ML20136G224
| ML20136G224 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/07/1997 |
| From: | Mccollum W DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20136G228 | List: |
| References | |
| NUDOCS 9703170238 | |
| Download: ML20136G224 (13) | |
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' DukeIbiterCompany MwwR McCowy Jk Catauba Nuclear Generation Department VicePraident 18M ConcordRoad (82)8311@ Ollice s'
%rk. SC73745 (80)83141% Fax DUKEPOWER March 7,1997 U.S.-Nuclear. Regulatory. Commission Attention: -Document Control Desk Washington, D.C.
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Subject:
Catawba Nuclear Station, Units 1 and 2 Dockets Nos. 50-413 and 50-414 Request for Facility Operating License Amendment Steam Generator Tube Rupture Evaluation Pursuant to 10CFR 50.90,. Duke Power Company hereby requests an-amendment to its Facility Operating License Nos. NPF-35 and NPF-52 for Catawba Nuclear Station, Units 1 and 2,
j respectively. The need for the requested license amendment was identified during a
self initiated review of dose.
analysis methodologies, assumptions, inputs and results in
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Chapter 15 of the:UFSAR.
1 Specifically, an amendment
.is requested to Technical
.Spe_cification 3/4.7.1.6 and Selected Licensee Commitments 16.10-1.to ' require four-instead of three steam generator
-PORVs. operable.
Also requested are changes to UFSAR Section Section 15.6.3, Steam Generator Tube Rupture'. This amendment addresses the identification:of a malfunction of:a different type-than: previously evaluated in the -'SAR.
The single failure-of'a-power supply would cause the loss of power.to two st.eam generator power operator relief val _ves (PORVs) and limit the ability to cooldown ~and depressurize the plant.
Upon. discovery of this failure, administrative limits were put into place to bound the consequences of this failure to those previously analyzed for steam generator overfill and offsite dose and an ENS notification was made to the NRC f
under 10 CFR Part 50.72.
Evaluations were started to determine corrective actions which-were entered into
/g Q k Catawba's corrective action program.-
These evaluations included an operability review and a 50.59 ' evaluation which
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concluded that. the plant was ~ operable but degraded and an Unreviewed Safety Question (USQ) existed.
A Licensee Event Report ~ (LER). will be submitted which discusses the. sequence of: events and corrective actions in more detail.
ADOCK 05000413@)
9703170238 970307 PDR P
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t Document Control Desk Page 2 March 7, 1997 The administrative limit requiring four PORVS is the Technical Specification amendment being submitted and a permanent change to the station.
The dose equivalent iodine limit will be in effect until final dose calculations are complete.
Additional information is provided in Attachments 1 and 2 of this letter. contains a description of the
- changes, safety
- analysis, and the determination of no significant hazards.
Attachment 2
contains marked-up Technical Specifications and UFSAR pages depicting the changes and the Environmental Assessment.
It is requested that these changes be approved in an expedited manner to support the upcoming Catawba Unit 2 refueling outage.
The schedule for startup is as follows:
Outage Start Date March 22, 1997 Enter Mode 4 April 22, 1997 0 1600 Enter Mode 3 April 23, 1997 0 0400 Enter Mode 2 April 25, 1997 0 0800 Enter Mode 1 April 25, 1997 0 2200 Any delay in approval of this amendment request would result in the prevention of a resumption of operation for Catawba Unit 2 and subsequent increase in power output up to the unit's licensed power level. Additionally, this situation would also apply to Catawba Unit 1 or Catawba Unit 2 prior to the start of the outage should it trip or have a forced outage.
In the event that this occurs, a request for an accelerated review would be forthcoming.
This proposed change to the Technical Specifications and UFSAR has been evaluated and has been determined to involve no significant hazards considerations. This proposed change and the determination of no significant hazards has been reviewed by the Plant Operations Review Committee (PORC) and the Nuclear Safety Review Board (NSRB).
It has been concluded that implementation of these changes will not result in an undue risk to the health and safety of the public.
In addition, a copy of this amendment request and a copy of the no significant hazards consideration
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Document Control Desk Page 3 March 7, 1997 determination have been provided to the appropriate South Carolina state official.
Should you have any questions regarding this matter, please contact Martha Pursor at (803)831-4015,
,i Very truly yours, C
,/
/
W.R. McCollum, Jr.
Attachments 4
xc (with attachments):
j L.A Reyes, Regional Administrator, Region II R.J.
