.1 Relief Valves 3.4.4 Applicability 3.4.4 Action 3.4.4 Surveillance 4.4.4.1 Surveillance 4.4.4.2 Surveillance 4.4.4.3 Surveillance
'4.4.4.4 l
Surveillance 4.4.4.5 3.3.4.2 Low Temperature Overpressure 3.4.9.3 Protection System a.
SLRV 3.4.9.3a b.
RCS Vent 3.4.9.3b Applicability 3.4.9.3 Action 3.4.9.3 Surveillance 4.4.9.3.1.
Surveillance 4.4.9.3.2
j 6-
, l
. Existing T.S. #
Description-Proposed RTS l
1 3.3.5-1 3.3.5.1 Reactor Coolant System Vents 3.4.11 1
Applicebility 3.4.11 y
Action 3.4.11 Surveillance 4.4.11.a Surveillance 4.4.11.b
' Surveillance 4.4.11.c l
3.4 Combined Heatup, Cooldown and 3-Pressure Limitations 3.4.A Reactor Vessel.
3.4.A.1 RCS pressure and temperature 3.4.9.lc
.l During hydrostatic and leak testing.
3.4.A.2 RCS pressure and temperature 3.4.9.1 heatup and~cooldown 3.4.A.3 Average rate of RCS temp Change 3.4.9.1.a and b of RCS Temp. Change 3.4.A.4 Allowable Pressure - Temp 3.4.9.1 Combinations 3.4.8 Pressurizer 3.4.B.1 500 psig. Limit 3.4.9.2.d 3.4.B.2 Heatup Rate 3.4.9.2.a.
3.4.B.3 Cooldown Rate 3.4.9.2.b 3.4.B.4' Temperature Difference 3.4.9.2.c 3.4.C.1 Steam Generator Pr/ Temp 3.7.2.a 3.4.C.2 Max heat up/cooldown 3.7.2.c 3.4.C.3 Tube sheet temp:
3.7.2.d 3.4.C.4 SG vessel temp 3.7.2.b 3.4 Applicability 3.7.2 2
'3.5 Chemical and Volume Control System 3.5.A.1 Charging Pumps
_3.1.2.2.a-& 3.1.2.4 3.5.A.2 Boric Acid Pumps 3.1.2.2.b 3.5.A.3 Boric Acid Tank 3.1.2.6 3.5.A.4 Maintenance 3.1.2.6 3.5.A.5 Flow Paths
~3.1.2.2.a 3.5.A.6 Valve BA-V-399 3.1.2.1 & 3.1.2.2.
3;5.B RCS Cold Legs Less than 315*F 4.1.2.3.3 & 3.1.2.4 3.6 Administrative Core Cooling System Technical Specification h,
i s
--e
. ~<
o I Existing T.S. #
Description Proposed RTS 3.6.A 1 Applicability 3.6.1 3.6.A-1.1 Pumps 3.6.1.a.1.2,3,5 j
3.6.A-1.2 RHR heat exchangers 3.6.1.a.1.4 3.6.A-1,3 Flow paths 3.6.1.a.1.6 & 3.6.1.b 3.6.A-1.4 One ECCS train 3.6.a Action Inoperable 3.6.A-II Applicability 3.6.2 3.6.A-II.1 One charging pums 3.6.2.a.1 3.6 A-l!.2 One RHR 41 eat exc1 anger 3.6.2.a.2 3.6.A-II.3 One RHR ') ump 3.6.2.a.3 3.6 A-l!.4 Flow pat 1s 3.6.2.a.4 4 3.6.2.b
{
3.6 A-II.5 No ECCS train 3.6.2 Action a o>erable because of one c1arging pump or flow path inoperable 3.6 A-II.6 No ECCS train 3.6.2 Action b operable because of the RHR pump or RHR heat exchanger inoperable 3.6 Core Cooling system 3.6 A See Administrative Technical i
=
Specification 3.6 A-1.1 3.6.B.1 Valve operability once Surveillance per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Requirements (SR)(a) 3.6.B.2 Valve operability on SR(b) startup prior to entering Mode 4 3.6.C Actions to disable Section3.6.2(SR)(b)
HPSI pumps 3.6.0 Actions to disable the Section3.6.1(SR)(b) l Centrifugal Charing i
pump 3.7 RWST Volume and Boron 3.1.2.5b, 3.6.3a, and 3.6.3b 3.8 Turbine Cycle 3.7.1 3.8.A.1 Safety Valves-Steam 3.7.1.1/
relieving capability ~
Table-3.7.-1 3.8.A.2.a Steam driven AFW pumps 3.7.1.2 3.8 A.2.b One AFW pump inoperai)le 3.7.1.2.a 3.8.A.2.c Two AFW pumps inoperable 3.7.1.2.b 3.8.A 3.a DWST/PWST min. Vol.
3.7.1.3 3.8.A.3.b DWST inoperable 3.7.1.3.a 3.8.A.3.c PWST inoperable 3.7.1.3.b 3.8.A.4 System piping 3.7.1.1 & 3.7.1.2 &
3.7.1.3 & 3.7.1.5 i
l
-7 Description Proposed RTS f
Existing T.S. #
3.8.B.1 AFW actuation system Table 3.3-2 instrumentation Item 3 3.8.8.2 AFW actuation contacts Table 3.3-2 and relays Item 3 i
Table 3.8-1 AFW actuation system Table 3.3-2 instrumentation Item 3 3.9 Operational Safety Instrumentation and Control Systems A
Logic Required for Full Power Table 3.3-1 Operations Table 3.3-2 B
Required Action if Logic Falls Table 3.3-1 Below 3
Limit Table 3.3-2 i
C Neutron Monitoring Note (3)
D Accident Monitoring Inst.
Table 3.3-7 Channel E
Required Action Table 3.3-7 Table 3.9.-1 Minimum Instrumentation Operating Conditions Item 1 Nuclear Overpower Reactor Trip Table 3.3-1 Item 2 s
Item 2 Pressurizer Variable Low Table 3.3-1 Pressure Reactor Trip Item 4 Item 3 Pressurizer Fixed High Pressure Table 3.3-1 Trip Item 5 Item 4 Pressurizer High Water Level Table 3.3-1 Reactor Trip Item 6 Item 5 Reactor Coolant Flow Table 3.3-1 Item 7 Item 6 Pressurizer Pressure Low Table 3.3-2 Item 1 Item 7 Deleted Item 8 Hanual Trip Table'3.3-1 Item 1 Item 9 Steam - Feedwater Flow Mismatch Table 3.3-1 Item 9 Item 10 High Steam Flow Table 3.3-1 Item 8 Item 11 Containment High Pressure Table 3.3-2 Item 5 Start-up Equipment Intermediate Range SUR Reactor Trip Table 3.3-1 Item 3 Source Range SUR Rod Stop Table 3.3-1 Item 17 m
~... -
I i
. Existing T.S. #
Description Proposed RTS Refueling Requirement j
Shutdown High Neutron Level Alarm 3.9.2 f
Table 3.9-2 Accident Monitoring Instrumentation Item 1 Pressurizer Level Table 3.3-6 Item 5 Item 2 Aux. Feedwater Flow Rate Table 3.3-6 i
Item 11 Item 3 Delete Item 4 PORV Position Indicator Table 3.3-6 Acoustic Flow Monitor Item 14 Item 5 PORY Block Valve Position Table 3.3-6 Indicator Item 13 Item 6 Safety Valve Position Indicator, Table 3.3-6 Acoustic Flow Monitor Item 14 3.10 Reactivity Control System 3.10.1.1 Shutdown Margin - Modes 1, 2 3.1.1.1 Applicability 3.1.1.1 Action 3.1.1.1 Surveillance la 3.1.1.1 Surveillance Ib 4.1.1.1.1.a Surveillance Ic 4.1.1.1.1.b Surveillance 1d 4.1.1.1.1.c Surveillance 2 4.1.1.1.2 3.10.1.2 Shutdown Margin - Mode'3 3.1.1.2 Applicability 3.1.1.2 Action 3.1.1.2 Surveillance (a) 4.1.1.2a Surveillance (b) 4.1.1.2b 3.10.1.3 Shutdown Margin - Modes 4, 5 3.1.1.3 Applicability 3.1.1.3 Action 3.1.1.3 Surveillance (a) 4.1.1.3.a Surveillance (b) 4.1.1.3.b 3.10.1.4 Shutdown Margin - three loop 3.1.1.4 Applicability 3.1.1.4 Action 3.1.1.4 Surveillance ((2)
Surveillance 1) 4.1.1.4.1 4.1.1.4.2 3.10.1.5 Moderator Temperature Coefficient 3.1.1.5 Applicability 3.1.1.5 Action 3.1.1.5 Surveillance (a),(b),(c) 4.1.1.5.a,b,c 3.10.1.6 Minimum Temp, for Criticality 3.1.1.6-Applicability 3.1.1.6 i
Action 3.1.1.6 Surveillance (a)(b) 4.1.1.6.a,b W
e-n-.
-,+- -
- - -. ~ -, -
~
a
i
. Existing T.S. #
Description Proposed RTS 3.10.2 Movable Control Assemblies 3.10.2.1 Bank Height 3.1.3.1 Applicability 3.1.3.1 Action 3.1.3.1 Surveillance (b)(b) 4.1.3.1.1 and 4.1.3.1.2 3.10.2.2 Positive Indication System-Operating 3.1.3.2 App 11 ability 3.1.3.2 Action 3.1.3.2 Surveillance Requirement 4.1.3.2 3.10.2.3 Positive Indication Systems-Shutdown 3.1.3.3 Applicability 3.1.3.3 Action 3.1.3.3 Surveillance 4.1.3.3 3.10.2.4 Rod Drop Time 3.1.3.4 Applicability 3.1.3.4 Action 3.1.3.4 Surveillance 4.1.3.4 3.10.2.5 Shutdown Insertion Limits 3.1.3.5 Applicability 3.1.3.5 Action 3.1.3.5 Surveillance Requirement 4.1.3.5 3.10.2.6 Control Group Insertion Limits -
3.1.3.6.1 Four Loops Applicability 3.1.3.6.1 Action 3.1.3.6.1 Surveillance 4.1.3.6.1 3.10.2.7 Control Group Insertion Limits -
3.1.3.6.2 Three Loo >s Applica sility 3.1.3.6.2 Action 3.1.3.6.2 Surveillance 4.1.3.6.2 3.11 Containment Administrative Tech. Spec.
3.11.A Leakage Limit 3.6.1.2.a 3.11.B.2 Containment Integrity with reactor 3.9.1 vessel head removed 3.11.C Internal Pressure 3.6.1.4 3.11.D.1 Air Recirculation System Performance 4.6.2.c Requirement 3.11.D.2 Air Recirculation System Cold Shutdown 3.6.2 Requirement 3.11 Containment 3.6.1.2.a 3.11A Leakage Limit (see 3.11A Admin.)
3.6.1.2.a 3.118 Containment Integrity 3.11.B.1 RCS above 300 psig. and 200*F 3.6.1.1 3.11.8.2 See Admin. 3.11.B.2 3.9.1 3.11.B.3 Positive Reactivity (See Admin. 3.11.c)
Changes 3.6.1.1 - 3.9.4 3.11.C Internal Pressure 3. 6.1. 4 -
, ~ -
,n.,
.--~ -
Existing T.S. #
Description Proposed RTS I
3.11.D See Admin. 3.11.D.1 and 3.11.D.2 4.6.2.c and 3.6.2 3.11.E Containment Spray System 3.11.F Containment Venting 3.11.F.1 Post-Accident Hydrogen Venting 3.6.1.7, Table 3.3-10 Item Ic 3.11.F.2 Purge Capability 3.6.1.7 3.11 G.
Containment Isolation Valve In FSAR 3.11.G.1 Restore Inoperable Valve 3.6.3.a 3.11.G.2 Isolate by use of automatic valve 3.6.3.b 1
3.11.G 3 Isolate by use of manual valve 3.6.3.c 3.11.G.4 Hot Standby 3.6.3.d l
3.11.H Trip Setpoint 3.3.2 3.12 Station Service Power 3.8 3.13 Refueling 3.13A Monitoring Radiation Levels 3.13B Honitoring Neutron Flux 3.9.2 3.13.C.1 Water Level in the Refueling Cavity 3.9.8.1, 3.9.10 3.13.C.2 RHR Pump & Heat Exchanger in Operation 3.9.8.1 3.13D Boron Concentration 3.9.1 3.13E Charging Pump 3.1.2.3 3.13F Yerification of Suberiticality 3.13G Director Communication 3.9.5 3.13H Handling of Spent Fuel Cask 3.131 Loading of Fuel for Offsite Lab Study No longer Applicable 3.14 Primary System Leakage 3.14.A.1 Unidentified Leakage 3.4.6.2.b 3.14.A.2 Identified Leakage 3.4.6.2.d 3.14.A.3 Combined Leakage 3.4.6.2.f 3.14.A.4 No Pressure Boundary Leakage 3.4.6.2.a 3.14.A.5 Steam Generator Tube Leakage 3.4.6.2.c
~
3.14.A.6 ECCS Valves Leakage 3.4.6.2.g 3.14.B.1 Action for Pressure Boundary Leakage 3.4.6.1 Action a 3.14.B.2 Action for Other Leakage 3.4.6.2 Action b 3.14.3 Action for SG Tube Leacage 3.4.6.2 Action c 3.15 Intentionally Left Blank 3,16 Intentionally Left Blank 3.17 Power Distribution Limits 3.17.1 Axial Offset -
3.17.1.1 Axial Offset - Four Loops 3.2.1.1 Applicability 3.2.1.1 l
Action 3.2.1.1 Surveillance 4.2.1.1.1 Surveillance 4.2.1.1.2 l
Surveillance 4.2.1.1.3 i
Surveillance 4.2.1.1.4 i
- See Section 3.2 of SER
i 11 -
Existing T.S. #
Description Proposed RTS 3.17.1.2 Axial Offset - three loops 3.2.1.2 Applicability 3.2.1.2 Action 3.2.1.2 Surveillance a 4.2.1.2.1 Surveillance b 4.2.1.2.2 Surveillance c 4.2.1.2.3 Surveillance d 4.2.1.2.4 3.17.2 Linear Heat Generator Rate 3.2.2.1 3.17.2.1 Four Loops Operating 3.2.2.1 Applicability 3.2.2.1 Action 3.2.2.1 Surveillance (1) 4.2.2.2.1 l
Surveillance (2) 4.2.2.2.2 3.17.2.2 Three Loops Operating 3.2.2.2 Applicability 3.2.2.2 Action 3.2.2.2 Surveillance (1) 4.2.2.2.1 Surveillance (2) 4.2.2.2.2 3.17.3 Nuclear Enthalpy Rise Hot Channel Factor 3.17.3.1 Four Loops Operating 3.2.3.1 Applicability 3.2.3.1 Action 3.2.3.1 Surveillance (1) 4.2.3.1.1 Surveillance (2) 4.2.3.1.2 3.17.3.2 Three Loops Operating 3.2.3.2 Applicability 3.2.3.2 Action 3.2.3.2 Surveillance (a) 4.2.3.2.1 Surveillance (b) 4.2.3.2.2 3.17.4 Quadrant Power Tilt Ratio 3.2.4 Applicability 3.2.4 Action 3.2.4 Surveillance a 4.2.4.1 Surveillance b 4.2.4.1 Surveillance c 4.2.4.1 3.17.5 DNB Parameters 3.2.5-Applicabiljty 3.2.5 Action 3.2.5 Surveillance a 4.2.5.1 Surveillance b 4.2.5.2 Surveillance c 4.2.5.3 3.18 Intentionally Left Blank 3.19 Snubbers 3.7.4 3.19.A Applicability 3.7.4 3.19.8 One inoperable 3.7.4
.-,-n-
-,, - ~, - -
g-,-
n
.., - - -...... -,, ~...
