ML20036A454
| ML20036A454 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 05/04/1993 |
| From: | Opeka J CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST UTILITIES |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20036A455 | List: |
| References | |
| B14461, NUDOCS 9305110240 | |
| Download: ML20036A454 (6) | |
Text
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$0RTHEAST UTILETIES cener.i Omce.. seinen street. Bertin Connecticut 9
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May 4, 1993 Docket No. 50-213 B14461 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Gentlemen:
Haddam Neck Plant Proposed Revision to Technical Specifications Additional Atmospheric Steam Dumo Pursuant to 10CFR50.90,. Connecticut Yankee Atomic Power Company (CYAPCO) hereby proposes to amend Facility Operating License DPR-61 by incorporating the changes identified in the Attachments into the technical specifications of the Haddam Neck Plant.
Description of the Proposed Chanaes During the upcoming Cycle 17 refueling outage, four main steam safety valves (one per main steamline) are being replaced with valves that will significantly increase the remote manual atmospheric steam dump capability of the Haddam Neck Plant.
These new valves can be operated remotely from the main control room as well as locally (self-actuating) for steam generator overpressure protection.
This modification will provide the ability to rapidly cool down and depressurize the plant without reliance on the. main
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condenser or other nonsafety-related balance of plant equipment.
This method of cooldown would be used following the rupture of a steam generator tube concurrent with a loss of offsite power or loss of the main condenser.
The present atmospheric ' steam dump capacity at the Haddam Neck Plant is limited and was originally designed to maintain ~ the plant at hot shutdown; following a reactor trip.
Therefore, procedures currently require reducing the primary side pressure to _ equal _ the secondary ~ side pressure if: offsite power.and the main condenser are available to terminate leakage which results from a steam generator' tube rupture (SGTR).
The reactor coolant system (RCS) loop stop valves can also be closed to terminate primary to secondary leakage
- due to a SGTR with a loss of offsite power or loss of main condenser.
With-the installation of the new type of code. safety valve, a SGTR may_ also be 100052-s i
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U.S. Nuclear Regulatory Commission B14461/Page 2 May 4, 1993 mitigated by dumping steam from two of the intact steam generators and rapidly cooling and depressurizing the primary system until the RCS pressure is below the lowest main steam code safety valve setpoint.
It should be noted that a SGTR is not part of the original licensing basis of the Haddam Neck Plant and that the loop stop valves were not originally considered safety grade.
Safety Assessment The proposed license amendment will make the following changes to the technical specifications:
- 1) This package will delete the operability requirements for the RCS loop stop valves in Section 3.4.1.1.
Associated with this is the deletion of the RCS loop stop valve surveillances in Section 4.4.1.1 and the deletion of the RCS loop stop valves from Bases Section 3/4.4.1; 2) This package modifies Section 3.7.1.1 and renumbers it as 3.7.1.1.1 to pertain to the self-actuating function of the steamline code safety valves, including the new dual function valves being installed.
- Also, the ACTION statement was changed to require the final plant mode to be HOT SHUTDOWN insteao of COLD SHUTDOWN.
Additional editorial changes were also made to the Section; 3) The package adds Section 3.7.1.1.2 which pertains to the operability of the remote actuation function of new steamline code safety valves that are being installed during this refueling outage.
Deleting the operability requirement of the RCS loop stop valves cannot affect the probability of occurrence of a previously evaluated accident.
The potential impact on the probability of occurrence of a previously evaluated accident reflected in this proposed license amendment is for the Excess Load Increase event.
The Excess Load Increase event includes spurious opening of safety valves.
The new valves will have the capability to be opened from the control room which means that a new method of having a safety valve open is added by the change.
The remote opening of the valve requires DC power and utilizes " energize-to-open" solenoid valves that fail closed on loss of DC power. The probability of a valve spuriously opening either due to mechanical reasons or a hot short is not expected to be much different for the new valves from the probability of spurious opening of the existing valves.
Also, the spurious opening of a safety valve is not the only means of causing an Excess load Increase event.
Based on the above, the change is judged to have a negligible impact on the probability of an Excess Load increase event.
