ML20214F259
| ML20214F259 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 05/12/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20214F257 | List: |
| References | |
| NUDOCS 8705260006 | |
| Download: ML20214F259 (2) | |
Text
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o UNITED STATES 8
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NUCLEAR REGULATORY COMMISSION h
WASHINGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOAgTO FACILITY OPERATING LICENSE DPR-77 AND AMENDMENT N047 TO FACILITY OPERATING LICENSE DPR-79 TENNESSEE VALLEY AUTHORITY INTRODUCTION By letter dated December 23, 1986, the licensee requested technical specification changes that would permit installation of slower acting valve operators on valves between the suction of the centrifugal charging pumps and the volume control tank.
The valves are designed to close following a safety injection signal to pemit the centrifugal charging (CC) pumps to take suction solely from the refueling water storage tank (RWST). Closure of the valves is required since full flow from the centrifugal charging pumps would cause the volume control tank (VCT) to be drained within a few minutes.
In addition closure of the valves ensures that the 2000 ppm boric acid solution in the RWST will be pumped into the core to tenninate the reactivity excursion which would be produced by a large steam line break. The boric acid concentration in the VCT is normally the same as the reactor system.
EVALUATION The valves between the VCT and the CC pumps suction are currently equipped with fast closing operators which require a device to limit the impact of the valve internals on the valve seats. Motor breaks were therefore installed for the valves at both Units 1 and 2.
The motor breaks at Unit 1 are currently l
inoperable. TVA has determined that the installation of slower closing valve
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operators will be acceptable and proposes to modify the operators at both units to decrease the closing speed and to remove the motor breaks at both units. The use of slower valve operators would increase the closing time by 15 seconds over that which was assumed in the FSAR safety evaluations and requires that the safety injection delay times in the Technical Specifications be increased by 5 seconds.
TVA detemined that the consequences of design basis transients and accidents would be unaffected by the slower closing of the valves. The licensee presented that Westinghouse performed an evaluation of delayed ECCS actuation consistent with the increased response times in the Technical Specifications.
It was detennined from that evaluation that delayed ECCS actuation has the potential of affecting the mitigation of steam line break and LOCA events.
Following a large break LOCA the accumulators would inject abundant ECCS water into the reactor during the blowdown period when the valves would be closing. Before the reflooding 8705260006 870512 DR ADOCK 0500 7
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4 period when flow from the'CC pumos aids in reflooding the core, the CC oump i
suction valves would have closed so that-the time to recover the core would be i
unaffected.
In the case of a small break LOCA the peak cladding temperature is not reached for over 1000 seconds after the break occurs. An additional I
delay in ECCS response as requested for the Technical Specifications would have an insignificant effect on the total pumped flow and hence on the amount i
of core heatup.
In the case of a break in a main steam line, previously approved analyses for removal of the boron injection tank' (BIT) concluded that recriticality and return to power during blowdown of one steam generator would i
not cause. fuel failure.
Injection of boric acid from the RWST would be-l required eventually to terininate the event; however sensitivity studies by-
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Westinghouse have determined that the requested delay in the ECCS response would not have a significant effect on the minimum DNBR which would occur at i
approximately 100 seconds into the event. Based on the above considerations the staff concludes that the Technical Specifications for Sequoyah Units 1 and.
2 may be modified as requested by.the licensee.
l ENVIRONMENTAL CONSIDERATION These amendments involve changes to the installation or use of the facilities' i
components located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendments involve no significant increase in j
the amounts, and no significant change in the types, of any effluents that may j
be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Connission has previously i
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issued a proposed finding that these amendments involve no significant hazards consideration, and there has been no public cocuent on such finding. Accordingly,
.the amendments meet the eligibility criteria for categorical exclusion set i
forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
i CONCLUSION The Commission made a proposed determination that the amendments involve no i'
significant hazards consideration which was published in 'the Federal Register on February 26, 1987 (52 FR 5870) and consulted with the state of Tennessee.
No public comments were received, and the state of Tennessee did not have any j
comments.
We have concluded, based on the considerations discussed above, that: (1)there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Consnission's regulations and the issuance of j
these amendments will not be inimical to the common defense and security or to
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the health ~and safety of the public, i
i Principal Contributors: Joe Holonich, PWR#4, DPWR-A i
Walt Jensen, PARS, DPWR-A Dated:
May 12, 1987 i
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UNITED STATES o
8 NUCLEAR REGULATORY COMMISSION n
g i-WASHINGTON, D. C. 20606 l
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TENNESSEE VALLEY AUTHORITY DOCKET NO.
