ML20095F983
| ML20095F983 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/11/1995 |
| From: | Hebdon F NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20095F988 | List: |
| References | |
| NUDOCS 9512190403 | |
| Download: ML20095F983 (11) | |
Text
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UNITE 3 STATES j
NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 2000H001 "s.,...../
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT. UNIT 1 AMEN 0 MENT TO FACILITY OPERATING LICENSE i
i i
Amendment No. 216 License No. DPR-77 1.
The Nuclear Regulatory Commission i 3 Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated August 7, 1995, complies witt, the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be.
conducted in compliance with the Commission's regulations D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
9512190403 951211 PDR ADOCK 05000327 P
l 2.
Accordingly, the license is amended by changes to the Technical.
Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-77 is hereby 1
amended to read as fo11cws-2 (2) Technical Specifications j
The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 216, are hereby incorporated in the i
license. The licensee shall operate the facility in accordance with the Technical Specifications.
1 3.
This license amendment is effective as of its date of issuance, to oe implemented within 45 days.
FOR THE NUCLEAR REGULATORY COMMISSION FrederickJ.Hebdh, Director
-Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications i
Date of Issuance: December 11, 1995 l
1
-. _. ~.... - - - -.
)
i i
ATTACHMENT TO LICENSE AMENDMENT NO. 216 FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT 3/4 2-7 3/4 2-7 6-21 6-21 i
l i
l
~
j o
+
t
+
POWER DISTRIBUTION LIMITS y
SURVEILLANCE REQUIREMENTS (Continued) e.
With measurements indicating l
1
'Ff(z) maximum over z
,(,}
has increased since the previous determination of Fn(z) either of the following actions shall be taken:
M 1.
P (z) shall be increased over that specified in 4.2.2.2.c by the
'j g
appropriate factor specified in the COLR, or M
2.
F (z) shall be measured at least once per 7 effective full power days until 2. successive maps indicate that
'Ff(z)'
maximum over 2 is not increasing.
,(,y f.
With the relationships specified in 4.2.2.2.c above not being satisfied:
1.
Calculate the percent F (z) exceeds its limit by the following expression:
1 l
I
'Ff(z) x W( z) '
l maximum over z
-1 x 100 for P at 0.5 lgggy x K(z) 1 i
l
'Ff(z) x W(z)'
maximum over z
-1 x 100 for P 4 0. 5 lgggy j
0.5 t
i 2.
r.ither of the following actions shall be taken:
a.
Plttee the core in an equilibriv.m condition where the limit in 4.'2.2.2.c is satisfied. Power level may then be increased l
provided the AFD limits of Specification 3.2.1 are reduced 1%
lR159 AFD for each percent F (z) exceeded its limit, or b.
Comply with the requirements of Specification 3.2.2 for F (z) exceeding its limit by the percent calculated above.
R144 i
SEQUOYAH - UNIT 1 3/4 2-7 Amendment Nos. 19, 95, 140, 155,216
~
ADMINISTRATIVE CONTROLS 14_QNTHLY REACTOR OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, l
including documentation of all challenges to the PORVs or Safety Valves, shall R76 be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.
CORE OPERATING LIMITS REPORT R159 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:
1.
Moderator Temperature Coefficient LOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3, 2.
Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, 3.
Control Bank Insertion Limits for Specification 3/4.1.3.6, 4.
Axial Flux Difference Limits for Specification 3/4.2.1, 1
5.
Heat Flux Hot Channel Factor, K(z), W (z), and the factor that accounts for the potential. decrease in Fo margin between surveillances for Specification 3/4.2.2, and R159 6.
Nuclear Enthalpy Hot Channel Factor and Power Factor Multiplier for Specifiention 3/4.2.3.
6.9.1.14.a The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in:
1.
WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION i.ETHODOLOGY",
July 1985 Qi Proprietary).
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 -
Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Hot Channel Factor.)
2.
WCAP-10216-P-A, Revision 1A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL Fn SURVEILLANCE TECHNICAL SPECIFICATION", FEBRUARY 1994 (W Proprietary).
(Methodology for Specification 3.2.1 - Axial Flux Difference R159 (Relaxed Axial Offset Control; and 3.2.2 - Heat Flux Hot Channel Factor (W(z) surveillance requiremeats for En Methodology).)
3.
WCAP-10266-P-A, Rev.
2, "THE 1981 REVISION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987, Qi Troprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).
4.
WCAP-13631-P-A, " SAFETY EVALUATION SUPPORTING A MORE NEGATIVE EOL R175 MODERATOR TEMPERATURE COEFFICIENT TECHNICAL SPECIFICATION FOR THE SEQUOYAH NUCLEAR PLANTS,* MARCH 1993 (H Proprietary).
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient)
SEQUOYAH - UNIT 1 6-21 Amendment No. 52, 58, 72, 74, 117, 152, 155, 156, 171,216
i
+RCEtv UNITED STATES j
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j NUCLEAR REGULATORY COMMISSION j
WASHINGTON. D.C. 30586 4 001 l
\\*****/
1 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.206 License No. DPR-79 j
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated August 7, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the i
Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
\\
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-79 is hereby amended to read as follows:
1 (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.206, are hereby incorporated in the l
license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance, to be l
implemented within 45 days.
]
FOR THE NUCLEAR REGULATORY COMISSION he
) l bkr r
Frederick J. HeMon, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications i
Date of Issuance: December 11, 1995 i
1
ATTACMENT TO LICENSE AMENOMENT NO. 206 FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 1
Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines
-indicating the area of change.
REMOVE INSERT l
3/4 2-6 3/4 2-6 1
6-22 6-22 6-22a 6-22a j
i i
~
l
?
l l
i l
POWER DISTRIBUTION LIMITS SURVEILLANO REQUIREMENTS (Continued)
R21 i
e.