The new safety / relief valves are being installed for use in mitigating an
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SGTR, without the main condenser available, and without use of the RCS loop stop valves. The design basis SGTR analysis assumes that primary-to-secondary leakage is terminated within 30 minutes from the time of the tube rupture.
This assumption is not being changed.
Therefore, there is no impact on the design basis consequences of an SGTR. The new valves would allow for a faster RCS cooldown and therefore RCS depressurization.
However, when using the RCS loop stop valves to terminate the primary-to-secondary leakage, the required RCS cooldown and depressurization is smaller. The impact of these differences is judged to negligibly impact the actual consequences of an SGTR.
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U.S. Nuclear Regulatory Commission B14461/Page 3 May 4, 1993 Consequences are also calculated for some accidents as a result of steam releases to the environment through the safety valves.
The most limiting steam generator over--pressurization transient, the Loss of Load event, does not predict lifting the fourth safety valve on any steam line.
The fourth safety valve has the highest lift pressure setpoin; of 1034 psia.
This means that replacing the safety valve with a safety / relief valve.cannot impact the consequences of this accident.
In addition, the new valves have essentially the same relief capacity as the existing valves. The new safety / relief valves will have the same opening set pressure requirements as the existing valves.
The safety valve blowdown is also expected to be similar to the current valves and will not exceed the maximum blowdown assumed in the safety analysis.
This proposed modification cannot introduce an accident of a different type since a safety valve opening spuriously is already considered in the Excess Load Increase design basis event and deleting the operability requirement for the RCS loop stop valves cannot initiate an event of a different type.
The change in the final mode from. C0LD SHUTDOWN to,H0T SHUTDOWN in.Section 3.7.1.1 potentially can increase the time the plant could be in Mode 3 with an inoperable safety valve (s).
This is due to the fact that it takes longer to.-
get to Mode 5, COLD SHUTDOWN, which means that the plant could exit Mode 3 sooner with the existing ACTION statement.
The likelihood for a new or-different accident requires a safety valve (s) to be inoperable, the potential i
time increase in Mode. 3, and an accident to occur that cannot be mitigated without exceeding design basis criteria with the safety valve (s) inoperable.
i The probability of. this scenario is judged too-low to represent a new or different accident.
The proposed uses of the new valves are to provide a more rapid plant cooldown and to mitigate an SGTR with the condenser unavailable.
This use of the valves will not result in a new type of accident.
The new safety / relief valves have essentially the same relief capacity and opening stroke time as the existing valve and will have the same lift pressure requirements.
In addition, the fourth safety valve'on each steam line is not predicted to lift for the limiting peak steam generator pressure event which is Loss of Load.
The change installs valves for use by the. operator in mitigating the consequences of an SGTR with loss of offsite power.
The SGTR analyses assume isolation of primary-to-secondary leakage within 30 minutes of the rupture.
The new valves do not require changing this assumption.
Therefore, the change does not require reanalysis of-the design basis accident.
Based on the above, there is no reduction in the margin of safety.
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'l U.S. Nuclear Regulatory Commission:
'i B14461/Page 4 q
May 4, 1993' l
Sianificant Hazards Consideration CYAPCO-has reviewed the proposed change in accordance with 10CFR50.92 and has-j concluded that 'the change. does not involve
'a significant hazards i
consideration.
The bases for this conclusion is that the three criteria Jof.
l 10CFR50.92(c) are not compromised.
The proposed change does not involve a significant hazards consideration because the change would not:
{
1.
Involve a significant increase in the probability of_ occurrence or consequences of. an accident previously analyzed.
1 The proposed changes reflect the replacement of one _ steam generator safety valve per steam generator with a' safety / relief: valve.
The replacement valve requires DC power and " energize-to-open" solenoid r
valves.
The probability of a spurious opening of a valve, mechanically =
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or due to hot shorts, is not expected to be significantly different than.
1 that _of the existing valves. The likelihood of excess load ~ transients is-not significantly increased by the change.
The' assumptions used to calculate the dose consequences for SGTR analyses are also not changed by l
this modification.
Therefore, the consequences of accidents previously' analyzed are not increased.
l 2.