Freudenberger, Senior Resident Inspector P.S. Tam, Senior Project Manager, ONRR Max Batavia, Chief, Bureau of Radiological Health, SC
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Document Control. Desk Page 4 March 7, 1997 i
W.
R. McCollum, Jr.,
being duly sworn, states that he is Site Vice-President,. Catawba Nuclear Station, Duke Power Company;
.that he is authorized on the part of said company to sign and
-file with the U.S.
Nuclear Regulatory Commission these revisions to the Catawba Nuclear Station License Nos.~NPF-35
[
and NPF 52; and that all statements and matters set forth
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therein are true and correct to the best of his knowledge.
. f. / f W.
R.
McCollum,'Jr., Site ice-President Catawba Nuclear Station i-Subscribed and sworn to before me this 7th day of March, 1997 e
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}
Notary Public
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j' My. commission expires:
4 MYCOMMISSION EXPIRES imAnvm wm
i ATTACHMENT 1 l
DESCRIPTION OF CHANGE (All changes are noted in italics.)
Technical Specifications Revise Technical Specification section 3/4.7.1.6 to show the following change: (This change is applicable to both units)
"Four steam generator power operated relief valves ( PORV ).."
Revise the Technical Specification Bases section 3/4.7.1.6 to show the following change: (This change is applicable to both units)
"..and that at least two are used to cool the Reactor Coolant System inventory to less than the saturation temperature..."
" Local operation of the steam line PORVs is credited in the event tha t remote opera tion is unavailable" UFSAR Revise the indicated portions of UFSAR section 15.6.3.1 as follows:
In paragraph 3,
delete the reference to ( Attachment 5 to Reference 32 on page 15-119).
Under number 2.
in this see. tion insert that a
safety injection signal ia initiated '.'nually or by low pressurizer pressure..
j Under number 7.
in this section insert, " Local operation of the steam line POR Vs is credited in the event that remote operation is unavailable."
Revise the indicated portions of UFSAR section 15.6.3.2 as follows:
Under number 1. of this section, replace the first sentence with, " Reactor trip occurs on manual operator action."
Under number 3.
of this section, note that the referenced figure should be 15-103.
The next to the last paragraph in this section should read, Detailed RETRAN-02 calcula tions are performed to evaluate
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p steam generator' overfill.
The result is that, even when the 1'
steam generator tube rupture event is analyzed with i
assumptions which are conservative with respect to overfill,
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including the most~ limiting single failure, there is margin i
such that overfill is avoided.
The' methodology used for A.
these calculations was prepared by the Westinghouse Owners Group and is documented in Reference 48 on page 15-120."
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The last two sentences in the last paragraph in this section should read, Because of the relative effects on. DNBR ' of the heatup and depressurization allowed by this trip j
- function, the steam generator tube rupture coastdown l
. transient from a lower RCS pressure is bounded by the complete loss of flow transient in Section 15.3.2."
1 l-Under Results in this section the figure numbers are incorrectly referenced. They should read as follows:
Figure 15-103 Break Flow j.
Figure 15-104 Reactor Coolant System Pressure Figure 15-105 Reactor Coolant System Temperature ( For Ruptured Loop)
Figure 15-106 Reactor coolant System temperature (For Intact Loops)
Figure 15-107 Pressurizer Water Level Figure 15-108' Steam Line Pressure i
Figure 15-168 Steam Generator Water Levels 4
Selected Licensee Commitments Revise Selected Licensee Commitments Section 16.10-1 to show l
the following change:
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" Four steam generator PORV safety-related gas supply.."
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]LACKGROUND j
Catawba Nuclear Station currently is engaged in a self initiated review to verify compliance with the UFSAR and accuracy of the UFSAR.
During this
- review, it was determined that Technical Specification (TS) 3/4.7.1.6 is j
not restrictive. enough to ensEre that the consequences of the Steam Generator Tube Rupture (SGTR) accident could be mitigated.
Furthermore, single failures not analyzed for
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effect on.the consequences of the SGTR accident were found.
This discovery was entered into the Catawba corrective s
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action program and an evaluation of operability performed.
During the evaluation it was determined that these single failures were malfunctions of a different type than any
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evaluated in the Safety Analysis Report.
This constitutes an Unreviewed Safety Question (USQ).
s The consequences associated with the SGTR event are overfill of the ruptured steam generator (S/G) and radiological consequences.
Analyses of S/G overfill and radiological
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consequences are reported in References 1 and 2 which define the current licensing basis.