4
. Existing T.S. #
Description Proposed RTS 3.20 Intentionally Left Blank 3.21 Safety-Related Equipment Flood Protection 3.21.1 Operability Requirement 3.3.4 3.21.2 Condensate Return Pump Operability 3.3.4 3.21.3 Screenwell House & D.G. Room Operability 3.21.4 Actions for 3.21.1 and 3.21.2 3.3.4 Can't Be Met 3.22.A.1 Fire Water System / Operability 3.7.6.1 3.22.A.2 One Pump Inoperable 3.7.6.1.a (Action)*
3.22.A.3 Two Pumps Inoperable 3.7.6.1.b (Action)*
3.22.8.1 C0, System / Operability 3.7.6.3 3.22.8.2 Action 3.7.6.3.a(Action) 3.22.B.3 Action - Reportability 3.22.C.1 Halon System / Operability 3.7.6.4 3.22.C.2 Action 3.7.6.4.a(Action) 3.22.C.3 Action - Reportability 3.22.0.1 Fire Water Stations / Operability 3.7.6.5/3.7.6.6 3.22.D.2 Action 3.7.6.5.a/3.7.6.6.a 3.22.E.1 Fire Detection System / Operability 3.3.3.6 3.22.E.2.a Action 3.3.3.6.b 3.22.E.2.b Action - Reportability 3.22.F.1 Penetration Fire Barriers / Operability 3.7.7 3.22.F.2 Action 3.7.7.a 3.22.G.1 Spray and/or Sprinkler Systems 3.7.6.2 3.22.G.2 Action 3.7.6.2.a(Action) 3.22.G.3 Action - Reportability 3.22.H Flammable Liquids Controls 3.7.8 3.22.H.1 Action - Written Permission 3.7.8.a 3.22.H.2.
Action - Container 3.7.8.b 3.22.H.3 Action - Fire Watch 3.7.8.c Table 3.22-1 Fire Water Stations Table 3.7-4/3.7-5 Table 3.22-2 Fire Detection Instruments Table 3.3-8 l
3.23 Post-Accident Monitoring Instrumentation 3.3.3.5 Table 3.23-1 and Accident Monitoring Instrumentation Table 3.3-7 and Table 3.23-2 Table 4.3.6 Item 1 Containment Pressure 1
Item 2 RCS - Cold Leg Temp.
2 Item 3 R25 - Hot Leg Temp.
3 Item 4 RCS Pressure 4
i Item 5 Containment Water Level 15 Item 6 CET 17 Item 7 Main Stack Wide Range Noble Gas Monitor 18 Item 8 Containment Atmosphere High Range Radiation 19 Monitor i
i
- See Section 3.2 of SER
. \\
Existing T.S. #
Description Proposed RTS Item 9 Reactor Vessel Water Level 20 Item 10 RC$ Subcooling Maring Monitor 12 1
3.24
$)ecial Test Rxceptions 3.10.1 3.24.1 51utdown Margin 3.10.1 i
Applicability 3.10.1 Action 3.10.1 1
Surveillance (a) 4.10.1.1 Surveillance (b) 4.10.1.2 3.24.2 Physics Test 3.10.2 Applicability 3.10.2 Action 3.10.2 Surveillance 4.10.2.1 Surveillance 4.10.2.1 Surveillance 4.10.2.3 3.24.3 Position Indication System - Shutdown 3.10.3 Applicability 3.10.3 Action 3.10.3 Surveillance 4.10.3 4.1 Introduction to Surveillance Requirements 4.01, 4.02 I
4.2 t
Administrative Operational Safety Items Table 4.2-2 PORV's and Block Valves Demonstrated Operable Ittm 15 A
Demonstrated Operable 4.4.4.1 B
Block Valve Demonstrated 4.4.4.3 C
The Emergency Air and Power Supply Demonstrated Operable C.1 Transfer From Normal to Emergency Power 4.4.4.6 C.2 Operate through Complete Cycle 4.4.4.6 D
Demonstration of Minimum Pressure on 4.4.4.5 Emergency Air Supply 4.2 Operational Safety Items Table 4.2-1 Minimum frequencies for Testing, Calibrating
]
and/or Checking Instrument Channels 4
1 Nuclear Power Table 4.3-1 Item 2 2
Intermediate Range Table 4.3 1, Item 3 3
Source Range 4
Reactor Coolant Temperature Table 4.3-6 5
Reactor Coolant Flow Table 4.3-1, Item 7 3.2.5 6
Pressurizer Level Table 4.3-1. Item 6 t
7 Pressurizer Pressure Table 4.3-1, Item 5 l
8 Yariable Low Pressure Trip Setpoint Table 4.3-1, Item 4 I
C61culator
- See Section 3.2 of SER l
l Existing T.S. #
Description Proposed RTS 9
Rod Position Digital Voltmeter 3.1.3.2 l
10 Rod Position Counters 3.1.3.2 11 Steam Generator Level Table 4.3 1 12 Steam Generator flow Mismatch Table 4.3-2 13 Charging Flow i
14 Residual Heat Pump Flow 4.5.1.g(7) 15 Boric Acid Tank Level 4.1.2.$a 4.1.2.6.1.b.
Table 4.3-6 16 Refueling Water Storage Tank Level Table 4.3 6, Item 11 17 Volume Control Tank Level 4.4.6.1.c 18 Blank 19 Radiation Monitoring System Table 4.3-3 20 Boric Acid Control 21 Blank 22 Valve Temperature Interlocks 4.4.1.7.2 and 4.4.1.11.3 23 Pump-Valve Interlock 4.4.1.11.4 24 Reactor Coolant System OPS 4.4.9.3.4 25 Auxiliary Feedwater Flow Rate Table 4.3-6, Item 11 26 Blank 27 PORY Position Indication Table 4.3-6, (AcousticMonitor)
Item 14 28 PORY Block Valve Indication Table 4.3-6, Jtem 13 29 Safety Valve Position Indication Table 4.3-6, item 14 (AcousticMonitor) 4.2 Operational Safety Items Table 4.2-2 Minimum Equipment Check and Sampling Frequency 1
Reactor Coolant Sample
. Table 4.4-4, Item 1 2
Reactor Coolant Boron 3.1.1.2, 3.1.1.3 3
Refueling Water Storage Tank Water Sample 4.1.2.5a 4
Control Rods 3.1.3.4 5
Control Rods 3.1.3.1 6
Pressurizer Safety Valves 4.4.2.1 and 4.4.2.2 7
Main Safety 4.7.1.1 8
Main Steam Isolation Valves 4.7.1.5 9
Reactor Containment Trip Valves 4.6.3 10 Refueling System Interlocks 11 Boric Acid Pumps 4.1.2.1.b, 4.1.2.2.b 12 RCS Overpressure Protection System 4.4.9.3.3 Isolation Yalve Interlocks and Alarms r
13 RCS Overpressure Protection Isolation 4.4.9.3.1 l
Valves i
14 RCSVent(s) 4.4.9.3.2
'I
- See Section 3.2 of SER
.. ~
. i i
Existing T.S. #
Description proposed RTS 4.3 Core Cooling Systems Periodic Testing Administrative Technical Specification 4.3.B.1.4 Once in 31 days - verify valves are 3.6.1SR(c.1) in correct position 4.3.G Visual inspection for no loose debris 3.6.1SR(d) 4.3 Core Cooling Systems Periodic Testing 4.3.A.1 Once per 18 months - s/d - automatic 3.6.1SR(f) operation of the ECCS 4.3.A.2 Verification of starting of 3.12.1.1.2SR(f)
D.Gs and pumps 4.3.A.3 Control board indications 3.12 4.3.A.4 Venting prerequisite for test 3.3.4.2 (Existin )
4.3.B.1 Monthly pump test on recirculation 3.6.1SR(c3 4.3.B.2 Monthly testing of charging and 3.6.1 SR(c4 metering purps 4.3.B.3 Cycling of safety injection and 3.6.1 SR(f1) core deluge valves 4.3.0.4 Exercise two valves 3.6.1SR(c2) 4.3.C Testing requirement on remaining pump 3.6.1 Action 4.3.0 Motor-operated containment spray 4.0.5 water valve 4.3.E Demonstrate pumps ino)erable Periodic 3.6.2SR(b) 4.3.T leak testing of each ECCS check valve 3.6.1 SR(h) shown in Table 4.3-1 (6 valves) 4.3.G Correct position of ECCS throttle valve 3.6.1 SR 4.3.H Flow Balance Test 3.6.1 SR 4.4 Containment Testing Administrative Tech Specs I.B.1 Acceptance Criteria 3.6.1.2a IV.A.4 Demonstrated condition for filteration 4.6.2.e unit IV.B.1 Acceptable filter efficiencies l
IV.C.3 Corrective Actions for Unusual Conditions IV.D Test frequency IV.D.1 18 month test frequency 4.6.2.c IV.D.2 Visual Inspection IV.D.3 Damper test IV.D.4 Charcoal Spray Valve 1Y D.S Halogenated Hydrocarbon Testing 4.6.2.g
.!Y.D 6 Cold DOP Test 4.6.2.f IV.D.7 15-Minutes Operational Requirements 4.6.2.a.1
- Done procedurally in accordance with Regulatory Guide 1.52 Rev. 2 j
i
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i 4
l Existing T.S. #
Description ProposM RTS i
4.4 Containment Testing l
i 4.4.1.A Integrated Leakage Test 4.6.1.2, 4.6.1.6.1 4.4.1.B.1 See Admin. Spec.
3.6.1.2.a i
i 4.4.1.B.2 Max. Allowable Reduced Pressure Test N/A i
In8)vidual Leak Detection Test (P
Leakage Rate l
4.4.!!.A 1
4.6.1.2. 4.6.1.3, 4.6.1.7.2 i
4.4.!!.B Acceptance Criterion 3.6.1.2.b l
4.4.!!.C Corrective Action 3.6.1.2 4.4.!!.D.1 Equipment hatch and fuel transfer
- 4. 6.1.1. c,
Tube 4.6.1.2 4.4.!!.D.2 Isolation Valves 4.6.1.2.d 4.4.ll.D.3 Personnel Air-lock Assembly 4.6.1.3, 4.6.1.1.b, 4.6.1.2.d 4.4.111 Recirculation System 4.4.!!!.A Recirculation System Test 4.4.6.2.1.g 4.4.6.2.1.f 4.4.!!!.B Acceptance Criteria 3.4.6.2.e 4.4.!!!.C Corrective Actions 3.4.6.2 Action 4.4.!!!D Test Frequency 4.4.6.2.1 9 Admin. 4.4.!!!.B Acce)tance Criteria 3.4.6.2 Action d 4.4.IV Air Filtration System 4.4.IV.A Tests 4.4 IV.A.1 Measurement of lodine Removal Efficiency 4.6.2.C.1 4.4.IV.A.2 In-place Freon 112 Test 4.6.2.g and 4.6.2.C.1 4.4 IV.A.3 Visual Inspection of Filter Banks 4.6.2.a.3 4.4.IV.A.4 Pressure drop across charcoal filter 4.6.2.a.3 4.4.IV.A.5 Damper Testing 4.6.2.a.4 and 4.6.2.e.2 l
4.4.IV.B Acceptance Criteria 4.6.2.c 4.4.IV.B.1 See Admin. 4.4.IV.B.1 4.4.IV.B.2 Acceptable Charcoal filter Efficiencies 4.6.2.g 4.4.IV.C Corrective Action 3.6.2 4.4.IV.C.1 Replacement of Charcoal 3.6.2 4.4.IV.C.2 Location of Leakage Paths 3.6.2 4.4.IV.C.3 See Admin. Tech. Spec.
4.4.!Y.D See Admin. Tech. Spec.
4.4.IV.E Sunenary of Technical Report 4.8 AFW system 3/4.7.1.2 4.8.1 AFW operability every 31 days 4.7.1.2.1 l
4.8.1.a Discharge pressure 4.7.1.2.1.a 4.8.1.b S/G level instrumentation Tables 4.3-2/4.2-1 l
4.8.1.c Verify correct valve position 4.7.1.2.1.b 4.8.2 DWST/PWST operability every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 4.7.1.3.1 l
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. Existing T.S. #
Description Proposed RTS 4.8.3 AFW operability every refueling 4.7.1.2.2 4.8.3.a Pump capability 4.7.1.2.2.a 4.8.3.b Verify correct valve position upon 4.7.1.2.2.b AFW actuation test signal 4.8.3.c Verify AFW pump starts upon AFW 4.7.1.2.2.c actuation test signal 4.9 MSIVs 3.7.1.5 4.10 Inservice Inspection and Reactor Vessel Surveillance 4.10A ISI of Class 1, 2, 3 Component 4.0.5a, 4.0.10 4.10B ISI of Class 1, 2, 3 Pumps and Valves 4.0.5a. 4.4.10 4.100 RCP Flywheel 4.4.10 4.10D Reactor Vessel Surveillance Capsule Table 4.4-5 4.10.1 In-service Inspection of Steam Generator Tubes 4.10.1A SG Sample Selection and Inspection 4.4.5.1 4.10.18 SG Tube Sample Selection and Inspection 4.4.5.2 4.10.1 B.1 Areas to Be Inspected 4.4.5.2.a 4.10.1.B.2 First Sample 4.4.5.1.b 4.10.1.B.3 Second and Third Sample 4.4.5.2.c 4.10.1.C Inspection Frequencies 4.4.5.3 4.10.1.D Acceptance Criteria 4.4.5.4 4.10.1.E Reports 4.4.5.5 Table 4.10.1-1 Minimum Nunter of SG Tube Inspected Table 4.4-1 Table 4.10.1-2 SG Tube Inspection Table 4.4-2 4.11 Deleted 4.0.6 4.12 High Energy Piping System 4.0.6 4.12A Augmented Inservice Inspection Program 4.0.6 4.12.A.1 First Ten-Year Inspection Program 4.0.6 4.12.A.2 Successive Inservice Inspection Program 4.0.6 i
4.12.A.3 Repairs, Reexamination and Test 4.0.6 4.13 Snubbees 4.7.4 4.13.A Visual inspection schedule 4.7.4.a 4.13.B Visua? inspection criteria 4.7.4.b 4.13.C functional tests 4.7.4.c 4.13.D Hydraulic snubbers test criteria 4.7.4.d 4.13.E Mechanical snubbers test criteria 4.7.4.e 4.13 F Snubber service life monitoring 4.7.4.f 4.14 Flood Protection Annunciators 4.14A Test 4.3.4 l
4.14B Acceptance Criteria 4.3.4 4.14C Corrective Action 3.3.4 4.14D Test Frequency 4.3.4 l
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Existing T.S. #
Descriotion Proposed RTS 4.15.A.1 Fire Water System Operability 4.7.6.1.1 4.15 A.I.a Pump Operability 4.7.6.1.1.a 4.15.A.1.b Valve Operability 4.7.6.1.1.b 4.15.A.1.c Valve Operability 4.7.6.1.1.c 4.15.A.I.d.1 Auto Actuation 4.7 6.1.1.d 4.15.A.1.d.2 Pump Flow / Pressure 4.7.6.1.1.d i,
4.15.A.1.d.3 Valve Operability 4.7.6.1.1.d 4.15.A.1.e Flow Test 4.7.6.1.1.e 4.15.8.1 C0 System Operability 4.7.6.3 4.15 B.I.a Cy$1nderWeight 4.7.6.3.b.1 4.15.B.1.b.1 Component Operability 4.7.6.3.b.1 4.15.6.1.b.2 Flow Test 4.7.6.3 b.2 4.15.C.1 Halon System Operability 4.7.6.4 4.15.C.I.a Cylinder Weight / Pressure 4.7.6.4.a 4.15.C.1.b.1 Component Operability 4.7.6.4.b.1 4.15.C.1.b.2 Visual Inspection 4.7.6.4.b.2 1
4.15.0.1 Fire Hose Station Operability 4.7.6.5/4.7.6.6 4.15.D 1.a Visual Inspection 4.7.6.5.a/4.7.6.6.a 4.15.D.I.b Removal / Inspection 4.7.6.5.b/4.7.6.6.c 4.15.0.1.c Flushing 4.7.6.5.c/4.7.6.6.c 4.15.D.1.d Valve Operability 4.7.6.5.c/4.7.6.6.c 4.15. D. I. e Hose Hydrostatic Test 4.7.6.5.b.1/4.7.6.6.c.1 4.15.E.1 Channel Functional Test 4.3.3.b.1 4.15.E.2 Circuit Supervision 4.3.3.6.2 4.15 F.1 Penetration Fire Barrier Operability 4.7.7.1 4.15.F.1.a Visual Inspection 4.7.7.1.a 4.15.F.1.b Post Repair inspection Done by Procedure 4.15 G.1 Spray and/or Sprinkle Operability 4.7.6.2 4.15.G.I.a Valve Operability 4.7.6.2.b 4.15.G.1.b.1 Functional Test 4.7.6.2.c.1 4.15.G.I.b.2 Visual Inspection - Headers 4.7.6.2.c.2 4.15.G.1.b.3 Visual Inspection - Nozzles 4.7.6.2.c.3 4.15.G.I.c Flow Test 4.7.6.2.d 5.0 Design Feactures 5.1 Introduction 5.2 Site Description 5.1.1 5.3.A Reactor Core 5.3.1 5.3.B Reactor Coolant System 5.4.1 5.4 Containment 5.2.1 6.0 Administrative Controls 6.1 Responsibility 6.1 6.2 Organization 6.2 6.2.1 Offsite Organization 6.2.1 6.2.1 Facility Staff 6.2.2 6.3 Facility Staff Qualification 6.3 6.3.1 Facility Staff Qualification 6.3.1 6.3.1.1 Health Physics Supervisor 6.3.1.1 6.3.1.2 STA 6.3.1.2
i Existing T.S. #
Description
. Proposed RTS t
6.4 Training 6.4 t
6.4.1 Retraining and Replacement 6.4.1 Training Program j
6.4.2 Fire Brigade Training Program 6.4.2 6.5 Review and Audit 6.5 l
6.5.