Create the possibility of a new or different. kind of accident.
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The change modifies the equipment-credited; for mitigating an. SGTR.with 4
the condenser unavailable.
This event is. not part ' of.. the original licensing basis of the' Haddam Neck Plant.. Currently,lthe RCS loop stop valves are not ~ designed for single -failures.
The failure probability of the new safety / relief valves is judged to' be very low.
Their failure in :
the mitigation of a SGTR.is Judged as highly unlikely, and believed to be-more reliable than the RCS loop stop valves. _ This means that a new or -
4 different kind of accident is not introduced'by the change.
i 3.
Involve a significant reduction in the margin of safety.
The replacement valves have essentially the same capacity and stroke time j
as the existing valves. The. lift setpoint is not being changed. The new l
valves do not require changing the assumptions for the design basis SGTR accident' analyses. Therefore, the margin of safety is not reduced.
j Moreover, the Commissien has-provided. guidance concerning the application of the standards in 10CFR50 92 by providing certain examples (March 6,
- 1986, 51FR7751) of amendments that-are considered not-likely to involve - a.
significant hazards consider ttion. Although'the change proposed herein is~ not
' enveloped by _a-specific ezample, it 'is considered safe. and is ?not-a
,j
- significant hazards. consideratian since these valves replace 'similar valves.
The major change is that these ;new valves will-be' capable:of being; controlled from the. control room as' well as-maintaining self-actuation.
Therefore, the
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1 U.S. fluclear Regulatory Commission j
B14461/Page 5 May 4, 1993 i
proposed technical specifications reflect both the self-actuated and remote actuation factors.
Further, the addition of the new remote steam dump capacity provides an additional method for mitigating the consequences of a SGTR.
Emergency procedures formerly required that the loop stop MOVs be closed to terminate leakage resulting from a steam generator tube rupture.
With the installation of the new type of safety valve, a steam generator tube rupture may be mitigated by dumping steam from the intact steam generators and rapidly I
cooling and depressurizing the primary system until RCS pressure is below the lowest main steam safety valve setpoint, thereby terminating the release through the open safety valve (s) on the faulted generator.
CYAPC0 has reviewed the proposed licensed amendment against the criteria of 10CFR51.22 for environmental considerations.
The proposed changes do not involve a significant hazards consideration, nor increase the types and amounts of effluents that may be released offsite, nor significantly increase individual or cumulative occupational radiation exposures.
Based on the foregoing, CYAPC0 concludes that the proposed changes meet the criteria delineated in 10CFR50.22(c)(9) for a
categorical exclusion from the requirements for an environmental impact statement.
Attachment I
presents the marked-up copy of the affected Technical Specifications.
The retype of the proposed changes to Technical i
Specifications are included in Attachment 2 and reflect the currently issued version of Technical Specifications. Pending Technical Specification changes,
-l or proposed Technical Specification changes issued subsequent to this submittal are not reflected in the enclosed retype.
The enclosed retype should be checked for continuity with Technical Specifications prior to issuance.
t Revision bars are provided in the right hand margin to indicate a revision to the text.
fio revision bars are utilized when the page is changed solely to 1
accommodate the shifting of text due to additions or deletions.
1 The Haddam tieck Plant fluclear Review Board has reviewed and approved the proposed changes and has concurred with the above determination.
In accordance with 10CFR50.91(b), we are providing the State of Connecticut
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with a copy of this proposed amendment.
i Regarding our proposed schedule for this amendment, we request issuance at your earliest convenience. The effective date of the amendment should be upon issuance for implementation within 30 days following issuance.
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F U.S. Nuclear Regulatory Commission B14461/Page 6 May 4, 1993 We trust you will find this information satisfactory and remain available to discuss.this with you at your convenience.
Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY
. F. CW.
J. F. $pe'ka O
Executive Vice President cc:
T. T. Martin, Region I Administrator A. B. Wang, NRC Project Manager, Haddam Neck Plant W. J. Raymond, Senior Resident Inspector, Haddam Neck Plant l
Subscribed and sworn to before me this //d day of~hwr
, 1993 LM/sb NM' #
Nptgry Public Date Commission Expires:
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