With measurements indicating
'Ff(z) maximum over z y(,)
has increased since the previous determination of (z) either of l
the following actions shall be taken:
M 1.
F (z) shall be increased over that specified in 4.2.2.2.c by the appropriate factor specified in the COLR, or M
R21 2.
F (2) shall be measured at least once per 7 effective full power days until 2 successive maps indicate that
' Ff(z) '
maximum over z h m hcreadm.
,(,,
L f.
With the relationships specified in 4.2.2.2.c above not being -
i satisfied:
1.
Calculate the percent F (z) exceeds its limit by the following expression:
'Ff(z) x W(z)'
i maximum over z
~1 x 100 for P t O'.S lR146 P
r l
W(z)'
' Ff(z) x maximum over z
~1 x 100 for P < 0. S lR146 l
Ff' I
O.5
.j
(
3 I
l 2.
Either of the following actions shall be taken:
a.
Place the core in an equilibrium condition where the R21 limit in 4.2.2.2.c is satisfied.
Power level may then be increased provided the AFD limits of Specification 3.2.1 are reduced 1% AFD for each percent FO(z) lR146 I
exceeded its limit, or b.
Comply with the requirements of Specification 3.2.2 for F (z) exceeding its limit by the percent calculated R21 above.
SEQUOYAH - UNIT 2 3/4 2-6 Amendment Nos. 21, 95, 131, 146, 206
- a y;
l t
ADMINISTRATIVE CONTROLS 1:
l l
l-MONTHLY REACTOR OPERATING REPORT l
lt 6.9.1.10 Routine reports of operating statistics and shutdown experience, I
' including documentation of all challenges to the PORVs or Safety. Valves, shall R64 l
be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the reg-i CORE OPERATING LIMITS REPORT R146 j
6.9.1.14 Core operating limits shall be established and documented in-the CORE l
OPERATING LIMITS REPORT before each reload cycle or.any remaining part of a reload cycle for the following:
l '.
Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3,
'2.
Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, 3.
Control Bank Insertion Limits for Specification 3/4.1.3.6, 4.
' Axial Flux Difference Limits for Specification 3/4.2.1, 5.
Heat' Flux Not Channel Factor, K(z), W(z), - and the f actor that accounts for.the potential, decrease in Fq margin between surveillances for Specification 3/4.2.2,-and R146
'6.
Nuclear Enthalpy Hot Channel. Factor and Power Factor Multiplier for Specification 3/4.2.3.
6.9.1.14.a-The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in:
1.
WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY",
July 1985 (E Proprietary).
(Methodology for Specification 3.1.1.3 Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit., 2.1.3.6 '
j
~ Control Bank Insertion' Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Hot Channel Factor.)
2.
WCAP-10216-P-A, Revision 1A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL Fo SURVEILLANCE TECHNICAL SPECIFICATION",' FEBRUARY 1994 (E Proprietary).
(Methodology for Specification 3.2.1 - Axial Flux Difference R146 i
(Relaxed Axial Offset Control) and 3.2.2 - Heat Flux Hot Channel Factor (w(z) surveillance requirements for Fq Methodology).)
j 3.
WCAP-10266-P-A, Rev.
2, "THE 1981 REVISION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987, (E Proprietary).
j (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).
4.
WCAP-136316-P-A, " SAFETY EVALUATION SUPPORTING A MORE NEGATIVE EOL R161 i
MODERATOR TEMPERATURE COEFFICIENT TECHNICAL SPECIFICATION FOR THE SEQUOYAh NUCLEAR PLANTS," March 1993, (E Proprietary).
-(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient).
SEQUOYAH - UNIT 2 6-22 Amendment Nos. 44, 50, 64, 66, 107, 134, 142,.146, 16L?06 l
. m. m
..m._
. _._.-._ _.._ _ ~.____
_ -... __.. _ _, _., _. _ _ - _. _... _ _ _. _ _ ~. _ _.
i i
ADMINISTRATIVE CONTROLS-CORE OPERATING LIMITS REPORT (continued)
R146 6.9.1'.14.b 'The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown. margin, and. transient and accident analysis limits) of the safety analysis are met.
6.9,..'14.c THE CORE OPERATING LIMITS REPORT shall be provided within 30 days
~
- after. cycle start-up (Mode 2) for each reload cycle or within 30 days of issuance of-any midcycle revision of the NRC Document Control _ Desk with copies to the Regional Administrator and Resident Inspector, j
SPECIAL REPORTS 6.9.2.1= Special' reports shall be submitted within the time period specified for each report, in accordance with 10 CFR 50.4.
R64 6.9.2.2 ' Diesel ~ (2anavator Reliability Imox6v---'t Procrram As a minimum the Reliability' Improvement Program report for NRC audit, required by LCO 3.8.1.1, Table 4.8-1, shall include:
R44 i
(a) a' summary of all tests (valid and' invalid) that occurred within the time period over which the last 20/100 valid tests were performed (b) : analysis of failures and determination of root causes of failures (c) evaluation of each of the recommendations of NUREG/CR-0660, " Enhancement of Onsite Emergency Diesel Generator Reliability in~ Operating Reactors,"
- with respect to their application to the Plant (d) -identification of all actions taken or to be taken to 1) correct the root causes of failures defined in b) above and 2) achieve a general
-improvement of diesel generator reliability
- (e).the schedule for implementation of each action from d) above (f) an assessment of the existing reliability of electric power to engineered-safaty-fenture equipment
-l d
- SEQUOl*,W - UNIT 2 6-22a Amendment Nos. 44, 50, 64, 66, 107, 134, 146, 206-
.-