PORC 6.5.1 l
6.5.1.1 PORC Function 6.5.1.1 6.5.1.2 Composition 6.5.1.2 6.5.1.3 Alternate 6.5.1.3 6.5.1.4 Meeting frequency 6.5.1.4 6.5.1.5 Quorum 6.5.1.5 6.5.1.6 Responsibilities 6.5.1.6 6.5.1.6a Responsibilities 6.5.1.6a 6.5.1.6b Responsibilities 6.5.1.6b 6.5.1.6c Responsibilities 6.5.1.6c 6.5.1.6d Responsibilities 6.5.1.6d 6.5.1.6e Responsibilities 6.5.1.6e 6.5.1.6f Responsibilities 6.5.1.6g 6.5.1.6g Responsibilities 6.5.1.6h' 6.5.1.6h Responsibilities 6.5.1.61 6.5.1.6i Responsibilities 6.5.1.6j 6.5.1.6j Res sensibilities 6.5.1.7b 6.5.1.7a Autaority 6.5.1.7a 6.5.1.7b Authority 6.5.1.7b 6.5.1.8 Records 6.5.1.8 6.5.2 NRB 6.5.2 6.5.2.1 Qualification 6.5.2.1 6.5.2.2 Composition 6.5.2.2 6.5.2.3 Consultants 6.5.2.3 6.5.2.4 Meeting Frequency 6.5.2.4 6.5.2.5 Quorum 6.5.2.5 6.5.2.6 Review 6.5.2.6 6.5.2.7 Audits 6.5.2.7 6.5.2.8 Authority 6.5.2.8 6.5.2.9 Records 6.5.2.9 6.6 Reportable Event Action 6.6 6.7 Safety Limit Violation 6.7 6.8.1 Written Procedures 6.8.1 l
6.8.2 Approval of Procedures 6.8.2 i
6.8.3 Temp. Changes to Procedures 6.8.3 6.8.4 Admin.
Written Procedures 6.8.1.d, e, f 6.8.5 Admin.
Written Procedures 6.8.4 6.8.6 Admin.
Written Requirements 6.8.5 6.9 Reporting Requirements 6.9.1 6.9.la Start-up Reports 6.9.1.1, 6.9.1.2, and 6.9.1.3 i
)
)
. Existing T.S. #
Description Proposed RTS i
6.9.1.b SG Tube Inspection 4.4.5.5.b 6.9.1.c Occusational Exposure Report 6.9.1.5.a 6.9.1.d Mont11y Operating Report 6.9.1.8 6.9.1.e 10 CFR 50.59b 6.9.1.f Admin.
Annual Radiological 6.9.1.6 i
Environmental Radioactive Report 6.9.1.g Admin.
Semiannual Radioactive Effluent 6.9.1.7 i
Release Report 6.9.2 Special Reports 6.9.2 6.9.2a ISI Results 4.0.5 6.9.2b Primary Containment Results Required By Appendix J 6.9.2c Reactor Yessel Material Surveillance 4.4.9.1.2 Specification Examination 6.9.2.d SG Tube Report 4.4.5.5.a 6.9.2.e Post-Accident Operability Table 3.3-7 6.9.2.f Fire Protection System Operability 6.9.2.g RCS Vent 3.4.11 6.9.2.h Radiological Effluent Reports 3.11.2.2 3.11.1.2 3.11.2.3 3.11.3 6.10.1 Record Retention 6.10.2 6.10.2 Record Retention 6.10.3 6.11 Radiation Protection Program 6.11.1 6.12 Deleted 6.13 High Radiation Area 6.1.2 6.14 Deleted 6.15 System Integrity 6.15 6.16 lodine Monitoring 6.16 6.17 REMODCM 6.13 6.18 Radioactive Waste Treatment Systems 6.14 6.19 PASS / Sampling and Analysis Plant 6.16 Effluents 7/8 Radioactive Effluents 3.11 7/8.1.1 Liquid Effluents 3.11.1.1 7.1.1.1 Concentration 3.11.1.1 Applicablity 3.11.1.1 Action 3.11.1.1" 8.1.1.1.1 Sampled and Analyzed 4.11.1.1.1 8.1.1.1.2 Assure Limits 4.11.1.1.2 7.1.1.2 Dose-Liquid 3.11.1.2 Applicability 3.11.1.2 Action 3.11.1.2 8.1.1.2.1 Determination 4.11.1.2.1 8.1.1.2.2 Confirmation 4.11.1.2.2 7.1.2.1 Dose Rate - Gas 3.11.2.1 Applicability 3.11.2.1 Action 3.11.2.1
t l
. Existing T.S. $
Description Proposed RTS 8.1.2.1.1 Determination 4.11.1.1.1 8.1.2.1.2 Control of Release Rates 4.11.1.1.2 8.1.2.1.3 Release Rate of I-131, etc.
4.11.1.1.3 l
7.1.2.2 Dose-Noble Gas 3.11.2.2 Applicability 3.11.2.2 Action 3.11.2.2 8.1.2.2.1 Cum. Dose 4.11.2.2.1 8.1.2.2.1 Confirmation 4.11.2.2.2 7.1.2.3 Dose-lodine 3.11.2.3 Applicability 3.11.2.3 l
Action 3.11.2.3 i
8.1.2.3.1 Cum. Dose Contributions 4.11.2.3.1 l
8.1.2.3.2 Confirmation 4.11.2.3.2 l
7.1.3 Total Dose 3.11.3 Applicability 3.11.3 Action 3.11.3 8.1.3 Determination 4.11.3 7/8.2 Instrumentation 7.2.1.1 Radioactive Liquid Effluent Instrumentation 3.3.3.7 Applicability 3.3.3.7 Action 3.3.3.7 8.2.1.1 Demonstrate Operable 4.3.3.7.1 7.2.2.1 Radioactive Gaseous Effluent Monitoring 3.3.3.8 Instrumentation Applicability 3.3.3.8 Action 3.3.3.8 8.2.2.1 Demonstrated Operable 4.3.3.8.1 Table 7.2-1 Radioactive Liquid Monitoring Table 3.3-9 Instrumentation Table 8.2-1 Radioactive Liquid Monitoring Table 4.3-7 Surveillance Table 7.2-2 Radioactive Gaseous Monitoring Table 3.3-10 Instrumentation l
Table 8.2-2 Radioactive Gaseous Monitoring Table 4.3-8 Surveillance g
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PART 2 0F SAFETY EVALUATION i
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RELATED TO AMENDMENT NO.125 INTRODUCTION By [[letter::B13324, Application for Amend to License DPR-61,revising Tech Spec 3/4.5.1, ECCS - ECCS Subsystem - Tavg Greater than or Equal to 350 F to Reflect Component Configurations Following 1989 Refueling Outage Mod to ECCS to Resolve Single Failure|letter dated August 2,1989]], Connecticut Yankee Atomic Power Company (CYAPC0/
licensee)requestedchangestoTechnicalSpecification(TS)3/4.5.1," Emergency Core Cooling System - ECCS Subsystem - Tavg Greater Than or Equal to 350"F to reflect modifications to be implemented by the end of the 1989 refueling outage.
These modifications will resolve single failure concerns identified on three separate occasions.
In addition, TS Section 3/4.5.2, " Emergency Core Cooling System - ECCS Subsystem - Tavg Less Than 350"F," Section 5.3 " Emergency Core Cooling System - Refueling Water Storage Tank," and Section 5.4, "F.mergency Core Cooling System - pH Control System" have been renumbered to be consistent withtheWestinghouseStandardTechnicalSpecifications(WSTS) format.
D_lSCUS$10N This TS will address three single failure vulnerabilities that were discovered by CYAPCO.
Thesesinglefailurevulnerabilitiesare1)SmallBreakLOCA, 2)MediumBreakLOCAand3)ChargingPumpFlowPaths.
Small Break LOCA On March conjunction with the Integrated Safety Assessment Program 31,1986, CYAPC0 su ISAP) for the Haddam Neck Plant,lant System (RCS) for which safety injection flow in the 11gh which identified a small range of break sizes in one loos of the Reactor Coo pressure recirculation mode may be insufficient to provide adequate core cooling.
To respond to these Small Break (SB) LOCAs, CYAPC0 took temporary measures which were approved by the NRC.
The emergency operating procedures were revised to provide an alternate flow path utilizing the High Pressure Safety Injection (HPSI) pumps for core cooling during the high pressure recirculation mode. The use of this flow path required realignment of two valves which did not satisfy the single failtre criterion.
Therefore, CYAPCO requested an exemption from the single failure criterion for these valves, pending implementation of the permanent modifications.
On April 28, 1986, the NRC granted the requested exemption anc requested that CYAPC0 provide by September 1986, a description of the long term resolution and a schedule for completion of any modifications.
By letters dated September 30, 1986 and April 1, 1987, CYAPC0 submitted a description of the proposed modifications and requested a one-cycle extension of the exemption because some of-the modifi-cations could not be completed until the end of the Cycle 15 outage.
On Septenter 2,1987, the NRC granted an extension until the end of the Cycle 15 outage.
This TS change will incorporate the new valves necessary for HPSI recirculation, j
i
, Medium Break LOCA i
While analyzing the design for the small break LOCA modifications, a medium size break in the core deluge system was identified which would not be sufficiently mitigated during sump recirculation. Procedures were developed, d
a flow control valve was repositioned and the TS were changed to provide a temporary resolution to this problem.
This TS change will incorporate a new valve necessary for the resolution of this issue.
Charging Pump Flow Paths During routine plant inservice inspection, CH-MOV-257, volume control tank (VCT) outlet valve failed to operate.
As part of the root cause analysis and sub-sequent evaluation, CYAPC0 identified two single failure vulnerabilities (failure of CH-MOV-257 or BA-MOV-373, Suction line from Reactor Water Storage Tank (RWST)) which could impact charging system performance. A temporary resolution consisting of automatically tripping both charging pumps cn a safety injectionsignal(SIAS)wasimplemented.
CYAPC0 will resolve these single failure vulnerabilities by adding redundant valves for CH-MOV-257 and BA-MOV-373.
These valves will be included in the TS.
EVALUATION A)
Small Break LOCA By letter dated September 7,1987, (Attachment 1) the staff approved the permanent ECCS modifications necessary to resolve the small break LOCA problem.
Details of the proposed modification can be found in the attachment.
This TS change will implement the modifications previously approved, to provide HPSI recirculation capability at the Haddam Neck site.
During the 1987 outage an eight inch cross-tie connection between the RHR pump discharge and the HPSI pump suction was added.
Inadditionmotoroperatedvalves(MOVs)SI-MOV-854A, 8548, 901, 902 and 873 were installed.
During the 1989 refueling outage the remaining modifications necessary for implementation of the HPSI recirculation will be completed.
These modifications will include removing valves SI-V-857A and B and SI-FCV-875 and the installation of valves:
SI-MOV-903, SI-MOV-904, SI-V-919, SI-Y-920, SI-CV-921, SI-CV-922, SI-CV-923, SI-CV-924 SI-Y-925, SI-V-926 SI-V-927, SI-V-928, SI-V-929, SI-Y-930 and SI-V-931.
The implementation of the HPSI recirculation will require the following specific TS changes:
1)
Valve SI-FCV-875, HPSI miniflow line, has been deleted.
This valve has been physically removed from the HPSI miniflow lir.e.
1
. b)
The asterisk and footnote for valve RH-MOV 874, RHR recirculation line, has been deleted.
The note required this valve to be cycled every 31 days.
This requirement was part of the compensatory measures taken because the temporary HPSI recirculation path was not single failure proof.
To insure reliability of the path CYAPC0 had agreed to increased surveillance on this valve and SI-MOV-24, RWST I
line isolation valve.
With the completion of the HPSI recirculation modifications this testing is no longer needed.
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c)
The positions for valves SI-MOV-854A and B in the TS have been changed to "Open-Manual Operator is Locked" from " Locked Open." These valves have been installed since the 1987 outage, however, power is not available for them until the completion of the new switchgear room.
Since the HPSI recirculation modificaticns were incomplete, these j
valves were locked open as required for the current plant safety i
configuration.
With the completion of the HPSI recirculation j
modifications and the new switchgear room, these valves will be powered with the manual operator locked.
During HPSI recirculation these valves will need to be closed to provide redundant isolation of the RWST.
d)
The positions for valves SI-MOV-901 and 902 RHR/HPSI crosstie, in the TS have been changed to " Closed.
ManualOperatorisLocked"from i
" Locked Closed." The situation with these valves is exactly the same as with the SI-MOV-854A and B valves except the required sosition for these valves was locked closed.
For HPSI recirculation tSese valves w'111 be opened to provide suction for the HPSI pumps from the discharge J
of the RHR pumps, e)
Valves SI-MOV-903 and 904 HPSI miniflow, have been added to the TS and are required to be open with the manual operator locked.
These valves were added to provide remote redundant isolation valves in the HPSI pump minimum flow line.
These valves replaced SI-FCV-875, i
HPSI miniflow line valve, SI-V-857A and B, manual HPSI pump minimum l
flow line valves.
During the recirculation phase these valves would l
be closed to isolate the RWST and prevent backfilling of the RWST with containnent sump water.
2)
Surveillance c.2, which currently requires valves SI-MOV-24 and RH-MOV-874 to be cycled every 31 days, will be deleted.
As noted in 1,b this surveillance was part of the compensatory measures taken because of the single failure vulnerability of the temporary HPSI recirculation path.
With the completion of the permanent HPSI recirculation path this increased surveillance is no longer necessary.
In addition, c.3 and c.4 were renumbered because c.2 has been deleted.
b)
Surveillance c.4 is being added to require monthly verification that containment sump valve RH-MOV-22 can be cycled manually from the control room and valve RH-V-808A can be manually cycled locally.
To assure the reliability of the recirculation path, CYAPCO has increased the surveillance interval of these valves to monthly from 18 months.
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. j l
3)
Surveillance f.2 is being revised to require all remote manual valves, which are required to change position during a LOCA, to be cycled once per
{
18 months.
These valves are also in the inservice testing (!$T) program.
i The original TS included only valves RH-MOV-22 and RH-Y-808A.
As noted in j
2.b these valves are now cycled monthly and no longer included in this i
surveillance.
CYAPC0 proposed this additional surveillance in the TS to highlight the importance of these valves.
4)
TS Table 4.5.2 This table lists all the valves to be tested by TS Section 4.5.1.f B)
Medium Break LOCA (Core Deluge Line Break) l 1)
TS Section 4.5.1.a The position of valve RH-FCV-796 in the TS has been changed to " Blocked open position" from " Blocked in throttled position." As part of the l
temporary resolution to the core deluge line break, CYAPC0 determined that RH-FCV-796 had to be throttled to prevent RHR pump runout.
