View source for W3P87-0537, Application for Amends to License NPF-38,raising Emergency Feedwater Initiation Low Level Setpoint from 30% Wide Range Level to 36.3% & Increasing Calibr Frequency for Control Room Chlorine Detection Sys.Fee Paid

{{Adams
| number = ML20211G994
| issue date = 02/23/1987
| title = Application for Amends to License NPF-38,raising Emergency Feedwater Initiation Low Level Setpoint from 30% Wide Range Level to 36.3% & Increasing Calibr Frequency for Control Room Chlorine Detection Sys.Fee Paid
| author name = Dewease J
| author affiliation = LOUISIANA POWER &amp; LIGHT CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000382
| license number = 
| contact person = 
| document report number = A00L-870223, AL-870223, W3P87-0537, W3P87-537, NUDOCS 8702260034
| package number = ML20211G999
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 8
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:LOUISIANA , 317BARONNESTREET                . P. O. BOX 60340 POWER & LIGHT /        NEW ORLEANS, LOUISIANA 70160      .    (504) 595-2781
$LES EYS E February 23, 1987                           J. G. DEWE ASE SENOR VKE 9RE5!OENT.
NUGEAR OPERATON$
W3P87-0537 A4.05 QA U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555

==Subject:==
Waterford SES Unit 3 Docket No. 50-382 Technical Specification Change Requests NPF-38-53, NPF-38-54, NPF-38-55, NPF-38-56, NH'-38-57 Gentlemen:
Louisiana Power & Light hereby files an application for five amendments to the Waterford 3 Technical Specifications. The amendments involve:
: 1. Raising the emergency feedwater initiation low level setpoint from 30%
wide range level to 36.3% (NPF-38-53),
: 2. Increasing the calibration frequency for the control room chlorine detection system (NPF-38-54),
: 3. Raising the required refueling water storage pool level from 82% to 83% (NPF-38-55),
: 4. Raising the required safety injection tank level from 60% to 61%
(NPF-38-56), and
: 5. Changing the function of two reactor auxiliary building smoke detectors from notification only to notification and actuation (NPF-38-57).
Ng22g8gy 738           2 P
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                      "AN EQUAL OPPORTUNITY EMPLOYER" Qee") wfcIL WWD

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I                    W3P87-0537
                    -. Page 2                                                                                                                                                                                *
                    . The proposed changes do no't involve an unreviewed safety question b r a significant hazards consideration. Should you have any questions or require additional information concerning the requested changes, please i                     contact Mike Meisner at (504) 595-2832.                                                                                                                                                                      ,
2                                                                                                                                 Yours very truly,                                                                      -
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f'.d.Dewease Eenior Vice President Nuclear Operations JGD/MJM/ssf r

==Enclosures:==
NPF-38-53 NPF-38-54 NPF-38-55 NPF-38-56 NPF-38-57 Filing fee, LP&L check - $150.00 5                      cc:    E.L. Blake W.M. Stevenson G.W. Knighton
:                            R.D. Martin
!                            J.H. Wilson NRC Resident Inspectors Office Administrator Nuclear Energy Division (State of Louisiana)
American Nuclear Insurers l
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
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In the matter of                                       )
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Louisiana Power & Light Company                        ) Docket No. 50-382 Waterford 3 Steam Electric Station                     )
AFFIDAVIT J.G. Dewease, being duly sworn, hereby deposes and says that he is Senior Vice President-Nuclear Operations of Louisiana Power & Light Company; that he is duly authorized to sign and: file with the Nuclear Regulatory Commission the attached Technical Specification Change Requests; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.
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8 nior Vice President-Nuclear Operations STATE OF LOUISIANA)
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l Subscribed and sworn to before me, a Notary Public in arf for the Parish I and State above named this      M f n / day of        fa ff7// A         > ,
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4 DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-53 This~is a request to revise Table 3.3-4, Engineered Safety Features Actuation System Instrumentation Trip Values of the Technical Specifications for Waterford Steam Electric Station Unit 3.
Existina Specification
      .See Attachment A.
Proposed Specification See Attachment B.
Description This change is requested to eliminate an unnecessary value in the Technical Specifications. Table 3.3-4 identifies trip values for the EFW Control Valve Logic Lo Level. Two levels are given for this parameter; 36.3% wide range with-out a safety injection actuation signal (SIAS) or 30.0% with a SIAS signal. The operability for this system is required for Modes 1 through 3 as protection against design basis events such as a Small Break LOCA, Steam Generator Tube Rupture or a Steam Line or Feedwater Line Break. (The 36.3% setpoint is necessary only for the Large Feedwater Line Break event.) In order to simplify the Technical Specifications the requested change will redefine the 30.0% wide range setpoint to 36.3% removing the dependence upon the SIAS.
Currently, under automatic control in the absence of a SIAS, the EFW control valve positions are determined based on steam generator level and EFW flow (see FSAR Section 7.3.1.1.6). For instance, at the critical level, 55.0% wide range, one control valve opens to allow 200 gpm to the steam generators, and the other valve is controlled by the flowmeter to ensure that at least 175 gpm is delivered should the first valve fail to function properly. At the Lo Level, 45% wide range, the flow controller setpoint increases to 400 gpm. However, with a SIAS, the control valves are no longer controlled by flow, but rather.by steam generator level, only. In this level mode, when steam generator level falls below 68% (and a SIAS is present) the EFW control valves are modulated to drive the level back to 68%. The rate at which the valves open and close is a function of the difference between the setpoint (68%) and the actual level, and the length of time that difference exists. For example, with SIAS present and steam generator level at the critical level, 55.0% wide range, the EFW control valves should be full open.
If the steam generator level drops to the "Lo-Lo" Level (30.0% with a SIAS, 36.3%
without a SIAS), a priority open signal is sent to the valves which overrides all automatic or manual controls. Once the level raises above the applicable value, the control returns to its previous mode; either " automatic" as summarized above, or " manual" under operator control.

The proposed change will increase the 30.0% setpoint to 36.3% resulting in one Lo Level setpoint, thereby simplifying the Technical Specifications. Because the setpoint increase is in the conservative direction, additional safety margin will be provided for the design basis events of concern with the exception of the Large Feedwater Line Break which will be unaffected by the change.
Safety Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:
: 1. Will the operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?
Response: No.
Raising the EFW control valve logic Lo Level setpoint has no effect on the probability of occurrence of the initiating event itself. The setpoint comes into play only when an analyzed event progresses to the point of reducing the steam generator level to the setpoint value. By raising the setpoint, feedwater will be available earlier than previously analyzed for most events, thus mitigating the consequences of the event.
: 2. Will operation of the facility in accordance with this propos.!d change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
Raising the EFW trip setpoint in a conservative direction does not create any new failure or accident path. Therefore, there is no possibility of occurrence for a new or different kind of accident from any accident previously evaluated.
i
: 3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
Response: No.
By raising the EFW setpoint, feedwater flow will be provided earlier than assumed in the analysis for most events, resulting in an additional margin
!         of safety for those events.
l l   The Commission has provided guidance concerning the application of standards for j   determining whether a significant hazards consideration exists by providing l   certain examples (48 CFR 14870) of amendments that are considered not likely to
,   involve significant hazards considerations. Example (ii) relates to a change i   that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications; for example, a more stringent sur-veillance requirement.
l

l In this case, the proposed change is similar to Example (ii) since.the EFW setpoint increase provides a more conservative operating range.
Safety and Significant Hazards Determination Based upon the above Safety Analysis, it is concluded that (1) the proposed change does not constitute a significant hazards consideration as defined by 10 CFR 50.92; (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental State-ment.
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