50-327 SEOU0YAH NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 55 License No. DPR-77 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment to the Sequoyah Nuclear Plant, Unit 1 (the facility) Facility Operating License No. DPR-77 filed by the Tennessee Valley Authority (licensee), dated December 23, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the license, as amended, the provisions of the Act, and the rules and regulations of the Comission; 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and 1
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is hereby amended by page changes to the Apoendix A Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. OPR-77 is hereby amended to read as follows:
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2-(2) Technical Specifications The Technical Specifications contained in Apoendix A, as revised through Amenoment No. 55 are hereby incorporated into the. 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.
FOR THE NUCLEAR REGULATORY COMMISSION i
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John
- i. Zwolinski, Assistant Director
- for i rojects P
Division of TVA Projects Office of Special Projects Attachment Appendix A Technical Specification Changes Date of Issuance:
May 12, 1987 i
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ATTACHMENT TO LICENSE AMENDMENT NO. 55 FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change. Overleaf page provided to maintain document completeness.*
y REMOVE INSERT 3/4 3-29 3/4 3-29 3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 3/4 3-32 3/4 3-32*
3/4 3-33 3/4 3-33 A--
TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 1.
Manual a.
Safety Injection (ECCS)
Not Applicable Feedwater Isolation Not Applicable Reactor Trip (SI)
Not Applicable Containment Isolation-Phase "A" Not Applicable Containment Ventilation Isolation Not Applicable Auxiliary Feedwater Pumps Not Applicable Essential Raw Cooling Water System Not Applicable l
Emergency Gas Treatment System Not Applicable b.
Containment Spray Not Applicable Containment Isolation-Phase "B" Not Applicable Containment Ventilation Isolation Not Applicable Containment Air Return Fan Not Applicable c.
Containment Isolation-Phase "A" Not Applicable Emergency Gas Treatment System Not Applicable Containment Ventilation Isolation Not Applicable d.
Steam Line Isolation Not Applicable 2.
Containment Pressure-High 32.0(1) l a.
Safety Injection (ECCS) i l
b.
Reactor Trip (from SI) 1 3.0 l
c.
Feedwater Isolation
-< 8.0(2) d.
Containment Isolation-Phase "A"(3) 1 18.0(8)/28.0(9) e.
Containment Ventilation Isolation Not Applicable f.
Auxiliary Feedwater Pumps 1 60 g.
Essential Raw Cooling Water System 1 65.0(8)/75.0(9) h.
Emergency Gas Treatment System 1 38.0(9) l l
l SEQUOYAH - UNIT 1 3/4 3-29 Amendment No. 55 l
n
TABLE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 3.
Pressurizer Pressure-Low 32.0(1)/28.0(7) a.
Safety Injection (ECCS) 1 b.
Reactor Trip (from SI) 1 3.0 c.
Feedwater Isolation
< 8.0(2)
Containment Isolation-Phase "A"(3) 18.0(8) d.
e.
Containment Ventilation Isolation Not Applicable f.
Auxiliary Feedwater Pumps s 60 g.
Essential Raw Cooling Water System 1 65.0(8)/75.0(9) h.
Emergency Gas Treatment System 1 28.0(8) 4.
Differential Pressure Between Steam Lines-High a.
Safety Injection (ECCS) 1 28.0(7)/28.0(1) b.
Reactor Trip (from SI) 1 3.0 c.
Feedwater Isolation
< 8.0(2) d.
Containment Isolation-Phase "A"(3) 18.0(8)/28.0(9) e.
Containment Ventilation Isolation Not Applicable f.
Auxiliary Feedwater Pumps 1 60 g.
Essential Raw Cooling Water System 1 65.0(8)/75.0(9) h.
Emergency Gas Treatment System 1 38.0(9) 5.
Steam Flow in Two Steam Lines - High Coincident with T
--Low-Low a.
S ety Injection (ECCS) s 30.0(7)/30.0(1) b.
Reactor Trip (from SI) 1 5.0 c.
Feedwater Isolation
< 10.0(2) d.
Containment Isolation-Phase "A"(3) 20.0(8)/30.0(9) e.