Reference 2 contains separate j
analyses for S/G overfill and radiation doses following SGTR postulated to occur at a three loop reference Westinghouse plant.
In those analyses, the following single failures i
were found to be limiting:
a
- 1) _ Failure of a power operated relief valve (PORV) on an intact S/G to open on demand (limiting for S/G overfill),
and f
- 2) Stuck open PORV on the ruptured S/G (limiting for radiological consequences).
In Reference 1,
it was reported that the overfill analysis bounded Catawba.
It was also determined that the single 4
failures of Reference 2 were the limiting failures for a 1
SGTR accident postulated to occur at Catawba.
Finally, it was reported.that the limiting failure with respect to S/G overfill would leave the affected Catawba unit with PORVs on two intact S/G's available for unit cooldown and termination l
of break flow.
The NRC agreed with the above ' findings in their review of Reference 1 as documented in Reference 5.
Technical Specification 3/4.7.1.6 requires that at least three S/G PORVs be operable for indefinite unit operations in Modes 1,
2, 3,
or 4 (when the S/G's are being used to remove decay heat).
If SGTR and the single failure identified in Reference 2 with respect to S/G overfill is postulated to occur with only tha.ee S/G PORVs initially i
operable, a PORV on only one intact S/G would be available for unit cooldown.
This is contrary to Reference 1.
l A total of six unanalyzed single failures were identified.
Of these six, three were determined to be limiting with respect to S/G overfill.
These single failures recently identified are as follows:
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- 1) failure of power from a Channel A or panelboard of the 120 VAC Vital Instrumentation and Control (I&C)
Power (EPG) System,
- 2) failure of power from an Channel A or D EPG inverter, and i
- 3) failure power from a Channel A or D distribution center of the 125 VDC Vital I&C Power (EPL) System.
With any of these failures, control power is lost to two S/G
- PORVs, rendering them incapable of being operated from within the control room.
Local operation of the affected PORVs (i.e.,
with handwheels) is not affected by these failures.
Superimposition of one of the above failures on the SGTR scenario with three PORVs available would leave the affected unit with only the PORV on the ruptured S/G available for remote operation from the Control Room.
Superimposition of one of these failures on the SGTR event with all four S/G PORVs initially operable leaves the affected unit with a PORV on one intact S/G available for operation from the control room.
As noted above, the affected S/G PORVs are still available for local manual j
operation.
The following administrative restrictions have been put in place:
- 1) all four PORVs be maintained operable with action times comparable to the existing specification (Technical Specification 3/4.7.1.6), and
- 2) *I dose equivalent concentration be restricted to a conservatively low value.
This action ensures that the i
latest dose analysis of record is bounding for Catawba Nuclear Station operation.
Restriction 1), along with local manual operation of one of the S/G PORVs affected by one of the single failures described above, will ensure that adequate margin to S/G overfill exists.
Restrictions 1) and 2) ensure that the offsite doses following a SGTR with one of these single failures remain within the values presented in the current dose analysis of record.(Reference 4)
It is expected that Restriction 2) will be lifted upon completion of dose analyses based on more detailed input in place of the conservative assumptions made to support the administrative controls.
T_ECHNICAL JUSTIFICATION AND SAFETY ANALYSIS Scoping evaluations were performed for the two cases outlined below:
1)
All four S/G PORVs initially are operable and therefore are available for remote operation from the Control Room at the initiation of the SGTR, single failure assumed to be one S/G PORV failing to open from the control room.
Cooldown is accomplished by remotely using two PORVs on intact S/Gs.
2)
All four S/G PORVs initially are operable and therefore are available for remote operation from the Control Room at the initiation of the SGTR, single failure assumed to be failure of control power to two S/G PORVs.
Cooldown is accomplished by remote operation from the Control Room using a PORV on one intact S/G.
In Case 2 above, it was demonstrated that S/G overfill occurs.
Case 1 above is enveloped by the following case, for which detailed RETRAN-02 computer code analyses were performed:
3)
All four S/G PORVs initially are operable and therefore are available for remote operation from the Control Room at the initiation of the SGTR, single failure assumed to be as outlined above (i.e.,
Vital I&C Power to S/G PORVs), leaving only one PORV on an intact S/G available to be remotely operated from the control room.
Local manual operation of another PORV on an intact S/G with a time delay is credited in this analysis.