SI-MOV-873, a remote, redundant valve to isolate the core deluge line from the ECCS in the event of a core deluge line break has been added.
Therefore, RH-FCV-796 no longer needs to be throttled and can be returned to the full open position.
2)
TS Section 4.5.1.b The position of valve SI-MOV-873 in the TS has been changed to " Locked Open.
Operator circuit breaker locked open" from " Valve is locked open and electrically disconnected." As part of the permanent solution to the core deluge line break valve SI-Y-873 was replaced with SI-MOV 873.
Since all the modifications necessary to resolve the LOCA problems were incomplete, SI-MOV-873 was locked open and electrically disconnected, which was j
consistent with the plant's current safety configuration. With the completion of the ECCS modifications during the 1989 refueling outage SI-MOV-873 will be provided with electrical power and an open breaker which will allow electrical energization to permit remote closure while still preventing inadvertent valve closure.
As noted before this valve provides 1
remote, redundant isolation capability for the core deluge line break.
1 3)
TS Section 4.5.1.1 Yalve RH FCV-796 is being deleted from the list of throttled valves.
As i
noted earlier RH-FCV-796 is no longer throttled during normal operation i
and is blocked in the full open position.
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- l 4)
TS Section 4.5.1.j Surveillance j.2, RHR pumps discharge flow balance test, is being deleted.
f This test was necessary to verify that valve RH-FCV-796 was throttled in the correct position as part of the temporary solution to the core deluge i
line break.
This test is no longer necessary since valve RH-FCV-796 is no t
longer throttled.
In addition this s changedtoincorporatesurveillancej.urveillanceisbeingeditorially i into surveillance j.
C)
Charging Pumps Flow Pat _hs f
1)
TS Table 4.5-1
~
This table has been revised tt; include valves BA-MOV-32, CH-MOV-257B and CH-50V 242B and their safety injection positions.
Valves CH-MOV-2578 and CH-50V-2428 were added to provide redundant isolation of the VCT from the charging pumps.
In addition valve BA-MOV-32, charging pump suction i
from the RWST, will be modified to receive an automatic open signal on a i
safety injection actuation signal (SIAS) and have a faster stro ce time.
This will assure an adequate suction supply to the charging pumps on a SIAS.
These modifications will allow the current charging pumps trip on
$1AS to be removed, as this trip was the temporary solution to the charging pump single failure vulnerabilities.
2)
TS Section 4.5.1 Bases The Bases section is being revised to reflect plant modifications and the associated proposed TS changes.
The staff has determined that all of the above TS changes are consistent with our Safety Evaluation dated September 2, 1987 relating to the ECCS modifications.
Therefore, the staff concludes that the proposed TS changes are acceptable.
ENVIRONMENTAL CONSIDERATION This amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements.
We have determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The staff has previously published a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding.
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuantto10CFR51.22(b),noenvironmentalim>actstatementorenvironmental assessment need be prepared in connection with tie issuance of the amendment.
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6.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be 4
l conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: Alan B. Wang
Attachment:
NRC letter dated 9/2/87 I
l
_PART 3 0F SAFETY EVALUATION RELATED TO AMENDMENT NO.125
1.0 INTRODUCTION
By letter dated July 31, 1989, Connecticut Yankee Atomic Power company (the licensee) requested approval of an amendment to the Haddam Neck Plant Technical Specifications.
The proposed changes reflect the installation of additional fire protection features associated with the licensee's efforts to conform with the requirements of Appendix R to 10 CFR 50, 2.0 DISCUSSION The Technical Specification Amendment includes the following changes:
1.
The inclusion of additional fire detection instruments and an increase in the minimum number of required operable fire detectors in several fire zones; 2.
The addition of new fire suppression systems in the new Switchgear Building; 3.
The installation of new fire hose stations in the new Switchgear Building; and 4.
Editorial changes to reflect reconfigured fire detector zones.
3.0 EVALUATION The staff initially, had several concerns with the licensee's proposed amendment.
The first was assurance that all of the fire protection features which were installed in conjunction with the licensee's efforts to conform with Appendix R to 10 CFR Part 50 would be reflected in the proposed Technical Specification changes.
Based on its review of the relevant design criteria documents, the staff finds that the proposed amendment is comprehensive in this regard.
The second staff concern was that the numbers of additional fire detectors and fire hose stations identified in the amendment request reflected an adequate design.
Based on its review of the system design details, the staff finds that these fire protection features conform with the relevant criteria contained in AppendixAtoBranchTechnicalposition(BTP)APCSB9.5-1.
Based on its review, the staff concludes that the licensee's proposed fire protection Technical Specification changes satisfy the guidelines of Appendix A to BTP APCSB 9.5-1 and are, therefore, acceptable.
2
4.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact have been prepared and published in the Federal Register on February 23,1990(55FR6563).
Accordingly, based upon the environmental assessment, we have determined that the issuance of the amendment will not have a significant effect on the quality of the human environment.
5.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1)there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conductedincompliancewiththeCommission'sregulations,and(3)theissuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
D. Kubicki
(
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PART 4 0F SAFETY EVALUATION RELATED TO AMENDMENT NO. 125
1.0 INTRODUCTION
f By letter dated July 28, 1989, supplemented September 29, 1989, Connecticut Yankee Atomic Power Company (CYAPC0/ licensee) proposed changes to the Technical i
Specifications (TS) for the Haddam Neck Plant.
The September 29, 1989 letter provided several pages of the TSs that were inadvertently not included in the t
July 28,1989 submittal.
These additional TS pages were within the scope of the original notice and do not affect the staff's determination in the original notice. The proposed changes would modify specifications having i
cycle-specific parameter limits by replacing the values of those limits with references to the Technical Report Supporting Cycle Operation (TRSCO) for the r
values of those limits.
The proposed changes also include the addition of TRSCO to the Definitions section and to the reporting requirements of the Aministrative Controls section of TS.
Guidance on the proposed changes was developed by NRC on the basis of the review of a lead-plant proposal submitted on the Oconee plent docket that was endorsed by the Babcock and Wilcox Owners G roup.
This guidance was provided to all power reactor licensees and appli-cants by Generic Letter 88-16, dated October 4 1988.
In addition made changes to IS Section 3/4.2.1.1, " Axial Offset-4 Loops," 3/4.E.CYAPCO has 1.2, " Axial Offset-3 Loops," 3/4.2.5, "DNB Related Parameters," 3.1.3.5, " Shutdown Rod Insertion Limit " 3.4.1.4.1, " Cold Shutdown Loops Filled" 3.4.9.1, " Pressure /
Temp. Limits-RCS " 3.4.9.3, "LTOP," Figures 3.4-3, 3.4-4 and 3.4 5 and Bases 3/4.23 and 3/4.4.1.
These changes were basically clarification or administra-tive changes.
EVALUATION The licensee's proposed changes to the TS are in accordance with the guidance-provided by Generic Letter 88-16 and are addressed below.
(1) The Definition section of the TS will be modified to include a definition of the TRSCO that requires cycle / reload-specific parameter limits to be l
l established on a unit-specific basis in accordance with an NRC-approved methodology that maintains the limits of the safety analysis.
The definition notes that plant operation within these limits is addressed by individual specifications.
(2) The following specifications will be revised to replace the values of cycle-specific parameter limits with a reference to the TRSCO that provides these limits.
i
.- ~.
b i
2-TS Section Title 4.1.1.1.1 15h"uTdown Margin - 4 Loops 4.1.1.4.1 Shutdown Margin - 3 Loops i
3.1.1.5 Mod. Temp. Coeff.
l 3.1.3.1 Moveable Cont. Assemblies l
3.1.3.6.1 Cont. Group Ins. Limit 4 Loops 3.1.3.6.2 Cont. Group Ins. Limit 3 Loops 3/4.2.1.1 Axial Offset i
3/4.2.1.2 Axial Offset 3/4.2.2.1 LHGR - 4 Loops 3/4.2.2.2 LgGR-3 Loops 3/4.2.3.1 F
' 4 LOOPS RH 3/4.2.3.2 F
- 3 Loops g
(3) Specification 6.9.1.9 " Technical Report Supporting Cycle Operation " will be added to the reporting requirements of the Administrative Controls section of the TS.
This specification requires that the TRSCO be submitted, upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.,
The report provides the values of cycle-specific parameter limits that are applicable for the current fuel cycle.
Furthermore, this specification requires that the i
values of these limits be established using the NRC-approved methodology in the references provided below and are consistent with a11' applicable limits of the safety analysis.
Finally, the specification requires that all changes in cycle-specific parameter limits be documented in the TRSCO before each reload cycle or remaining part of a reload cycle and submitted upon issuance to NRC, prior to operation with the new parameter limits.
F. M. Akstulewicz to E. J. Mroczka, " Review of NUSCO Topical Report I
a.
i on Physics Methodology for PWR Reload Design (NUSCO 152),"
l August 3, 1987.
b.
A. B. Wang to E. J. Mroczka, " Safety Evaluation for Northeast Utilities Topical Report 140-1, NUSCO Thermal Hydraulic Qualification, Volume I (RETRAN) " July 26, 1988.
c.
F. M. Akstulewicz to J. F. Opeka, "NUSCO Thermal Hydraulic Model Qualification, Volume II (VIPRE), Topical Report NUSCO 140-2,"
October 16, 1986.
d.
A. B. Wang to E. J. Mroczka, " Safety Evaluation of Northeast Utilities Topical Report 151, Haddam Neck Non-LOCA Transient Analysis," October 18, 1988.
1 I
e.
Supplement to the Safety Evaluation by the Directorate of Licensing, U.S. Atomic Energy Commission Docket No. 50-213, Connecticut Yankee Atomic Power Company, Haddam Neck Plant, December 27, 1974.
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i
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(4) The following Figures were deleted and the information provided in the TRSCO:
Fioure Title i
3.1-1 Rod Ins. Limit vs. Power Level
)
3.1-2 Rod Ins. Limit vs. Power Level 3.2-la Power Level vs. Axial Offset 3.2-lb Power Level vs. Axial Offset 3.2-2a Power Level vs. Axial Offset 3.2-2b Power Level vs. Axial Offset (5) The following Bases Sections were changed to reflect that certain operational limit 5 will be provided by TRSCO:
Bases Section Title 3/4.1.3 Mgveable Cont. Assemblies 3/4.2.3 FAH On the basis of the review of the above items, the NRC staff concludes that a
the licensee provided an acceptable response to those items as addressed in the NRC guidance in Generic Letter 88-16 on modifying cycle-specific parameter limits in TS.
Because plant operation continues to be limited in accordance with the values of cycle-specific parameter limits that are established using i
an NRC-approved methodology, the NRC staff concludes that this change is administrative in nature and there is no impact on plant safety as a consequence.
Accordingly, the staff finds the proposed changes acceptable.
The following two changes to the TSs were proposed to clarify certain surveillance requirements during plant start-up following a refueling outage.
Axial Offset The applicability statement of Technical Specifications 3.2.1.1, " Axial Offset-Four Loops" and 3.2.1.2, " Axial Offset-Three Loops" requires monitoring the axial offset when operating above 40% of rated power.
However, the excore/incore axial offset correlation cannot be accurately performed until a minimum of three days operation at 80% power (50% power for three loop operation) after start-up.
While the proposed TS surveillance requirement i
specifies continuous monitoring using the excore/incore axial offset correlation above 40% power, proposed Surveillance Requirement 4.2.1.3 does not require the correlation to be determined after a refueling or major change in excore Power Range instrumentation until exceeding 80% power.
The revised proposed TS will not require continuous monitoring of the Axial Offset after a refueling or major change in excore Power Range instrumentation using the excore/incore Axial Offset correlation until the excore/incore correlation can be determined and implemented prior to exceedin The requirement of not exceeding 80% power (50%g 80% of Rated Thermal Power.
power for three loop operation), combined with the successful completion of the zero power testing will provide assurance that the.LHGR will not exceed the initial conditions
't assumed for the loss of coolant accident (LOCA) analyses prior to determining the correlation.
All other required surveillances have been maintained.
Therefore, the staff concludes that the proposed changes are acceptable.
I DNB Parameter The present Surveillance Requirement 4.2.5.1.c requires verification of the reactor coolant system (RCS) total flowrate once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when operating i
in MODE 1.
However, Surveillance Requirement 4.2.5.2 allows the RCS total i
flow rate to be determined by heat balance within seven EFPD of Achieving Rated Thermal Power after a refueling.
In addition CYAPC0 states that Surveillance 4.2.5.2 cannot be accurately performed until achieving 100%
power.
The revised proposed TS transfers the RCS flow rate Surveillance 4.2.5.1c to Surveillance 4.2.5.2.
This will clarify the TS by stating that the RCS total flow rate need not be verified at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> until after the RCS total flow rate has been established.
The maintenance of the two other DNB-related parameters will prevent departura from DNB prior to establishing the RCS flow rate.
Therefore, the staff concludes that the proposed changes are acceptable.
Control Rod Insertion Limits The proposed change to TS 3.1.3.5 redefines the fully withdrawn position to be 317 steps instead of 320 steps.
All the physical models used in the cycle design and determination of safety analysis input parameters assume that the "all rods out" position to be 317 steps.
The 317 step position is based on the interface between the fuel assemblies and the control rods.
This change will allow greater operational flexibility in the positioning of control rods to minimize future control rod wear concerns and provide additional margin to accomodate drift in the individual rod position indicators.
Based on the above, the staff concludes that the proposed change is acceptable.
RCS Heatup TS 3.4.1.4.1 requires that at least one RHR loop be in operation in MODE 5.
One of the recommendations which resulted from analysis of the thermal shield repair was that no more than two reactor coolant pumps be operated at temperatures less than 350'F.
Recent experience has demonstrated that the RCS heatup is very slow with two reactor coolant pumps and one RHR pump l
operating.
The proposed change allows the RHR pump to be deenergized during heatup provided the following constraints are met:
1)
The deenergized RHR pump and LOOP are OPERABLE, 1
2)
The reactor coolant pumps in at least two unisolated loops are operating, with steam generator secondary side narrow range water i
level greater than 25%,
3)
No operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 4)
Core outlet temperature is maintained at least 10 F below saturation temperature.
m
i These constraints provide an adequate heat sink for operation in MODE'S because of the low decay heat.
Deenergizing the operating RHR pump in MODE 5 will allow a controlled RCS heatup without affecting the protective boundaries.
Based on the above, the staff concludes that the proposed changes are acceptable..
I RCS Hydrostatic 3nd Leak Testing The proposed changes to TS 3.4.9.1 allow the low temperature overpressure protection system (LTOPS) to be isolated during performance of RCS hydrostatic and leak testing.
In addition, the applicability of the LCO has been changed to apply during heatup, cooldown inservice leak and hydrostatic testing but not during criticality.
TS 3.4.9.3, "LTOPS" has also been changed to reflect that the LTOPS can be isolated during performance of RCS hydrostatic and leak testing.
CYAPC0 has stated it is not possible to perform the RCS hydrostatic i
and leak testing with the LTOPS inservice.
The failure mode of a low temperature, overpressurization event occuring below 315*F while the LTOPS is isolated has been evaluated.
It was determined that the 10 CFR Part 50 Appendix G margin of safety is maintained during the tests if the hydrostatic and/or leak test are performed above the required minimum temperatures of 245*F and 235 F respectively and a heatup rate of less than or equal to 10*F/ hour for one hour prior to and during the tests is maintained.
The minimum operating temperature requirement while critical is maintained by TS 3.1.1.6, " Minimum Temperature for Criticality" and therefore the reference to criticality in this TS can be removed.
TS 3.4.9 3," LTOPS restates that the LTOPS can be isolated during hydrostatic and leak testing.
Based on the above, the staff concludes that the proposed changes'are acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
This amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 i
CFR Part 20.
This amendment also involves changes in recordkeeping, reporting, or administrative procedures or requirements.
We have determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be releas'ed offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The staff has previously published a proposed' finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding.
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (10).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be preapred in connection with the issuance of the amendment.