Containment Ventilation Isolation Not Applicable f.
Auxiliary Feedwater Pumps s 60 g.
Essential Raw Cooling Water System 5 67.0(8)/77.0(9) h.
Steam Line Isolation
< 10.0 1.
Emergency Gas Treatment System 40.0(9)
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SEQUOYAH - UNIT 1 3/4 3-30 Amendment No. 55 rt
TABLE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 6.
Steam Flow in Two Steam Lines-High Coincident with Steam Line Pressure-Low a.
Safety Injection (ECCS) 1 28.0( )/28.0(1) b.
Reactor Trip (from SI) 1 3.0 c.
Feedwater Isolation
< 8.0(2)
Containment Isolation-Phase "A"(3) 18.0(8)/28.0(9) d.
e.
Containment Ventilation Isolation Not Applicable f.
Auxiliary Feedwater Pumps 1 60-g.
Essential Raw Cooling Water System 1 65.'0(d)/75.0(9) h.
Steam Line Isolation
< 8.0 1.
Emergency Gas Treatment System 38.0(9) 7.
Containment Pressure--High-High a.
Containment Spray 1 58.00(9)
'65(8)/75(9) b.
Containment Isolation-Phase "B" 1
c.
Steam Line Isolation 1 'i. 0 '
d.
Containment Air Return Fan
> 540.0 and 1 660 8.
Steam Generator Water Level--High-High a.
Turbine Trip-Reactor Trip i 2.5 b.
Feedwater Isolation i 11.0(2) 9.
Main Steam Generator Water Level -
Low-Low a.
Motor-driven Auxiliary 1 60.0 Feedwater Pumps (4) h.
Turbine-driven Auxiliary 1 60.0 Feedwater Pumps (5)
SEQUOYAH - UNIT 1 3/4 3-31 Amendment No.J2, 55 i
n
TABLE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS 10.
Station Blackout
- ?'
a.
ALxiliary Feedwater Pumps 5 60 i,
11.
Trip'of' Main Feedwater Pumps a.
Auxiliary Feedwater Pumps 1 60
- 12. Loss of Power a.
6.9 kv Shutdown Board - Degraded i 10(10)
Voltage or Loss of Voltage 13.
RWST Level-Low Coincident with Containment Sump Level-High and Safety Injection i
a.
Automatic Switchover to Containment Sump 1 250 14.
Containment Purge Air Exhaust Radioactivity - High
, 10(6) a.
Containment Venti 1'ation Isolation 15.
Containment Ga.; Monitor Radioactivity High a.
Containment Ventilation Isolation i 10(6) 16.
Containment Particulate Activity High a.
Containment Ventilation Isolation 1 10(6)
- NOTE:
This technical specification to be implemented at the startup following the second refueling outage or following completion of the modification, whichever is earlier.
r SEQUOYAH - UNIT 1 3/4 3-32 Amendment No. 29 t
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TABLE 3.3-5 (Continued)
TABLE NOTATION (1) Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps, SI and RHR pumps.
(2) Using air operated valve (3) The following valves are exceptions to the response times shown in the tat'le and will have the values listed in seconds for the initiating sig-nals and function indicated:
Valves:
FCV-26-240, -243 Response times:
2.d. 21(( / 31( )
3.d. 22 4.d. 21(8)/ 31(9) 1 31 Valves:
FCV-61-96, -97, -110, -122, -191, -192, -193, -194 Response times:
~ 2.d. 31(8) 3.d. 32 4.d. 31(8) 5.d. 34(8) 6.d. 31 Valve:
FCV-70-143 Response times:
2.d. 61(8)f71(9) 3.d. 62((8) l 4.d. 61(8)f77(9) 5.d. 64(0)/74 I) 6.d. 61 0)/7159)
(4) On 2/3 any Steam Generator (5) On 2/3 in 2/4 Steam Generator (6) Radiation detectors for Containment Ventilation Isolstion may be excluded from Response Time Testing.
(7) Diesel generator starting and sequence loading delays not included. Offsite power available.
Response time limit includes opening and closing of valves l
to establish SI path and attainment of discharge pressure for centrifugal charging pumps.
(8) Diesel generator starting and sequence loading delays not included.
Response
time limit includes operating time of valves.
(9) Diesel Generator starting and sequence loading delays included.
Response
time limit includes operating time of valves.
