ML20216G312
| ML20216G312 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 09/11/1997 |
| From: | Ewing E ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| W3F1-97-0227, W3F1-97-227, NUDOCS 9709150048 | |
| Download: ML20216G312 (5) | |
Text
- - -.
'+ g-Ente gy Oper:llons,Inc.
)
D/
Kmona. LA 70006 Tel 504 739 6242 Early C. Ewing,111 t
tety a rugutavy Mars W3F1-97-0227 -
A4.05 PR September 11,1997 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 Request For Additional Informatica Concerning Technical Specification Change Request NPF-38-199 Gentlemen:
During an August 27,1997, telephone conversation the NRC staff was provided verbal responses to questions concerning proposed changes to Technical Specification (TS) Tables 3.7-1 and 3.7-2, and the associated TS Bases. At the end of the conversation Waterford 3 personnel were asked to provide written responses.
The questions and associated responses are attached.
Should you have any questions or comments concerning this information, please contact Paul Caropino at (504) 739-6692.
Very truly yours, l
~
E.C. wing Director Nuclear Safety and Regulatory Affairs ECE/WMW/ssf Attachments:
Affidavit Questions and Responses 9709150048 9769'11
~
PDR ADOCK 05000382
, ff!
lllll
4 4
4 l
Request for Additional Information Concerning Technical Specification Change Reliuost NPF-38-199 W3F197-0227 Page 2 September 11,1997 cc:
E.W. Merschoff (NRC Region IV)
C.P. Patel (NRC-NRR)
J. Smith N.S. Reynolds NRC Resident inspectors Office Administrator Radiation Protection Division (State of Louisiana)
American Nuclear Insurers
UNITED STATES OF AMERICA NUCLEAR. REGULATORY COMMISSION In the matter of
)
)
Entergy Operations, incorporated
)
Docket No. 50-382 Waterford 3 Steam Electric Station
)
AFFIDAVIT Early Cunningham Ewing, being duly sworn, hereby deposes and says that he is Director, Nuclear Safety & Regulatory Affairs - Waterford 3 of Entergy Operations, Incorporated; that he is duly authorized to sign ard file with the Nuclear Regulatory Commission the attached Request for Additional Information Concerning Technicai Specification Change Request NPF-38-199; 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.
lA
-tarly C' Ewing Director, Nuclear Safety and Regulatory Affairs STATE OF LOUISIANA
)
) ss PARISH OF ST. CHARLES
)
Subscribed and sworn to before me, a Notary Public in and for the Parish and State above named this
~c day of
.4<>.
4
,1997.
s%
b h
<uj_6 Notary Public My Commission expires ~/ id
e QUESTIONS AND RESPONSES PROPOSED CHANGE NPF-38-199 Question #1:
How does Waterford 3 verify the acceptability of the Linear Power Levei-High trip setpoints listed in Technical Specifications Table 3.7-2, " Maximum Allowable Linear Power Level-High Trip Setpoint with inoperable Steam Line Safety Valves During Operation with Both Steam Generator 37' Answer #1:
NRC Information Notice (IN) 94-60, Potential Overpressurization of Main Steam System, discusses the inadequacy of the linear function of the main steam safety valve capacity which determines the basis for the Linear Power Level High trip setpoints listed in Technical Specifications Table 3.7-2. In Attachment 1 to IN 94-60, Westinghouse offered three recommendations which could be taken by plants to address the issue. At that time, Waterford 3 followed recommendation #2 and re-analyzed the Loss of Condenser Vacuum (LOCV) event (limiting event for secondary pressure) with the initial power and safety valve conditions dictated by Technical Specification Table 3.7-2. Since that time, the acceptability of the Linear Power Level High trip setpoint has and will be verified through the performance of the LOCV safety analysic for Waterford 3.
Question #2:
The Technical Specification Change Request states that the pressure drop from the steam generator outlet up to the Main Steam Safety Valve (MSSV) inlet is 72 psia. During the August 27,1997, telephone conversation it was stated that the pressure drop in the main steam line is less than 11.5 psia. What is the basis for these numbers and how do they affect the steam generator pressure during a LOCV event?
Answer #2:
The pressure drop which occurs from the steam generator, through the MSSV spool piece, and up to the MSSV during full flow conditions with all of the MSSVs full open is no greater than 72 psia. However, most of this pressure drop occurs at the 90 degree turn into the MSSV spool piece once the valve is open. A much smaller pressure drop exists in the main steam line itself that will affect the opening setpoints of the MSSVs. When the LOCV occurs, a turbine trip causes the steam flow rate to rapidly decrease to zero. Therefore the pressure drop in the steam line is zero and the first MSSV will open at a steam generator pressure equal to the MSSV setpoint. Once this valvo opens, flow is re-established in the steam line. The blowdown setting of the MSSV is sufficient to offset the total pressure drop as the valve opens and maintains the valve open.
c e
Upon approach to the setpoint of the second MSSV, a pressure drop due to the flow in the main steam line (approximately 1/6 rated main steam flow) will cause the' steam generator pressure to be 0.5 psia greater than the pressure at the second MSSV Note that since there is no flow through the spool piece leading up to the second MSSV when the valve is closed, the only pressure drop which needs to be considered with respect to the opening setpoint occurs in the main steam line. As more valves begin to open, the steam flow, and therefore the pressure drop, in the main steam line will increase to a maximum pressure difference of 11.5 psia between the steam generator and the opening setpoint of the sixth (and last) MSSV. The table below lists the nominal lift setpoint for the MSSVs and the steam generator pressure (lift setpoint used in analysis) at which the valve is assumed to open in the analysis.
Nominal Lift Setpoint Lift Setpoint +3%
Lift Setpoint to be Used (psla)
Tolerance in Analysis (psia)
(psia)
[a Above Lift Setpoint]
1084.7 1117.2 1117.2 (0) 1099.7 1132.7 1133.2 (0.5]
1114.7 1148.1 1149.9 (1.8]
1129.7 1163.6 1167.7 (4.11 1139.7 1173.9 1181.3 (7.4]
1149.7 1184.2 1195.6 (11.5]
l 1
_