L I
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4.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
Daniel B. Fieno Thomas G. Dunning Alan B. Wang 4
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PART 5 0F SAFETY EVALUATION RELATED TO AMENDMENT NO.125
3.0 INTRODUCTION
~
t On November 17, 1987, Connecticut Yankee Atomic. Power Company (CYAPC0/ licensee) submitted a proposed amendment to Facility Operating License No. DPR-61, to add operability requirements for onsite and offsite power sources with-limiting conditions for operation (LCO) and time requirements for corrective actions to Technical Specification (TS) 3.12. " Station Service Power." - In accitb, TS 4.2 " Operational Safety Items," was modified to include-requirements for testing and channel calibration of the undervoltage instruments. As a resu.t of a meetin 25, 1988, CYAPC0revisedandcombinedTS3.12,gwiththeNRConFebruaryStation Service Power"
" Emergency Power System Periodic Testing" into a newly titled TS 3/4.8, )
" Electrical Power System." The new TS submitted August 29, 1988 will: 1 incorporate the degraded grid voltage protection requirements, 2) incorporate emergency diesel generator requirements of Generic -Letter (GL) 84-15, " Proposed-Staff Actions To Improve and Maintain Diesel Generator Reliability," 3) incorporate industry improvements, 4) change the-custom TS format to one that is similartotheWestinghouseStandardTechnicalSpecification-(WSTS) format,and
- 5) incorporate requirements for battery discharge testing as required by the Systematic Evaluation Program-(SEP) Topic VIII-3.A, " Station Battery Test Requirements."
In addition the degraded grid undervoltage setpoints in the TS_
were changed.
These changes reflect the new station service transformers that were installed during the 1987 refueling outage. The proposed TS were supplemented by additional information provided in' letters dated June 9, 1989 July 19 and August 1,1989. The supplemental letters provided additional bases for several of the TS request. The supplemental information were within the scope of the original notice and did not affect the staff's determination in that original notice. This evaluation relates only to items (1) through (5).
A separate Safety Evaluation has been prepared for the degraded grid undervoltage setpoints TS changes, l
l 2.0 DISCUSSION As part of the SEP, CYAPC0 committed to convert their custom formatted TS to the WSTS.
Since the conversion effort did not start until October 1988 and with the impending issuance of a newly revised WSTS (merits), the staff proposed that it would be advantageous to await the issuance of the revised.WSTS.before addressing the full WSTS conversion.
In the interim, the staff and CYAPC0 agreed that the custom TS format could be upgraded to the WSTS format. The staff concluded that this interim step would:
- 1) provide a substantially im
- 2) proved TS while facilitating a future conversion effort to the revised WSTS,
-provide definitive LCO and action statements for several safety related systems, 3) eliminate the use of administrative TS at the Haddam Neck Plant,
- 4) provide a mechanism to close prior TS commitments associated with NUREG 0737, SEP and various other GL recommendations, and 5) eliminate the ambiguities inherent with the wording and format of the current TS.
Based on the above -the staff concluded that the improved TS would enhance public safety and therefore justified this interim step to improve the Haddam Neck TS.
The staff has informed CYAPC0 several times that this TS upgrade will not fulfill CYAPCO's SEP commitment to convert to the WSTS.
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This amendment is one of several that is part of the TS upgrade.
By letter dated September 22, 1987, the NRC provided Northeast Utilities with an acceptable revision of the WSTS.
The TS upgrade will be using the provided WSTS revision as a guidance while maintaining its current TS requirements.
Since the overall upgrade is primarily a format change, the staff did not pursue all deviations and omissions from the provided WSTS with the same I
intensity as would have been done for a normal WSTS conversion.
Therefore, if the proposed TS omits portions of the requirements that appear in the provided WSTS revision and these s ee requirements did not already exist in the current TS, the review of these omissions will be deferred to the full WSTS conversion.
However, where new TS statements have been proposed (statements not previously found in the current TS) that deviate from the provided WSTS revision, a review of the deviation will be given.
The deviations will be reviewed in part, based on three )reviously agreed upon criteria:
- 1) plant specific design, 2) previously approved.1ardware, structural or organizational changes, and 3) past operating experiences that can be shown to provide an equivalent degree of protection to that provided by the WSTS. Any deviations from the current custom TS will also be reviewed.
The format change and the additional restrictions resulting from this amendment make substantial improvements in the clarity and readability of the TS. As a result, the staff considered this TS upgrade beneficial from both a public safety and an operational perspective.
3.0 EVALUATION The evaluation has been divided into two sections.
Section I will address proposed TS that are consistent with the provided WSTS and/or the current TS.
In addition, many of these TS sections will add restrictions to the current TS.
Section II will address proposed TS that relax restrictions from either the current TS or the provided WSTS revision.
As noted earlier, the staff did not perform a " completeness" review to ensure that all sections of the WSTS were included in the proposed amendment.
Therefore, this review will exclude complete omissions of WSTS sections that did not already exist in the current TS.
Each of the deviations will be addressed individually.
If a GL or a.SEP issue has been addressed by the proposed TS change then it will also be noted.
3.1 Section I, Previously, the NRC staff provided a version of the WSTS to CYAPC0 and excluding plant specific alterations, has stated that the provided WSTS would be an acceptable guidance for a STS conversion.
Although this amendment is not intended as a STS conversion, CYAPC0 has submitted the amendment following the guidance of this WSTS revision.
The logic for this TS upgrade has been stated in the Discussion section of this Safety Evaluation.
Figure I provides a list of proposed TS that are consistent with the provided WSTS and/or the current TS.
In many cases the proposed TS impose added restrictions to the current TS or add restrictions that do not currently exist.
In all cases, the proposed TS listed in Figure 1 do not relax any of the restrictions found in the current custom TS.
Based on the above, the staff has concluded that the TS changes associated with Figure 1 are purely administrative (format change) or provide additional limitations, restrictions, or controls not previously included in the Haddam Neck TS.
Therefore, the staff concludes that the proposed TS listed in Figure 1 are acceptable.
3 3.2 Section II.
The TSs reviewed in this section will be addressed by number and subsection as it appears in the proposed TS. The following clarifications have been prov_ided-for this section of the review:
1)
The " current (or existing) TS" refers to the TS that is currently part of CYAPCO's operating license.
2)
The " admin TS" refers to an administrative 1y controlled TS that CYAPCO has been using in conjunction with the current TS. The admin TS is used by CYAPC0 to clarify the current TS and to provide additional requirements that CYAPC0 has found advantageous, through past operating experience.
3)
The " proposed TS" refers to the TS that CYAPCO has submitted-for NRC review as part of the TS upgrade.
4)
The "WSTS" refers to the copy of the Westinghouse Standard Technical Specifications that was provided by the NRC to-Northeast Utilities. This revision of the WSTS was provided-with a letter dated September 22, 1987-and has been used by CYAPCO as a guidance in the proposed TS upgrade.
Hereafter, "WSTS" will refer to this revision.
3.8.1.1 LCO b.1 The purpose of the LCO is to require that the diesel generator (DG) be equipped with a separate engine mounted fuel oil tank and to require that a minimum of 400 gallons of fuel oil be maintained in this tank.
i The proposed LCO is consistent with the WSTS except that it allows the fuel i
volume in the tank to drop below the stated minimum volume during DG' operation.
j The fuel oil transfer pumps take suction from the underground fuel oil storage l
tanks and transfer the fuel oil to the engine mounted tanks. The transfer pumps are controlled by level switches that are set to maintain a level of 400 gallons in the engine mounted tanks.
However, the differential setting of the level switches will allow the tank level-to drop below the 400 gallons before activating the transfer pumps.
Once activated, the pumps will refill the tanks to the required 400 gallon level. Therefore, the TS exception statement is necessary to prevent inadvertent TS violations that would result from the transfer pump controller design.
The staff determined that the proposed TS has met the intent of the WSTS and finds the proposed TS to be acceptable.
I 3.8.1.1 Action a The principal intent of this Action statement is to limit:the time allowed for continued power operation with less than two offsite AC power sources operable.
l If the failed circuit is not restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> then-the unit must be in-1 Cold Shutdown within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
During this time, the Action i
statement requires breaker alignment checks and DG operability tests. The purpose of these checks is to insure that alternate AC power sources are available to maintain the safety function of critical systems.
4.
The proposed TS meets the 72 and 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> requirements that are specifically stated in the WSTS and recommended by Regulatory Guide (RG) 1.93.
However, the proposed TS deviates from the WSTS in the surveillance intervals between breaker checks and DG tests.
In the first deviation, the proposed TS requires the breaker alignment to be checked within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.
The WSTS requires the breaker alignment to be checked within I hour and every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.
Both the WSTS and the proposed Action statement assume that an operable offsite circuit and both DGs are available. Following that assump-tion, there would be an alternate and diverse means to provide AC power to the safety related loads. The intention of the breaker alignment surveillance is
+
to insure that the preferred, operable offsite AC-source is available. The proposed TS checks this alignment six times during the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> interval and thereby, does provide assurance that the operable offsite source would be available if needed.
In addition, the WSTS 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> interval-implies that the surveillance should be performed once per shift.
CYAPC0 has stated that the intent of the proposed 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval, is that the surveillance will be performed once per shift while allowing some latitude in timing during that shift in which to perform the surveillance.
Therefore, the staff finds the proposed deviation to be acceptable.
The second deviation from the WSTS is that the DG needs to be demonstrated operable only if either DG has not been-successfully tested within the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The WSTS would require that the DGs be tested every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> during the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> interval. As a result, the WSTS requirement could lead to nine DG tests.
Following the guidance of GL 84-15 on DG reliability and testing frequency, the staff concluded that nine DG tests would be excessive.in this time frame. Futhermore, GL 84-15 states that frequent fast start testing from ambient conditions could result in an increased probability of DG failure. Therefore, after reviewing the basis of a similiar proposed TS change that was previously approved for the North Anna Power i
Station, Unit 2 (Amendment No. 48 issued April 25,1985) and using the guidance of GL 84-15, the staff concluded that this deviation is acceptable.
The current TS contains no such Action statements and only requires one offsite power source and one DG to to be operable for power operation.
However, CYAPCO currently uses a supplemental admin TS that has similar requirements to the proposed TS and has operated successfully in the past using this supplemental TS.
Based on the above, the current TS requirements and the availability of alternate AC sources, the staff has determined that the proposed TS meets the intent of the WSTS Action statement. Therefore, the staff finds the proposed Action statement to be acceptable.
3.8.1.1 Action b The principal intent of this Action statement is to limit the time allowed for l
continued power operation with less than two DGs operable.
If the inoperable DG is not restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> then the unit must be in Cold Shutdown within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
During this time frame, the Action I
statetent requires breaker alignment checks and the testing of the remaining operable DG, The purpose of these surveillances is to insure that alternate AC sources are available to maintain the safety function of critical systems.
1
5
- The proposed TS meets the 72 and 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> requirements that are specifically stated in the WSTS and are recommended in RG 1.93. However, the proposed TS deviates from the WSTS_in the surveillance intervals between brea(er alignment checks and DG testing.
Following the guidance of RG 1.93, one inoperable onsite source (DC)offersthesameseverityasthe-lossofoneoffsitesource.
Since the surveillance intervals for breaker alignment checks and the DG testing are the same as those stated in 3.8.1.1 Action a, the evaluation of
-these-two deviations is consistent with the evaluation of 3.8.1.1 Action a.
In addition, a statement has been added to this Action' statement that does not require the operable DG to be challenged if the inoperable DG was rendered inoperable due to preplanned maintenance or surveillance testing.
If the DG is inoperable due to preplanned maintenance it is assumed that the staggered testing frequency as recommended by GL 84-15, is sufficient to insure that the redundant DG is operable.
However, if one of the DGs has become inoperable due -
to some anomaly, it is.necessary to test the remaining operable DG to insure that it has not also been similarly affected. Determining'that the redundant DG is operable insures that the critical safety system loads can be powered should they be required. The proposed Action statement is specific and does.
require that the redundant DG be tested in this situation. This same exception statement was previously approved in Amendment No. 48 issued April 25, 1985 for the North Anna Power Station, Unit No. 2 and the basis for that approval is applicable to Haddam Neck.
Currently, CYAPCO's TS do not directly specify an action, or place'a time constraint en operation while the plant is in this l
degraded condition.
Based on the above, the current TS requirements, and the l
evaluation'of 3.8.1.1 Action a, the staff has concluded that the proposed TS has met the intent of the WSTS Action-statement.
Therefore, the staff finds the proposed Action statement to be acceptable.
3.8.1.1 Action e The principal intent of this Action statement is to limit the time allowed for continued power operation with one offsite AC source and one-onsite' AC source (DG) inoperable.
In addition, the Action statement provides a time constraint during which all AC sources must be made operable.
If at least one of the inoperable sources is not restored to operable status within the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> then the unit must be in Cold Shutdown within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.. In addition, if all AC sources are not restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> then the unit must be in Cold Shutdown within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
During this time frame, the Action statement requires breaker alignment checks and the testing of the remaining DG.
The purpose of these surveillances is to insure that the remaining AC sources are operable and available to maintain the safety function of critical systems.
CYAPCO's current TS only requires that one offsite and one onsite AC source be available during power operation. Oneofthedesignbasisevents-(DBE)ofthe plant is a LOCA with a loss of offsite power and a loss of a DG. With one AC offsite and one AC onsite source operable, redundancy is still provided by two
- diverse sources of power and this factor is considered in the DBE.
- However, the allowed time for continued operation in this configuration should be kept minimal.
The intent of the WSTS is to recognize the severity of the loss of both an onsite and offsite AC power source and to address it accordingly.
The
-proposed TS meets the 12 and 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> requirements that are specifically stated
~.
6 in the WSTS and recommended by RG 1.93..As a result, fore, proposed TS is imposing the an additional requirement over the current TS.
There the staff concluded that CYAPC0 has recognized the severity of this condition by imposing the added restrictions and by meeting the 12 and 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> WSTS requirements.
The proposed TS deviates from the WSTS in the surveillance intervals between breaker alignment checks and DG testing and by adding a statement that does not require the operable DG to be challenged if the inoperable DG was rendered inoperable due to preplanned maintenance or surveillance testing. These deviations are.
consistent with the proposed Action statements a and b.
Since the deviations are consistent with the previously proposed TS and the proposed Action statement meets the intent of the WSTS by recognizing the severity of this operating condition and imposing added restrictions to the current TS, the ~ staff finds the proposed Action statement to be acceptable.
3.8.1.1 Action d The principal intent of this Action statement is to provide assurance that a loss of offsite power event will not result in a complete loss of the safety function of critical systems while one DG is inoperable. The Action statement requires that with one DG inoperable, in addition to Action b or c, the operability of the charging pump, HPSI' pump, LPSI pump and RHR which depend on the remaining operable DG as a source of emergency sower must be verified..In addition, if these conditions are not satisfied wit 11n 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> the unit must be in Cold Shutdown within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The proposed TS meets the 2 and 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> time requirements for continued operation as specifically stated in the WSTS. The deviation from the WSTS arises from the wording of which equipment should be verified operable. The wording of the WSTS provides a general description of equipment that must be operable. The proposed TS provides a specific list of equipment to be verified operable.
The listed equipment in the proposed TS is the equipment that the operable DG must carry to maintain the safety function of critical systems.
I Furthermore, since the intent of the Action statement is to insure that the safety function of critical systems is not lost, the wording of the proposed-TS I
does reflect that intent.
In addition, the proposed TS deletes the WSTS references that require verification that the steam-driven auxiliary feedwater pump is operable. This deletion can be justified due to the Haddam Neck Plant design.
Unlike the standard Westinghouse plant that has two electric driven and one steam-driven auxiliary feedwater pumps, Haddam has two steam-driven auxiliary feedwater pumps. Therefore, having one inoperable DG would would not signifi-cantly affect auxiliary feedwater availability. As a result, the deletion will have no adverse impact on plant safety.
The current TS provides a similar restriction to the proposed TS and lists the-same equipment to be verified operable.
However, the current TS does not have the shutdown time requirements that the proposed TS has added.
Based on the above, the current TS and the additional proposed time constraints, the staff finds the proposed TS to be acceptable.