SEQUOYAH - UNIT 1 3/4 3-33 Amendment No.J7, 55
UNITED STATES l
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9, NUCLEAR REGULATORY COIMilSSION
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WASHINGTON,D C.20005 a
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i TENNESSEE VALLEY AUTHORITY DOCKET NO.
50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 47 License No. DPR-79 l
1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment to the Sequoyah Nuclear Plant, Unit 2 (the facility) Facility Doerating License No. DPR-79 filed by the i
Tennessee Valley Authority (licensee), dated December.23, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations as set forth in 10 CFR Chapter I; i
B.
The facility will operate in conformity with the license, as amended, the provisions of the Act, and the rules and _ regulations of the 4
Commission, C.
There is reasonable assurance (1) 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 Comission's regulations-D.
The issuance of this amendment will not be inimical.to the comon 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 a
the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the-license is hereby amended by page changes to the Appendix A 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:
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. (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 47 are hereby incorporated into the license. ine licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMlSSION
(,
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John
. Zwolinski, Assistant Director for Projects Division of TVA Projects Office of Special Projects Attachment Appendix A Technical Specification Changes Date of Issuance: May 12, 1987 y
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ATTACHMENT TO LICENSE AMENDMENT NO. 47 FACILITY OPERATING LICENSE NO DPR-79 DOCKET NO. 50-328 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change. Overleaf page provided to maintain document completeness.*
REMOVE INSERT 3/4 3-29 3/4 3-29 3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 3/4 3-32 3/4 3-32*
3/4 3-33 3/4 3-33
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TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 1.
Manual a.
Safety Injection (ECCS)
Not Applicable Feedwater Isolation Not Applicable Reactor Trip-(SI)
Not Applicable Containment Isolation-Phase "A" Not Applicable Containment Ventilation Isolation Not Applicable Auxiliary Feedwater Pumps Not Applicable Essential Raw Cooling Water System Not Applicable Emergency Gas Treatment System Not. Applicable b.
Containment Spray Not Applicable Containment Isolation-Phase "B" Not Applicable Containment Ventilation Isolation Not' Applicable Containment Air Return Fan Not Applicable c.
Containment Isolation-Phase "A" Not Applicable.
Emergency Gas Treatment System Not Applicable Containment Ventilation Isolation
_Not Applicable d.
Steam Line Isolation Not Applicable 2.
Containment Pressure-High a.
Safety Injection (ECCS)
< 32.0(1) b.
Reactor Trip (from SI) 10 3
<8.0(2) c.
Feedwater Isolation
~118.0(8)/28.0(9)
Containment Isolation-Phase "A"(3) d.
e.
Containment Ventilation Isolation Not Applicable f.
Auxiliary Feedwater Pumps 160 65.0(8)/75.0(9) g.
Essential Raw Cooling Water System 1
h.
Emergency Gas Treatment System 138.0(9) i SEQUOYAH - UNIT 2 3/4 3-29 Amendment No. 47
. cr
L TABLE 3.3-5-(Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES
. INITIATING SIGNAL'AND-FUN'CTION RESPONSE TIME-IN SECONOS'
~ 3.
Pressurizer Pressure-Low.
III/28.0(7)
. a.
Safety Injection (ECCS)
-1 2.0 3
i b.
' Reactor Trip (from SI).
<-3.0; y
c.
Feedwater Isolation-8.0(2) d.
. Containment Isolation-Phase "A"(3)'
- {18. 0(8) e.
Containment-Ventilation Isolation Not' Applicable-f.
Auxiliary Feedwater Pumps' 160' g.
Essential Raw Cooling Water System 165.0(8)/75.0(9) h.
Emergency Gas Treatment System-128.0(8) 4.
Differential Pressure Between-Steam Lines-High' a.
Safety Injection (ECCS)
< 28. 0(7)/28.'0(1).
b.
Reactor Trip (from SI) 13.0 c.
Feedwater Isolation:
<8.0(2).
d.
Containment Isolation-Phase "A"(3) 18.0(0)/28.0(9) e.
Containment' Ventilation _ Isolation
.Not Applicable-f.
Auxiliary Feedwater Pumps 160 g.
Essential Raw Cooling Water System 165.0(8)/75.0(9) h.
Emergency Gas Treatment System 138.0(9)'
l 5.