In Case 3 above, it was demonstrated that there is adequate j
margin to S/G overfill. Therefore, to avoid S/G overfill in light of the single failures identified herein, credit must be assumed for local manual operation of a PORV on at least one intact S/G.
Limited manual action to effect plant cooldown is allowed under the Standard Review Plan (Reference 3).
This action needs to be taken only for a SGTR with a LOOP and one of the control power failures outlined above.
Furthermore, the action involves j
dispatching an operator to only one location to operate the handwheel on one S/G PORV.
Both time trials and simulator tests have been performed to show that operators can begin local manual operation of a S/G PORV and otherwise respond
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as assumed in the safety analysis.
Additionally, the emergency operating procedure which is used to respond to a SGTR accident currently includes directions to dispatch an operator to manually operate S/G PORVs to effect plant s
- cooldown, No procedure changes' had to be made to accommodate case 3 above.
i Administrative restrictions currently in place, based on l
conservative assumptions, will ensure that the resultant doses from a postulated SGTR will not exceed the last dose analysis of record (Reference 4).
These administrative controls are expected to be lifted upon completion of dose i
analyses based on more detailed input in place of the conservative assumptions made to support the restrictions.
The requirement to maintain all four PORVs operable rather than the current requirement for three PORVs is more restrictive, and ensures Catawba Nuclear is capable of l
mitigating a SGTR accident, along with the single failure described
- above, while also avoiding S/G overfill.
Alternatives to this proposed Technical Specification i
amendment will involve costly plant modifications which may either
~ require additional Technical Specifications amendments, or involve unreviewed safety questions, 4
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DE_ TERMINATION OF NO SIGNIFICANT HAZARDS This proposed change has been evaluated against the standards in 10 CFR 50.92 and has been determined to involve no significant hazards, in that operation of the facility in accordance with the proposed amendment would not:
1)
Involve a significant increase in the probability or consequences of an accident previously evaluated:
The proposed Technical Specification amendment ensures that the consequences of a postulated SGTR accident are enveloped by current analyses.
The proposed Technical Specification amendment, together with credit for local manual operation of one S/G PORV, will ensure that adequate margin to overfill exists for the SGTR accident.
Furthermore, with administrative controls currently in place regarding reactor coolant specific activity, this requirement ensures that offsite doses following the SGTR accident remain within the dose analysis of record.
These administrative controls are expected to be lifted with the completion of dose analyses based on more detailed input in place of the conservative assumptions made to support the restrictions.
The requirement to maintain all four S/G PORVs operable is more restrictive than the current requirement, and therefore does not adversely affect the consequences of any analyzed accident.
The accident in which the S/G PORVs are considered to be accident initiators is discussed in Section 15.1.4 of the Catawba UFSAR.
Considering the number, design features and reliability of steam dump to condenser valves (nine), atmospheric dump valves (nine), S/G Code Safety Valves (twenty),
and S/G PORVs (four),
the requirement to maintain all four S/G PORVs operable does not significantly increase the probability of inadvertent opening of steam dump valve as analyzed in Section 15.1.4 of the Catawba UFSAR.
As reported in Section 15.1.4 of the Catawba
- UFSAR, inadvertent opening of a S/G PORV is enveloped by the consequences of a postulated Main Steam Line Break.
The requirement to maintain all four S/G PORVs operable does not in any way change this.
l 2)
Create the possibility of a new or difforent kind of l
accident from any previously evaluated:
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No new accident types have been identified for the S/G PORVs or any SSCs associated with or connected to the S/G PORVs.
With respect to the types of accidents that should be considered, the Standard Review Plan and the l
Catawba UFSAR are considered to be complete for Catawba Nuclear Station.
3)
Involve a
significant reduction in the margin of safe ty:
Margin of safety is related to confidence in fission product barriers.
The proposed Technical Specification amendment, along with credit for local manual operation of one S/G PORV, will ensure that there is adequate margin to overfill.
Therefore, the steam lines, S/G PORVs and the code safety relief valves will not be degraded following a design basis SGTR.
This amendment will also ensure that steaming of the ruptured S/G is not necessary to effect plant cooldown after a
postulated SGTR.
Along with administrative controls currently in place regarding reactor coolant specific activity, this requirement ensures that offsite doses following SGTR remain within values of the dose analysis of record. These administrative controls are expected to be lifted with the completion of dose analyses based on more detailed input in place of the conservative assumptions made to support the restrictions.
In summary, this proposed amendment does not involve a significant reduction in the margin of safety.
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e ATTACHMENT 2 l
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