3.8.1.1 Action e The principal intent of this Action statement is to minimize the risk associated with two DGs (onsite sources) inoperable while avoiding the risk associated with an immediate shutdown.
The Action statement allows 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in
_. _ ~
7 which to restore one of the DGs to operable status or be in Hot Standby within 1
the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
In addition, if both DGs are not restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> the unit must be in Cold Shutdown within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
During the allowed time for continued operation, the Action statement requires that the offsite AC sources be demonstrated operable by performing breaker alignment surveillances.
The proposed TS meets the 2 and 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> requirements that are specifically stated in the WSTS and are recommended by RG-1.93.
The deviation from the WSTS is in the surveillance intervals between breaker alignment checks.
The surveillance interval in this Action statement is consistent with the intervals found in Action statements a, b and c.
Although the severity of this Action statement differs from that of the other Action statements, the staff concluded that the additional restrictions imposed by this Action statement do meet the intent of the WSTS.
Since the proposed TS does meet the intent of the WSTS by providing shutdown time requirements where none currently exist and is consistent with previously proposed surveillance intervals, the staff finds the proposed TS to be acceptable.
4.8.1.1.2 Surveillance a.1 The )urpose of this surveillance requirement is-to verify that the fuel volume in tie engine mounted fuel tank is at least 400 gallons.
The proposed TS is consistent with the WSTS except that it allows the fuel volume in the tank to drop below the stated minimum volume during DG operation.
This same exception statement. appears in proposed TS 3.8.1.1 LCO b.1) ano the design circumstances that apply to that TS are also applicable to this surveil-lance. As'a result, the evaluation for 3.8.1.1 LCO b.1) is applicable to this surveillance.
Therefore, the staff finds the proposed TS to be acceptable.
4.8.1.1.2 Surveillance a.4 The purpose of this surveillance requirement is to verify that the DG starts from an ambient condition and within 10 seconds is at a designated speed, voltage and frequency.
This surveillance is footnoted to provide limitations on the frequency of fast start surveillance testing and to specify that the mechanical stress and wear created by these tests be minimized.
The proposed TS deviates from the WSTS in the wording of the footnote and by not listing the start signals that are listed in the WSTS.
The WSTS footnote states that the surveillance testing should be preceded by an engine prelube period and/or manufacturer recommended procedures.
The proposed TS states that the testing shall allow for gradual acceleration to reduce stress and wear on the DG, The intent of this footnote is to reflect the recommendations of GL 84-15 and current industry standards for the reduction of wear on DGs.
GL 84-15 concluded that an overall improvement in diesel engine reliability can be gained by performing DG starts for surveillance testing using manufacturer recommended procedures..Rather than make the general statement of following the manufacturer's recommendations, CYAPC0 has stated that the proposed TS reflects their manufacturer's recommendation of gradual acceleration.
Therefore, the proposed TS does meet the intent of the WSTS and GL 84-15 through the l
proposed wording.
In addition, the proposed TS deviates from the WSTS by not l
8 1
s providing a specific list of start signals for the DG test.
The WSTS provides a diverse list of possible start signals and allows the operator to use any one of the listed signals for test initiation.
CYAPCO's current operating )rocedures already designate how the DG surveillance should be initiated.
CYAPC0 aas operated in the past with the existing procedures and found them to be effective t
l in demonstrating DG reliability.
As a result, CYAPC0 did not include a list l
of possible start signals as part of the TS.
In view of the diversity of the WSTS list, the staff determined that CYAPC0's operating procedures for DG starting do provide an equivalent level of protection to that of the WSTS.
Therefore, the staff finds the proposed deviation to be acceptable.
The proposed TS also deviates from the current TS.
The current TS requires that a DG surveillance must be performed monthly.
TheproposedTS(inthefootnote) requires the DG surveillance to be performed once in 184 days.
The increased surveillance intervals result in a reduction of DG fast starts which is consistent with the guidance of GL 84-15 and the WSTS.
GL 84-15 determined that frequent fast cold starts resulted in undue wear and stress on engine parts.
However, GL 84-15 also stated that the demonstration of fast start capability for DGs cannot totally be eliminated.
Combining these two conclusions, GL 84-15 provided an acceptable TS to reflect the findings.
The sample TS provided by GL 84-15 did specifically state the 184 day interval.
The proposed TS has met the intent of the WSTS by providing criteria to determine whether or not a DG start is successful.
Futhermore, it provides additional restrictions over the current TS and incorporates the guidance of GL 84-15. Therefore, the staff finds the proposed TS to be acceptable.
4.8.1.1.2 Surveillance a.5 This surveillance requires verification that once the DG is synchronized and connected to the bus, it is manually loaded to between 2750 KW and 2850 KW in less than or equal to 60 seconds and that it operates in that range for at least 60 minutes.
The surveillance statement is footnoted to limit the testing frequency and to require gradual loading for limiting mechanical stress and wear.
The proposed TS deviates from the current TS in the length of time in which the DG is required to remain loaded.
The current TS requires that the DG be loaded for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> while the proposed TS requires that the DG be loaded for at least 60 minutes.
The intent of the surveillance requirement is to provide sufficient assurance that the DG is available and can successfully operate in a steady state condition. Although the proposed change reduces the length of DG operation required by the current TS, it is consistent with the WSTS, GL 84-15 and the manufacturer's recommendations.
In addition, CYAPCO's operating procedures require a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> running time consistent with the current TS and past operating experiences.
CYAPC0 has stated that they intend to continue running the DG for the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period but have followed the WSTS in the wording
-("at least 60 minutes") of the proposed TS.
Therefore, the staff concluded that the proposed TS does meet the intent of the current surveillance, the intent of the WSTS and does reflect the guidance of GL 84-15.
The evaluation of the footnote for 4.8.1.1.2 surveillance a.4) also applies to this footnote.
Therefore, the staff finds the proposed TS to be acceptable.
. _ _ _ _ - ~
9 N
q 4.8.1.1.2 Surveillance b
]
This surveillance requires the verification that the automatic load sequence timers are within 10% of their design intervals.
The proposed TS deviates from the WSTS in the wording of the surveillance.
The WSTS provides a general statement that the interval between each load block is within 10% of its design interval.
The proposed TS provides a list of equipment to be sequenced on by the automatic timer and their perspective allowable elapsed times. The allowable elapsed times that are listed in the proposed TS are the exact times as required by the plant design basis.
The current TS has no such requirements but CYAPC0 has been operating with Admin TS that contain similar requirements to the proposed TS.
The staff concluded that the proposed TS meets the intent of the WSTS and finds the proposed TS to be acceptable.
_4.8.1.1.2 Surveillance d and e These two surveillances specify API, water, sediment, viscosity and testing requirements for fuel oil upon delivery and during underground storage.
The proposed TS deviates from the WSTS in wording and API gravity.
In particular, the proposed TS does not specify that new fuel oil will be sampled in accordance with ASTM-D4057.
Since 1976, CYAPC0 has procedurally followed the recommendations of RG 1.137 which references all ASTM procedures necessary
(
to meet the standards.
In following RG 1.137, CYAPC0 does sample in accordance with ASTM-D4057 and has stated that they will continue to sample in accordance with RG 1.137. However, CYAPC0 chose not to include the specific procedural number as a part of TS which may be subject to frequent revisions or replacement.
Since the current TS requires no such testing and CYAPC0 has successfully used the recommendations of RG 1.137 through its operating procedures in the past, the staff determined that the proposed TS does offer an equivalent level of protection to that of the WSTS.
The second deviation is in differing numerical values for API gravity.
The values for API gravity in the proposed TS differed from the WSTS by only a small amount.
These values were obtained through plant specific data and based on considerable past operating experience.
Since no such requirements exist in the current TS and CYAPC0 has i
used these numerical requirements successfully in the past, the staff l
determined that the numerical variations were acceptable based on the ground rules of this TS upgrade. The staff determined that the added restrictions of the proposed TS do meet the intent of the WSTS and are acceptable.
4.8.1.1.2 Surveillance f The purpose of this surveillance is to verify that with a loss of offsite power coincident with a Safety Injection Actuation Signal (ESF) the:
1 1) emergency busses will deenergize and shed load; 4
2)
DG will auto-start and energize the emergency busses with permanently connected loads within 10 seconds and energize the auto-connected shutdown loads and will operate for greater than or equal to 5 minutes (loaded);
i and maintains voltage and frequency requirements;
- - - =.
10 3) correct DG trips are bypassed; and 4)
DG capability to reject a load of greater than the largest single load.
CYAPC0 has submitted the proposed TS f.1 as an equivalent TS to WSTS 4.8.1.1.2.f.4 and WSTS 4.8.1.1.2.f.6.
WSTS 4.8.1.1.2.f.4 requires the above mentioned surveillances while simulating a loss of offsite power by itself.
WSTS 4.8.1.1.2.f.6 requires the above mentioned surveillance while simulating a loss of offsite power in conjunction with an ESF actuation test signal.
The Haddam Neck Plant uses discrete time delay relays for loading safety injection motors onto the electrical system. Whether a DG start is initiated by an ESF signal or loss of offsite power signal, the same diesel start and loading logic are used.
The difference between the signals results from the fact that an ESF signal will also initiate the loading of the safety injection i
loads.
Since a loss of offsite power signal alone or a loss of offsite power in conjunction with an ESF actuation signal will initiate the same diesel start and loading logic, one of the tests will verify the operability of the diesel start and loading logic.
By performing the surveillance requiring both the loss of offsite power and the ESF actuation signal, verification of the loading of the safety injection loads and the verification of the diesel start and loading logic are both accomplished.
Through surveillance procedures, CYAPC0 initiates the proposed TS surveillance first by an undervoltage condition which initiates eFtG, and then by an ESF signal which initiates a second DG and the safety injection loads.
By initiating the surveillance in this fashion, both initiation signals are tested.
Considering the proposed TS in conjunction.with the surveillance procedures and the plant hardware design, the staff determined that the proposed TS does meet the intent of the WSTS.
l The proposed TS in part 1.b and part 2 deviate from the WSTS by omitting.the l
statement requiring that the voltage and frequency shall be maintained within set limits af ter the bus is energized. The design of the Haddam Neck Plant on-site power system utilizes two GM/EMD 20 cylinder, turbo charged, low.
impedance generators. This design uses power current transformers to supply the needed energy to the exciter during motor starts while the voltage is depressed to as low as 50% of the DG rated value.
In addition this design allows for frequency swings during motor starts (loading). DurIngthe1980 refueling outage, a special test was conducted that simulated runout safety injection flow and worst case DG loading. CYAPC0 has stated that the test successfully demonstrated the on-site power systems capability to start and run the design basis loads without maintaining the voltage and frequency guidelines as set forth in the WSTS.
Based on the above, the staff determined that the plant design would not permit the precise wording of the WSTS without incurring unwarranted TS violations.
Therefore, the staff concluded that the proposed TS f.1.b and f.2 are acceptable.
The current TS only requires the demonstration of the readiness of the emergency power system to automatically start and restore power to the vital equipment by initiating a loss of normal AC power to each emergency bus.
The detailed requirements of the proposed TS and the WSTS are not in the current TS.
However, CYAPC0 has been performing the proposed surveillance through their Admin TS in the past.
Based on the above reviews of the individual parts of 1
l...,_ -
~.
11 this surveillance,-the absence of similar surveillance criteria in the current TS and the added restrictions imposed by the proposed TS, the staff has determined that the proposed TS is acceptable.
4.8.1.1.3 Surveillance-Reports This TS requires the licensee to report all DG failures to the Commission and i
include the information recommended in RG 1.108.
Additional information is required based on the number of failures within a valid test sample.
The proposed TS deviates from the WSTS by requiring that if the number of failures in the last 20 valid tests is greater than three, additional information will be reported in accordance with RG 1.108.
The WSTS requires that if the number of failures in the last 100 valid tests is-greater than seven, additional'information will be reported in accordance with RG 1.108.
Through the guidance of GL 84-15, the testing requirements for DGs were changed.
Subsequently, the reporting requirements were also changed. The reporting requirements of GL 84-15 are different from both the WSTS and the proposed TS.
However, the proposed TS does incorporate a portion of the reporting requirements found in GL 84-15. Although the proposed TS does not completely follow GL 84-15, the staff determined that it does meet the intent of the GL reporting requirements.
Therefore, the staff finds the proposed TS to be acceptable.
Table 4.8-1 Diesel Generator Test Schedule This table aetermines the DG testing f requency based on the number of failures in the last 20 valid tests.
As a result of GL 84-15, the testing requirements for DGs were changed to improve reliability and reduce unnecessary DG wear.
The GL reduced the testing frequency of the DGs and based the testing criteria on the number of failures per valid tests.
GL 84-15 provided a sample modified WSTS that reflected a number of the recommendations fc,und throughout the GL.
Included with the sample TS, was a DG test schedule table.
The sample TS table included the reduced testing frequency based on the number of failures in the last 20 valid tests.
The proposed TS follows the guidance of GL 84-15 and the sample TS table.
The WSTS revision used in this TS upgrade presents the testing frequency in a different form and includes tests not required in the GL 84-15.
Since the proposed TS table does follow the guidance of GL 84-15, the staff finds it to be acceptable.
l 3.8.1.2 Action a When in N0 DES 5 and 6, the purpose of this Action statement is to immediately suspend all operations involving Core Alterations, positive reactivity changes, movement of irradiated fuel or crane operation with less than one DG and one offsite circuit operable.
The Action statement also requires immediate action in MODE 5 if less than two steam generators are operable and in MODE 6 if water level is less than 23 feet above the reactor vessel flange.
l G
w-e
~
,,,eaw-a,--,
12 The proposed TS deviates from the WSTS by omitting the statement requiring RCS venting.
The Haddam Neck Plant has a separate, dedicated system called the Low Temperature Overpressure Protection (LTOP).
The LTOP is a system capable of protecting the RCS against pressure transients which could exceed the limits of Aspendix G to 10 CFR Part 50 when one or more of the RCS cold-legs are less tien or equal to 315 degrees F.
The operation of the LTOP is currently covered by TS 3.3.4.2.
The LCO, Applicability and Action statements of TS 3.3.4.2 do coincide with the plant conditions in proposed TS 3.8.1.2.
Since the LTOP is capable of venting the RCS and since by TS, the LTOP is required to be operational in the plant conditions of proposed TS 3.8.1.2, the staff determined that the proposed TS does meet the intent of the WSTS.
Therefore, the staff finds the proposed TS to be acceptable.
3.8.2.1 Action 1
The purpose of this Action statement is to limit the time allowed for continued operation with the available onsite DC supplies one less than the LCO.
The-4 Action statement allows a short time interval in which the affected DC supply must be restored.
If the affected DC supply is not restored within that time, the unit must be in Cold Shutdown within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The proposed TS meets the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> requirement that is specifically stated in the WSTS and recommended by RG 1.93.
The proposed TS deviates from the WSTS in the time allowed for continued power operation.
The proposed TS allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of continued operation in comparison to the WSTS which allows 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of-continued operation. The primary intent of the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> requirement stated by the WSTS is to minimize the risks associated with only one operable DC source and provide constraints on continued operation.
By comparison, the current Haddam Neck TS only requires that one battery charger must be in service ano provides no direct Action statements for a degraded condition.
However, the proposed Action statement follows from a proposed LC0 requiring two battery banks and hssociated chargers to be operable.
It was the opinion of.the staff, that when considering the current TS, the added restrictions and clarity of the proposed TS are a substantial improvement over the current TS and do reduce the current risk associated with this degraded condition.
Furthermore, the licensee contends that the proposed 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period would allow time to attempt successful repairs on the inoperable DC supply while minimizing the risks associated with continued operation and a forced shutdown with no redundant onsite DC supply.
Based on the above and the increased level of safety resulting from the proposed TS, the staff determined that the proposed TS does meet the intent of the WSTS.
Therefore, the staff finds the proposed TS to be acceptable.
4.8.2.1 Surveillance b This surveillance requires that once per 92 days and within 10 days after a battery discharge or overcharge, that specified battery parameters be verified and that the resistance of terminals or connectors be verified to be less than a specified value.