Steam Flow in Two Steam Lines - High' Coincident with T
--Low-Low a.
Safety Injection (ECCS) 130.0(7)/30.0(1) -l b.
Reactor Trip (from SI) 1 '. 0 -
5 c.
Feedwater Isolation
<10.0(2) d.
Containment-Isolation-Phase "A"(3) l 20.0(8)/30.0(9) e.
Containment Ventilation Isolation.
Not Applicable f.
Auxiliary Feedwater Pumps
'160 -
g.
Essential Raw Cooling Water System
_167.0(0)/77.0(9) h.
Steam Line Isolation
<10.0 i.
Emergency Gas Treatment System 140.0(9)-
m SEQUOYAH - UNIT 2 3/4 3-30
' Amendment No.'47, y.
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4 TABLE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 6.
Steam Flow in Two Steam Lines-High Coincident with Steam Line Pressure-Low a.
Safety Injection (ECCS)
< 28.0(7)/28.0(1) b.
Reactor Trip (from SI)
< 3.0 c.
Feedwater Isolation
< 8.0(2) d.
Containment Isolation-Phase "A"(3) h18.0(8)/28.0(9) e.
Containment Ventilation Isolation Not Applicable f.
Auxiliary Feedwater Pumps
$60 g.
Essential Raw Cooling Water System 5 65.0(8)/75.0(9) h.
Steam Line Isolation
< 8.0 1.
Emergency Gas Treatment System 38.0(9) 7.
Containment Pressure--High-High a.
< 58.00(9) b.
Containment Isolation-Phase "B" 65(8)/75(9) c.
Steam Line Isolation 5 7.0 d.
Containment Air Return Fan
> 540.0 and 1660 8.
Steam Generator Water Level--High-High a.
Turbine Trip-Reactor Trip 5 2.5 b.
Feedwater Isolation
-< 11.0(2) 9.
Main Steam Generator Water Level -
Low-Lp a.
Motor-driven Auxiliary 5 60.0 umps (4)
Feedwater c b.
Turbine-driven Auxiliary 5 60.0 Feedwater Pumps (5)
SEQUOYAH - UNIT 2 3/4 3-31 Amendment No. 47
TABLE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL ~AND FUNCTION.
RESPONSE TIME IN SECONDS 10.
Station Blackout a.
Auxiliary Feedwater Pumps 1 60
- 11. Trip of Main Feedwater Pumps a.
Auxiliary Feedwater Pumps 1 60
- 12. Loss of Power a.
6.9 kv Shutdown Board - Degraded 5 10(10)
Voltage or Loss of Voltage 13.
RWST Level-traw Coincident with Containment Sump Level-High and Safety Injection a.
Automatic Switchover to Containment Sump 1 250 14.
Containment Purge Air Exhaust Radioactivity - High a.
Containment Ventilation Isolation 1 10(6) 15.
Containment Gas Monitor Radioactivity High a.
Containment Ventilation Isolation i 10(6) 16.
Containment Particulate Activity High a.
Containment Ventilation Isolation 1 10(6)
- NOTE: This technical specification is to be implemented during the startup following the first refueling outage.
SEQUOYAH - UNIT 2 3/4 3-32 Amendment No. 18 r>
INSTRUMENTATION TABLE 3.3-5 (Continued)
TABLE NOTATION (1) Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps, SI and RHR pumps.
(2) Using air operated valve (3) The following valves are exceptions to the response times shown in the table and will have the values listed in seconds for the initiating signals and function indicated:
Valves:
FCV-26-240, -243 Response times:
2.d. 21
/31(9) 3.d. 22 4.d. 21(0)/ (9)
(9) 5.d. 24
/
(9) 6.d. 21
/
Valves:
FCV61-96, -97, -110, -122, -191, -192, -193, -194 Response times 2.d. 31(8) 3.d. 32(8) 4.d. 31(8) 5.d. 34((8) 0)
6.d. 31 Valve:
FCV-70-143 2.d. 61((0)/71(9)
Response times:
8) 3.d. 62 4.d. 61(0)/71(9)
(4) On 2/3 any Steam Generator (5) On 2/3 in 2/4 Steam Generator (6) Radiation detectors for Containment Ventilation Isolation may be excluded from Response Time Testing.
(7) Diesel generator starting and sequence loading delays not included.
Offsite power available.
Response time limit includes opening and closing of_ valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.