The proposed TS meets the 92 day interval that is specified by the WSTS but deviates in the surveillance interval after a battery discharge or overcharge.
The WSTS requires that within 7 days af ter a discharge or overcharge, that this surveillance be performed.
The proposed TS allows a 10 day interval.
Currently, CYAPC0 has procedures that follow both the manufacturer and IEEE 450
13 i
recommendations.
These procedures require that an equalizing charge may take as long as 6.5 days to complete at which time the batteries are placed on a float charge for 3 days.
CYAPC0 has operated with these procedures in the past and has found them to be an effective means in which to verify battery surveillance parameters.
Since CYAPC0 is following both the manufacturer and industry recommendations and has an effective procedure already in place for this surveillance, the staff finds the proposed TS to be acceptable.
Table 4.8-2 Battery Surveillance Requirements 1
i This table lists the parameters for battery surveillance requirements for
~
weekly and quarterly surveillances.
The proposed TS table is consistent with the WSTS except for slight numerical deviations. The numerical values in the proposed TS reflect both CYAPCO's past operating experiences and the manufacturer s recommendations.
Since the intent of this table is to insure that the batteries are maintained in a reliable operating condition, the staff concluded that the plant specific and manufacturer's data warranted the numerical deviations.
Based on the above, the staff determined that the proposed TS does meet the intent of the WSTS.
Therefore, the staff finds the proposed TS to be acceptable.
3.8.2.2 Action a The purpose of this Action statement is to immediately suspend all operations involving core alterations, positive reactivity changes, movement of irradiatea fuel or crane operation with less than one battery bank and associated charger operable.
In addition, the Action statement requires the RCS to be vented.
l The proposed TS deviates from the WSTS by omitting the statement requiring RCS venting.
As with the proposed Action statement of 3.8.1.2 the LTOP system is also used in this case as an equivalent vent path.
Since the deviation and operating conditions of that TS are consistent with this LCO, the evaluation of 3.8.1.2 applies to this LCO.
Based on the evaluation of proposed TS 3.8.1.2 the staff finds the Action statement of 3.8.2.2 to be acceptable.
_3.8.3.1 Action b The purpose of this Action statement is to require operator action with the loss of a vital bus and/or its associated inverter.
The Action statement provides time constraints in which to restore the vital bus to its normal configuration or be in COLD SHUTDOWN within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The proposed TS deviates from the WSTS by allowing an optional means in which to energize a failed vital bus from another source.
The proposed optional means is a Haddam Neck Plant specific design feature.
The WSTS assumes that there is available an alternate, independent hource of_ power for the vital busses (other than the associated inverter n Accordingly, the WSTS provides a limited time in which to reenergize the vital bus and restore it to its normal operating conditions.
The Haddam Neck Plant design does not have an alternate, independent source that can be used to reenergize the vital busses.
However, the Haddam design does allow the crosstying of vital busses between inverters.
CYAPC0 has i
l proposed this option as part of the proposed Action statement.
14 The current configuration would allow the vital busses to be crosstied across safety divisions.
After review, the staff found the crosstying between safety divisions to be unacceptable.
However, as part of SEP, CYAPCO committed to the addition of a new separated switchgear room and bus arrangement.
The design would permit the DC system to meet current plant design and separation criteria.
Along with the new design the bus arrangement would be altered such that vital busses would have the ability to be crosstied with another inverter within the same safety division.
Haddam has built the new switchgear room and intends to put the new configuration in service during the current outage.
The electrical portion of the new switchgear room and bus design has been reviewed and approved.
Based on the new design the staff has analyzed the proposed course of action. The staff finds the proposed Action statement to be acceptable for the following reasons:
1)
The new switchgear room and bus design will maintain the separation (both electrical and physical) between the two safety divisions.
Therefore, the crosstying of two vital busses will only be within one safety division.
Based on the staff's analysis of the provided information, the staff concluded that it is not acceptable to crosstie between two safety divisions at power.
2)
CYAPC0 has performed an analysis and determined that a single inverter can adequately carry the loads of two vital busses for the duration of the Action statement.
3)
For the duration of the Action statement, a compensatory measure will be ta ken.
This measure will consist of placing the reactor protection system channel of the failed bus in the tripped condition.
4)
CYAPC0 has performed analysis and determined that the isolation devices at the output of each inverter will protect the crosstied inverter from a faulted condition that may exist on the failed vital bus.
5)
The length of continued operation in this configuration will be limited to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Af ter 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> the plant will shutdown if the vital busses have not been restored to their normal configuration.
l 6)
CYAPC0 has stated that the time to reenergize the failed vital bus (8 l
hours), results from the method by which the loads of the failed vital bus will need to be loaded onto the nonfailed (crosstied) vital bus.
7)
A prior Safety Evaluation of SEP topic VI-7.C.1 has stated that with an acceptable new bus separation design (the switchgear room and bus configuration changes), such a crosstie would be permitted.
l Furthermore, the staff compared the severity of a failed vital bus without any means to be reenergized with a limited continued operation time for a crosstied configuration.
Based on the above review and this comparison, the staff determined that this Action statement is acceptable.
15 i
The Haddam Neck Plant design also permits the crosstying of other redundant busses between safety divisions (trains).
This evaluation should not be construed as to find such a procedure acceptable.
In fact, the staff has a
found it to be unacceptable to crosstie redundant busses between safety divisions.
3.8.3.1 Action c This Action statement addresses the operator response and time constraints with one DC bus not energized from Its associated battery bank.
i The proposed TS deviates from the WSTS in the time allowed for continued operation.in this condition.
The proposed.TS allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of continued operation whereas the WSTS allows 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The current TS simply requires one battery charger to be operable with no definitive Action statements.
The added restriction does show CYAPCO's recognition of.the severity of this operating condition and does define a course of action for this condition.
Furthermore, the proposed time is consistent with the LC0 and Action statements of 3.8.2.1.
Based on the above, the staff concluded that.the proposed TS is acceptable.
3.8.3.2 Action a During MODES 5 and 6, this Action statement requires operator action with the loss of the electric service busses as listed in the LCO.
The deviation from the WSTS is in both wording and the RCS venting requirement.
The operating conditions are consistent with the LCOs 3.8.1.2 and 3.8.2.2.
Based on the evaluations of thuse Action statements, the staff concluded that the proposed Action statement a, of 3.8.3.2, is acceptable.
4.0
SUMMARY
After checking the current TS sections 3.12 and 4.5, the staff determined that the current TS requirements have been maintained by the proposed TS.
Furthermore, the proposed amendment offers not only an improved format over the l
current TS but also adds numerous TS restrictions to plant operation.
Based on the considerations discussed in the above evaluation, the staff concluded that i
j the proposed amendment will make overall improvements in the operational safety while maintaining the current safety analysis.
Therefore, the staff finds the I-proposed amendment to be acceptable.
5.0 ENVIRONMENTAL CONSIDERATION
1 This amendment changes a requirement with respect to the installation or use of i
a facility component located within the restricted area ~as defined in 10 CFR Part 20 and changes surveillance requirerrents..We have determined that the amendment involves no significant increase;in the amounts, and no significant L
change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational.
radiation exposure.
The staff has previously published a proposed finding that the amendment involves no significant hazards consideration and there has been i
1
16 no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment, t
1 I
6.0 CONCLUSION
1 We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and-safety of the public.
Principal Contributor: G. E. Garten 4
1 1
i
17 1
Figure 1 LC0 - Limiting condition for operation APP - Applicability l
SURV - Surveillance L
ACT - ACTION-l TS NUMBER SUBSECTION TYPE l
3.8.1.1 a
LCO l
- 3. 8.1.1.
b-2, 3 LCO i
3.8.1.1 APP..
4.8.1.1.1 a
SURV-4.8.1.1.2 a-2,3,6 SURV 4.8.1.1.2 c
SURV 4.8.1.1.2 f-3 SURV 3.8.1.2 a,b LC0 3.8.1.2 APP i
3.8.1.2 b
ACT.
l 4.8.1.2 SURV 3.8.2.1
.a,b LCC 3.8.2.1 APP 4.8.2.1 a
SURV 4.8.2.1 c
SURV 4.8.2.1 d e.f SURV 3.8.2.2 LCO 3.8.2.2 APP 3.8.2.2 b
ACT l
4.8.2.2 SURV L
3.8.3.1 a,b,c,d e LC0 L
3.8.3.1 f,g.h LCO 3.8.3.1 APP 3.8.3.1 a
ACT i
l
. 4.8.3.1 SURV 3.8.3.2 a,b,c d LC0 3.8.3.2 APP 3.8.3.2 b
ACT 4.8.3.2 SURV bases l
t f
i t
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o
]
PART SA 0F SAFETY EVALUATION RELATED TO AMENDMENT NO.125
1.0 INTRODUCTION
i Connecticut Yankee Atomic Power Company (CYAPCO): replaced station service transformers during the 1987 refueling outage to eliminate a potential PCB-hazard.
Because of differences in the replacement transformer impedances; the-degraded grid voltages available to Haddam Neck safety equipment are different than those previously approved by NRC in the degraded grid operating procedure safety evaluation letter dated July 2,1985.
Evaluations have been made by the-licensee with the new transformers in the system under various plant and grid. conditions including conditions of degraded grid voltage. As a consequence of these evaluations, CYAPC0 proposed by letter dated November 17, 1987, as revised August 29, 1988, to amend the Technical Specification (TS) t degraded grid undervoltage setpoints.
This safety evaluation covers-these changes.
2.0 BACKGROUND
By letter dated October 21, 1981, CYAPC0 proposed. technical specification:
changes to include the additional requirements and limiting conditions for.
operation associated with a degraded grid voltage protection. system proposed in response to NRC staff positions letter dated June 3, 1977..The NRC safety evaluation dated July 9,1982 concluded that the proposed technical specification modifications for degraded voltage were acceptable.
- However, since manual operator actions were required in response to degraded grid conditions, the-staff requested submission of appropriate operating procedures. Accordingly, CYAPC0 submitted Abnormal Operating Procedure l
A0P-3-2-25 on February 3, 1983.
By letter dated July. 2.1985, the staff!
L provided a safety evaluation of the A0P procedure, finding that it was acceptable.
However, the degraded grid voltage action-level numerical values in the procedure were not consistent with those in the TS.
Therefore, the staff requested that CYAPC0 revise and resubmit the TS to reflect the proper numerical values as contained in the approved procedure.
CYAPC0 submitted the proposed Technical' Specification degraded grid voltage changes by letter dated November 17, 1987. However due to voltage differences caused by replacement of the feeder transformers,,the numerical values are different from those previously approved.
This Safety Evaluation (SE) is only for the numerical.
voltage setpoint change values. A separate SE will. evaluate the remaining portions of the licensee's November 17,1987 (as revised) submittal (Part 5).
i-
3.0 PROPOSED CHANGE
S The proposed changes consist of revising Technical Specifications Sections 3.12 and 4.2 as follows:
b
c a
L 2
o 3.1' Section 3.12. Station Service Power Section 3.12 B)(1), revise the 4160 volt emergency bus specification level three undervoltage set oint range from "below the level three undervoltage setpoint L3980V, but above 3642 volts" to "below the level three undervoltage setpoint 4019V)butabove3684 volts."
Section 3.12 B)(2), revise the 4160 volt emergency bus specification level two undervoltage setpoint from "3642 volts to "3684 volts."
i Section 3.12 8) Basis, revise the 4160 volt emergency bus basis undervoltage values from 3980 and 3642 volts, respectively, to 4019 and 3684 volts, respectively, i
3.2 Section 4.2. Operational Safet.y items Revise the Table 4.2-1 undervoltage protection calibration setpoints as follows:
Channel 31, 4.16kV Emergency Bus Undervoltage - Level 2; change both' the 1
4.16kV emergency bus undervoltage level two trip setpoint and allowable value from "3642 volts" to "3684 volts."
3 Channel 32, 4.16kV Emergency Bus Undervoltage - Level 3; change both the 4.16kV emergency bus undervoltage level three trip setpoint and allowable value from "3980 volts" to "4019 volts."
4.0 REVIEW CRITERIA / REQUIREMENTS NUREG-0452, Standard Technical Specifications for' Westinghouse Pressurized Water Reactors, Revision 4.
l Code of Federal Regulations,10 CFR Part 50 Appendix A, General Design Criteria 17 - Electric Power Systems.
NRC letter to CYAPCO, Safety Evaluation and Statement of Staff Positions Relative to the Emergency Power Systems for Operating Reactors, June 3, 1977.
5.0 TECHNICAL EVALUATION
/ DISCUSSION Review of the November 17, 1987 CYAPC0 proposed TS degraded grid voltage revisions consisted of an evaluation of the licensee's basis for the numerical values for the level'three and level two undervoltage allowable values and the degraded grid voltage instrumentation setpoint values.
The licensee's basis for revising these values are electrical system impedance changes due to replacing feeder transformers which provide the power to the Class IE safety-related systems. As a consequence of these changes, the voltages available to the loads are different and they vary depending upon the conditions of the grid and the magnitude and characteristics of the load.
The licensee has conducted evaluation case study analyses involving a total of 35 different electric grid supply and load configurations including both steady
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state and transients in order to envelope the range of voltages which coula occur on the 4160 volt safety related buses.
Based upon the analyses and upon previously established minimum starting and operating voltages required for the safety-related equipment the licensee has established the revised 4019 voltlevelthreeand3684voltleveltwosetpointsandallowablevalues.
The 1
staff has-reviewed the licensee analysis and voltage values resulting from the impedance changes due to the replacement of the feeder transformers and find the new values to be acceptable.
6.0 ENVIRONMENTAL CONSIDERATION
This amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. We have determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The staff has previously issued a proposed finding that his amendment involves no significant hazards consideration and there has been no public comment on finding. Accordingly, the amendment meets the eli categorical exclusion set forth in 10 CFR 51.22(c)gibility criteria for (9). Pursuant to 10 CFR 51.22(b), no environnental imsact statement or environmental assessment need be prepared in connection wit) the issuance of this amendment.
7.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will beconductedincompliancewiththeCommission'sregulations,and(3)the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
8.0 REFERENCES
DOCUMENTS j
CYAPC0 Degraded Grid Voltage Protection Response Letters to NRC June 3, 1977, letters within the period August 1,1977 to April 21, 1982, i
l NRC letter to CYAPCO, Safety Evaluation of Proposed Degraded Grid Voltage l
Protection System, July 9,1982.
(
CYAPC0 Degraded Grid Voltage Protection Systems Proposed Operating Procedures, February 3 and 14,1985.
NRC letter to CYAPCO, Safety Evaluation of Degraded Grid Voltage j
Protection System Operating Procedures, July 2, 1985.
l CYAPC0 Revised Degraded Grid Voltage Protection System letter to NRC, l
November 17, 1987.
Principal Contributor:
C. H. Woodard, Region I
1 j
Part 6 of Safety Evauation Related to Amendment Ro!*125
1.0 INTRODUCTION
Connecticut Yankee Atomic Power Company (licensee) has upgraded portions of the ReactorProtectionSystem(RPS)intwophases. Phase I was evaluated in a previous Safety Evaluation (SE).
This SE will provide the evaluation of the second phase of the upgrade. Certain aspects-of this evaluation are the same as previously performed for Phase I and will refer to the-previous SE where appropriate to reduce repetition. The Phase I RPS upgrade SE.was issued e
to the licensee March 21, 1990.
This SER will also evaluate the Nuclear InstrumentationSystem(NIS)upgradeandtheassociatedTechnicalSpecification (TS) changes for both parts of the Phase;II upgrade.
By letter dated September 1, 1988 the licensee submitted preliminary information concerning the RPS Phase II and NIS upgrades. The licensee stated that the information was provided for information purposes only and was not requesting NRC review or approval. The licensee has stated that these changes will be implemented in compliance with 10 CFR 50.59.
In addition to the technical evaluation of the physical changes this SE will also address the appropriateness of making the changes via the 10 CFP,50.59 rule.
By letter dated July 28, 1989 the licensee submitted the proposed changes to Technical Specifications associated with the RPS Phase II and NIS upgrades.