(8) Diesel generator starting and sequence loading delays not included.
Response time limit includes operating time of valves.
(9) Diesel generator starting ar.d sequence loading delays included.
Response
time limit includes operating time of valves.
SEQUOYAH - UNIT 2 3/4 3-33 Amendment No. 8, 47
'[
o UNITED STATES g
8 NUCLEAR REGULATORY COMMISSION o
WASHINGTON, D. C. 20555 g
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOAgTO FACILITY OPERATING LICENSE DPR-77 AND AMENDMENT NO.a7 TO FACILITY OPERATING LICENSE DPR-79 TENNESSEE VALLEY AUTHORITY INTRODUCTION By letter dated December 23, 1986, the licensee requested technical specification changes that would permit installation of slower acting valve operators on valves between the suction of the centrifugal charging pumps and the volume control tank.
The valves are designed to close following a safety injection signal to pemit the centrifugal charging (CC) pumps to take suction solely from the refueling i
water storage tank (RWST). Closure of the valves is required since full flow from the centrifugal charging pumps would cause the volume control tank (VCT) to be drained within a few minutes.
In addition closure of the valves ensures that the 2000 ppm boric acid solution in the RWST will be pumped into the core to tenninate the reactivity excursion which would be produced by a large steam line break. The boric acid concentration in the VCT is normally the same as the reactor system.
EVALUATION The valves between the VCT and the CC pumps suction are currently equipped with fast closing operators which require a device to limit the impact of the valve internals on the valve seats. Motor breaks were therefore installed for the valves at both Units 1 and 2.
The motor breaks at Unit 1 are currently inoperable. TVA has determined that the installation of slower closing valve operators will be acceptable and proposes to modify the operators at both units to decrease the closing speed and to remove the motor breaks at both units. The use of slower valve operators would increase the closing tima by 15 seconds over that which was assumed in the FSAR safety evaluations and requires that the safety injection delay times in the Technical Specificadons be increased by 5 seconds.
TVA detennined that the consequences of design basis transients and accidents would be unaffected by the slower closing of the valves. The licensee presented that Westinghouse performed an evaluation of delayed ECCS actuation consistent with the increased response times in the Technical Specifications.
It was determined from that evaluation that delayed ECCS actuation has the potential of affecting the mitigation of steam line break and LOCA events. Following a large break LOCA the accumulators would inject abundant ECCS water into the reactor during the blowdown period when the valves would be closing. Before the reflooding l
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period when flow from the.CC pumps aids in reflooding the core, the CC pump suction valves would have closed so that the time to recover the core would be-unaffected.- In the case of a small break LOCA the peak cladding temperature is not reached for over 1000 seconds after the break occurs. An additional delay in ECCS response as requested for the Technical Specifications would:
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- have an insignificant effect on the total' pumped flow and hence on the amount of core heatup.
In the case of a break in a main steam line, previously approved analyses for removal of the boron injection tank (BIT) concluded that recriticality and return to power during blowdown of one steam generator would not cause fuel failure.
Injection of boric acid from the RWST would be required eventually to terminate the event; however sensitivity studies by Westinghouse have determined that the requested delay in the ECCS response 4
would not have a significant effect on the minimum DNBR which would occur at approximately 100 seconds into the event. Based on the above considerations the staff concludes that the Technical Specifications for-Sequoyah Units 1 and 2 may be modified as requested by the licensee.
ENVIRONMENTAL CONSIDERATION These amendments involve changes to the installation or us'e of the facilities' components located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendments involve no significant increase in
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the amounts, and no significant change in the types, of any effluents that may.
l be released offsite and that there-is no significant increase in individual or cumulative occupational radiation exposure. The Comission has previously issued a proposed finding that these amendments involve no significant hazards consideration, and there has been no public coment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection i
with the issuance of the amendments.
1 CONCLUSION The Commission made a proposed determination that the amendments involve no significant hazards consideration which was published in the Federal Register on February 26,1987 (52 FR 5870) and consulted with the state of Tennessee.
No public comments were received, and_the state of Tennessee did not have any Coments.
We have concluded, based on the considerations discussed above, that: (1) there
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is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be a
conducted in compliance with the Comission's regulations and'the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: Joe Holonich, PWR#4, DPWR-A Walt Jensen, PARS, DPWR-A 1
Dated:
May 12, 1987 l
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