These changes were described using the new Standard Technical-Specification (STS) format. This SE will address only those changes specifically associated with the described upgrades and is not intended to review the remainder of the STS format changes which will be evaluated by a separate SE.
l
2.0 DESCRIPTION
AND EVALUATION This section will describe the physical changes being implemented, discuss the-NRC review criteria and provide our evaluation of the changes.;
2.1 Reactor Protection System Phase 11 Upgrade Descriptjon t t s
The RPS Phase II changes are a continuation of the modernization effort of Phase I which includes the replacement of sensors. transmitters and Main Control Board equipment. PhaseIIisbeinginstalledviaPlantDesignChange Record (PDCR) No. 952. The following Systems are affected:
a Reactor Coolant System Flow b
Reactor Coolant System Pressure c
Primary Containment Pressure
i s
2 d)
Steam Generator Narrow Range level (transmitter replacement only) e)
SteamGeneratorSteamFlow(transmitterreplacementonly) l The Reactor trip relay logic system is being replaced with a solid state j
Foxboro Spec 200 Micro logic system. This change will involve changing the logic implementation, the field interfaces, bypass and defeat abilities and l
on-line testing capabilities.
In addition Power Dependent Insertion Limit 1
(PDIL) circuitry is being added to the Rod, Control System.
The details of each l
change are listed below f
The Reactor Coolant System Low Flow trip circuit has been substantially changed
(
are being replaced with twelve-(three channels per loop)p) flow transmitters from the existing system. The four (one channel per loo new qualified Foxboro transmitters. The three transmitters per loop-will use the same tap so there-are no additional pressure boundary penetrations.
Each of the-three transmitters per channel will be powered.from one of the A, C or D vital power buses. The output of each transmitter is input to individual Foxboro Spec 200 microprocessors which compare the flow to the setpoint and provide an l
electrically isolated (via Foxboro L2CR isolator) output to each of four separate Spec 200 micros.
Each of the four microprocessors receives the output from each of the three transmitters and does a 2 out of 3 coincidence which if satisfied provides an isolated output trip signal.
Each of the 2/3 comparators is powered from a different vital bus. Each channel (total of 4 channels, one per fluid loop) has four isolated separate trip outputs for a. total of sixteen l
trip output signals.
1 One output from each loop then is input to another set of four microprocessors (also powered from each of the four vital buses) where the P7 and P8 permissive are compared with the transmitter low flow trip signals.
This section is.the same as the existing design except that it is accomplished with software within the microprocessor and there are four complete sets of coincidence lo of the four isolated outputs are hardwired together (two out of two) gic.
Two for the Train A breaker trip and the other two are combined (also hardwired two out of two) for the Train B breaker trip.
This total logic train from transmitter to breaker is designed such that there is no single failure of sensor, transmitter, microprocessor, cable or power supply that would cause a trip-or prevent a valid trip signal.
This configuration also allows increased bypass and testing abilities without a single failure during testing causing a reactor trip.
This logic configuration is acceptable to the staff.
l PDCR 952 will remove the two existing Reactor Coolant System Pressure wide range (0-3000 psig)). transmitters from loop 4 and will install qualified wide L
range (0-3000psig and narrow-range (0-600 psig) transmitters on loop 4 and add a redundant pair of wide and narrow range transmitters to loop 1.
In I
addition to the added redundancy the narrow-range transmitter will provide a more accurate pressure signal to the Residual Heat Removal (RHR) system and the l
Low Temperature Overpressurization System (LTOPS) interlocks.
This j
. modification is acceptable to the staff.
l
-~ -
~
. The existing six Primary Containment Pressure Switches (mercoid) will be replaced with four Primary Containment Pressure Transmitters.
The logic will change from a 2 out of 3 taken twice configuration to 2 out of 4. The new pressure transmitters are expected by the licensee to provide higher accuracy and better repeatability.
This change is acceptable to the staff.
The Steam Generator Narrow Range I.evel transmitters will be replaced with the new qualified transmitters.
This change is primarily to replace obsolete equipment with new qualified reliable equipment and is acceptable to the staff.
The Steam Generator Feedwater Flow transmitter upgrade described in the September 1, 1988 submittal has been postponed by the licensee until other feedwater modifications are scheduled are not included in PDCR 952 and therefore are not considered as part of this SE.
2.1.1 System and Hardware Evaluation The changes described above will use Foxboro transmitters and input / output modules. The Foxboro Spec 200 Micro equipment used are digital microprocessors which use software to implement the various functions, The changes described above will also require new wiring, cable, instrument racks and tubing which are seismically qualified.
The requirements that this equipment must meet are the same as the Phase I RPS upgrade and were discusseo in the previous SE.
The equipment used is the same as Phase I and the applications are similar.
The equipment is acceptable to the staff for use in the Phase II upgrade.
The licensee has made many changes which will reduce the possibility of inadvertent reactor trip from a single component or power supply failure.
The remaining open issue is potential common mode problems which would defeat the redundancies added by the licensee. The staff reviewed potential common mode failure mechanisms.
Use of qualified Class-1E components and verification and validation of software reduces the potential for common mode mechanistic or programming errors to an acceptable level.
The one open area which the staff believes was not adequately addressed was the potential for electromagnetic interference or voltage perturbations on the power buses to cause unacceptable operation of several microprocessors at one time.
As described in the Phase I SER, Foxboro had performed specific testing which established a level of electrical environment qualification.
The staff required that the licensee determine that the electrical environment at the installed equipment was enveloped by the vendor testing.
The staff required that a plan for determining this be submitted to the NRC for review prior to startup from the Phase II installation.
The conclusions from Phase I apply for Phase II with additional emphasis due to the new use of microprocessors to implement two out of two reactor trip logic.
EMI induced problems in microprocessors may be more complex than a loss of power or electromechanical problem in an analog sys tem.
2.1.2.
Software Assessment The Phase I SE described the verification and validation of the software used in the Foxboro Spec 200 Micro and the Haddam Neck Plant configuration control and found them acceptable for that application. As described in that SE a few
4 0
notes of caution were listed that some applications may be extremely complex or require extreme speed and were not addressed in that SE.
The staff has reviewed the Phase II RPS upgrade and has concluded that the situations which were cautioned against are not being used in Phase II and therefore are not a concern.
Each specific segment of the logic which has been implemented with this software is relatively simple which provides a high degree of confidence that the verification and validation reviewed for the Phase I upgrade is adequate for Phase II.
2.2 Nuclear Instrumentation System Upgrade Description il This change will involve replacement of the existing ex-core Nuclear Instrumentation System (NIS) and is implemented with PDCR 954. The ex-core detectors, cabling, scaler / timer, rod disconnect panel, preamplifiers and main control board equipment will be replaced.
The Boger Communication Module and the Refueling Cavity Level Indicator will be relocated.
The primary equipment supplier will be Gammametrics.
This upgrade will retain the four power range channels but there will now be ten reactor trip steps instead of the previous three.
The new ranges are selected by a ten position switch on the Power Range Drawer.
Rod Stops will also be increased to ten.
The Power Range channels will use fission chambers instead of the previous uncompensated. ion chambers.
Four wide range channels will replace the existing two intermediate range channels.
These channels will use fission chambers to replace the current compensated ion chambers.
Reactor trip on Hi SUR will now be a 2/4 logic.
Four Source Range Channels will replace the existing three source channels and will share the same detectors as the wide range channels, 2.2.1 NIS System and Hardware Evaluation The new NIS is expected by the licensee to be much more reliable than the old system, easier to maintain and less noise sensitive.
This equipment has been previously accepted for use at other facilities.
Northeast Utilities performed a reliability analysis (dated June 23,1989)for the NIS which concluded that the overall NIS reliability would be improved due l
to increased redundancies and modern equipment.
The previous equipment had become obsolete and was having an increasing negative effect on system and l
plant availability.
The existing relay matrix logic system which develops the permissives and reactor trips is being replaced with two completely redundant coincidentors.
lhese coincidentors will develop the trips and permissives utilizing solid state electronics. The resultant signal from each coincidentor will then go to-the RPS logic cabinets.
The NI' upgrade is acceptable to the staff.
'[
. 1 2.3 Techniccl Specifications The changes to the TS were provided by letter dated July 28, 1989. This letter described the changes using the new Standard Technical Specification (STS) format.
This review addresses only those changes specifically related to the RPS Phase II and NIS upgrades.
The proposed changes are described below.
1)
A definition of Reactor Trip System Response Time was.added.
A surveillance of response time was also added.
A new table providing the RTS instrumentation response times was added.
2)
The RTS/ESF and Accident Monitoring tables were revised to reflect the new NIS LC0's and surveillance requirements were added.
3)
The NIS Analog Channel Operational test was changed from every 14 to every 41 days.
A 41 day requirement of trip actuation and device operational testing was added.
The 14 day requirement is no longer required since the equipment operational difficulties have been resolved via replacement with the new NIS.
The TS changes are acceptable to the staff.
2.4 RPS Response Time As a result of the RPS upgrade the time responses for the power range nuclear flux, start-up rate and low-flow reactor trips have been lengthened relative to previous analysis.
the change in response times are as follows:
1)
PowerRangeNuclearFlux(Overpowertrip) 0.25 sec. to 0.5 sec.
- 2) Start-up Rate Reactor Trip 0.4 sec. to 1.0 sec.
j
- 3) Low-flow Rate Reactor Trip 1.15 sec, to 1.85 sec.
The licensee determined that of the thirteen transients evaluated for the Chapter 15 non-LOCA transient analysis only seven are affected.
CYAPC0 has evaluated the affect of the response time delays on the following transients:
- 1) uncontrolled rod withdrawal from subcritical, 3) steam line break, 4) power, 2) uncontrolled rod withdrawal from flow, 6) locked rotor / sheared shaft and 7) jection, 5) loss'of reactor coolant RCCA e idled and isolated loop start-up.
1)
Steam Line_ Break The steam line break (SLB) is a cooldown event which is concerned with post trip return to power.
In the SLB analysis, reactor trip is on high power, caused by positive moderator feedback resulting from the break. The stored energy in the fuel and the initial negative Doppler and moderator reactivity insertion due to fuel and moderator heat up prior to trip are minimized by assuming no delay on the reactor trip signal.
If the RPS delay were included, the pre-trip l
heat-up would result in a slight negative Doppler and more negative moderator reactivity insertion.
In addition, the stored energy in the fuel would also be increased. The slight negative reactivity insertion would reduce the positive reactivity contribution from the j
cooldown.
For steam line break transients the consequences are more severe by having an earlier trip.
Therefore, delaying the trip will still result in minimum DNBRs and peak fuel centerline temperatures.
that are bounded by previous analyses.
2)
TnTEinaHTEo~plMQ, Locked Rotor / Sheared ShafL and Idled Loss of Reactor Coolant Flow For the loss of reactor' coolant flow and locked rotor / sheared shaft transients the increase in the low-flow trip delay time has been-compensated by a reduction-in the conservatism assumed in:the radial peaking factors.
For the idled and isolated loop start-up transient the increase in the overpower trip response time is also compensated by the reduction in the conservatisms assumed in the radial peaking factors.
In all three cases the radial peaking factors, assured-by the TSs, continue to bound the analysis assumptions and the predicted minimum DNBRs and peak fuel centerline temperatures are bounded by previous analysis.
3)
Uncontrolled Rod Withdrawal from Suberitical For the uncontrolled rod withdrawal from subcritical transient the increase in the'hightstart-up rate trip response time is compensated for by the new wide range channels. The new wide range channels will
't be able to detect and trip the event at initiation, while the original equipment could not detect and trip the event until the power levels reached the intermediate range channels.
This will result in a trip.
at a much lower power level. The longer delay times with the expanded sensitivity of the new wide range channels resulted in minimum DNBRs -
and peak fuel centerline temperatures that are bounded by previous analyses.
4)
RC M jection for the RCCA ejection transient the increase in-the overpower trip delay is compensated for by an improved pin census for Cycle 16 and the use of a less conservative, but still bounding, gap conductance.
The pin census is a distribution of the number of fuel pins as a function of post ejection radial peaking factor.
This census is used incombinationwiththecriticalheatflux(CHF)analysistodetermine how many fuel pins have a radial peaking factor greater than or equal.
to that radial peaking factor which results in a calculated DNBR below that which is assumed to result in DNB.
The pin census became less severe in Cycle 16.
Even though the longer RPS response tire resulted in a lower radial peaking factor leading to DNB, the pin census improved so much that the total number of fuel pins calculated to enter DNB is lower than previously calculated.
On a monthly basis L
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7 the axial and radial power distribution are measured. The measured values are compared to calculated values of axial and radial power distribution to confirm they 6re bounded by the actual vslues. This provides assurance that the codes are still accurately predicting core behavior. Axial offset, nuclear enthalpy rise hot channel and a core reactivity balance are factor, quadrant power tilt ratio,irm that the analysis assumptions factors assured by TSs. which conf and predicted minimum DNBRs and peak fuel centerline temperatures are bounded by the reload analysis during operation.
The peak fuel centerline temperatures for the hot full power and hot zero power cases, assuming the increase in response times, are bounded by the previous analysis.
5)
Uncontrolled Rod Withdrawal from Power for the uncontrolled rod withdrawal from power transient the increase in the everpower trip response time does not impact minimum DNBR since it occurs immediately prior to reactor trip. The RCCA withdrawal event is a relatively slow transient.
In the minimum DNBR analysis, reactor trip occurs at approximately 83 seconds following event initiation.
The rate of temperature rise over this time is slow.
Power is also rising slowly. The core power increases an additional 0.1% of full power due to the increased delay time.
As reactor trip 2260psiain2 seconds)ystempressureisincreasing(2240psiato on high power occurs, s Reactor trip on high pressure is not credited in order to assess the high power and variable low pressure trips.
The rise in pressure compensates for the further increase in core power resulting from the 0.25 second increase in trip delay tine.
Although the predicted peak fuel centerline terrperature increases by 6 F out of 4400 f, the increase is insignificant and does not irapact consequences of the transient, i
3.0 Sut4 MARY The equipment upgrades for both the RPS Phase II and the NIS are acceptable.
In addition the staff has concluded the increase in response times for the power range nuclear flux, start-up rate and low-flow reactor trips do not impact the consequences of any design basis event.
3.1 RPSPhaseIIJpp,radeJonclusions The staff has concluded that the RFS upgrade is acceptable with the exception that qualification to the electrical environment has not been determined.
Care must be taken to assure that there is no common mode EMI/SWC problems which could prevent a reactor trip when required.
It is also irnportant to assure that no inadvertent trip can occur due to EMI/SWC. The staff requires that the installed configuration of the foxboro Spec 200 Hiero equipment be shown to be enveloped by the vendor testing.
The staff requires that the licensee determine the method to be used to verify the electrical environment qualification and document the plan to the staff prior to restart with the RPS Phase II operational, e
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3.2 NIS Upgrade Conclusions The NIS upgrade using the Gammametrics system hat, been previously approved for use at several other plants and is acceptable to the staff for use at Haddam Neck.
3.3 Technical Specification The TS are consistent with the equipment changes, conform to the STS and are acceptable to the staff as shown in the July 28, 1989 letter.
3.4 10 CFR 50.59 Evaluation
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The staff has been aware that the RPS Phase II upgrade would involve the use of Microprocessors since the original audit for Phase I.
The Phase ! SER described the staff conclusion that the RPS upgrade should not have been done under 10 CFR 50.59 because the change from an analog system to a digital microprocessor system has inherent (software) failure modes which present a malfunction of a different type than previously evaluated and is, therefore, an unreviewed safety question.
Since the equipment has been found acceptable and additional guidance to the industry via generic communication is being considered, the staff does not consider this a significant violation. The effects of EMI/SWC may also have a greater impact on the digital systems than on the original analog system.
These conclusions also apply to the RPS Phase II,
4.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32, and 51.3b, an environmental assessment and finding of no significant impact have been prepared and published in the Federal Register on February 23, 1990 (55 FR 6563).
Accordingly, based upon the environmental assessment, we have determined that the issuance of the i
amendment will not have a significant effect on the quality of the human environment.
5.0 C0t<CULS10N We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and(3)theissuance of the amendment will not be inimical to the common defens,e and security or to the health and safety of the public.
Principal Contributors:
J. Stewart A. Wang l
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