ML19209C171
| ML19209C171 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/10/1979 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML19209C168 | List: |
| References | |
| NUDOCS 7910120124 | |
| Download: ML19209C171 (14) | |
Text
.
ENCLOSURE 1 GUIDE TO PROPOSED UNIT 3 TECHNICAL SPECIFICATION CHANGES Page 26.
. Relief valves Page 27.... Relief vaJves Page 30.... Relief valves Page 153.... LPCI modification Page 154.
. LPCI modification Page 225.... Relief valves Page 225a.
. Relief valves 1143 127 2Y 1
7910120 r-
sal'ETY LI"IT LIMITING SAFETY SYSTEM SETTING 1./
HEAS T_OR COOLANT SYSTEM 2.2 REACTOh COOLANT SYSTEM INTEGHITY INTEGRITY A pplica ul lit y Applicability Applies to limits on reactor Applies to trip settings of the coolant system pressure.
instruments and devices which are provided to prevent the reactor system safety limts from being exceeded.
Obiective Obiective To establish a limit below To define the level of the which the integrity of the process variables at which reactor coolant system is not automatic protective action is threatened due to an initiated to prevent the overpressure condition.
pressure safety limit from being exceeded.
Specification Speci f i ca tion The limiting safety system settings shall be as specified A.
The pressure at the lowest below:
point of the reactor vessel chall not exceed Limiting 1,375 psig whenever Safety irradiated fuel is in the Protective System teactor vessel.
Action Setting A.
Nuclear system 1,250 psig safety valves 1 13 psi open--nuclear (2 valves) system pressure lb B.
Nuclear system J
relief valves open--nuclear system pressure Target - Rocks 1,105 psig 2 11 pri
( 4 valves) 1,115 psig 2 11 psi
( 4 valves) 26 i143 128
LIMITING SAFETY SYSTEM SETTING SAFETY LIMIT
- 1. 2 PEACTOR COOLANT SYSTF.M 2,2 REACTOR COOIANT SYSTEM
,IJJTEGRITY ItTTEGRITY 1,1.25 Psig 1 11 psi (1 valve )
Crosbys**
1,150 Psig i 11 psi (2 valves)
OR Target-Rock **
I,125 psig 1
11 psi (2 valves)
C.
Scram-nuclear 6, 1,055 psig system high pressure
- Analyses have been run which allow operation with either 9 Target-Rocks and 2 Crosby's or 11 Target-Rocks as indicated in the above specification.
The results of these analyses are presented in
the. Bases.
1143 129 27
2.2 BASES REACTOR COOLANT SYSTEli INTEGRITY The pressure relief system for each unit at the Browns Ferry Nuclear Plant has been sized to meet two design bases. First, the total safety / relief valve capacity has been established to meet the over-pressure protection criteria of the ASME Code. Second, the distribution of this required capacity between safety valves and relief valves has been set to meet design basis 4.4.4-1 of sub-section 4.4 which states that the nuclear system relief valves shall prevent opening of the safety valves during normal plant isolations and load rejections.
The details of the analysis which shows compliance with the ASt1E Code requirements is presented in subsection 4.4 of the F5AR and the Reactor Vessel Overpressure Protection Sumary Technical Report submitted in response to auestion 4.1 dated December 1, 1 M i, 9 Target Rock And 2 Crosby Valves To meet the safety design basis, thirteen safety-relief valves have been installed on each Jnit with a total capacity of 81.08%ef nuclear boiler rated steam flow. The analysis of the worst overpressure transient, (3-second closure of all main stean lire isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure of 1293 asig if a neutron flux scram is assumed. This results in a 82 h
psig margin to the code allowable overpressure limit of 1375 psig.
To meet the operational design basis, the total safety-relief capacity of 81.08%of nuclear boiler rated has bean divided into 66.88% relief (11 valves) and 14.2% safety (2 valves). The analysis of the plant isolation transient (turbine trip with bypass valve failure to open) assuming a turbine trip scram is presented in Reference 5 on page 29.
This analysis shows that the 11 relief volves limit pressure at the safety valves to 1218 psig, well below the setting of the safety valves. Therefore, the safety valves will not open. This analysis shows that peak system pressure is limited to 1243 psig which is 132 psig below the allowed vessel overpressure of 1375 psig, 11 Target Rock Valves Only To meet the safety design basis, Dirteen safety-relief valves have t.een installed on each unit with a total capacity of 84.2% of nuclear boiler rated steam flow. The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure of 1280 psig if a neutron flux s. ram is assumed. This results in a 95 psig margin to the code allowable overpressure limit of 1375 psig.
To meet the operational design basis, the total safety-relief capacity of 84.2% of nuclear boiler rated has been divided into 70% relief (11 valves) and 14.2% safety (2 valves). The analysis of the plant isolation transient (turbine trip with bypass valve failure to open) h assuming a turbine trip scram is presented in Reference 5 on page 29.
This analysis shows that the 11 relief valves limit pressure at the safety valves to 1206psig, well below the setting of the safety valves. Therefore, the safety valves will not open. This analysis shows that peak system pressure is limited to 1232 psig which is 143 psig below the allowed vessel overpressure of 1375 psig.
30 1143 130
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS A
A 3.5 COhE_AND CONTAINMENT 4.5 CORE AND CONTAINMENT COOLING COOLING SYSTEMS SYSTEMS 8.
If specifications second operable 3.5.B.1 through access path for the sama phase of the 3.5.B.7 are not met, an orderly shutdown mode (drywell sprays, snall be initiated suppression chamber and the reactor shall st nys and be shutdown and suppression pool placed in the cold cooling) shall be condition within 24 demonstrated to be hours.
operable daily thereafter until the 9.
When the reactor Sr.cond path is vessel pressure is returned to normal atmospheric and service.
irradiated fuel is in the reactor vessel at 8.
No additional least one RHR loop surveillance with two pumps or two r equir ed.
loops with one pump per loop shall be o
n 9.
When the reactor operable.
The pumpse associated diesel vessel pressure is generators must also atmospheric, the RHR be operable.
pumps and valves that are required to be 10.
If tae conditions of operable shall be demon-specification 3.5.A.5 strated to be operab:>
are met, LPCI and monthly.
containment cooling are not required.
10.
No additional surveill mce required.
11.
When there is irradiated fuel in 11.
The B and D RHR pumps the reactor and the reactor vessel on unit 2 which pressure is greater supply cross-connect than atmospheric, capability shall be unit 2 RHR pumps B operable monthly when and D with associated the cross-connect heat exchangers and capability is required.
valves must be operable and capable 12.
When it is determined of supplying cross-that one RiiR pump or connect capability associated heat except as specified in specification exchanger located on e
3.5.B.12 below.
the unit cross-connection in the 153 ll4}
j
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.5 CORE AND CONTAINMENT COOLING l.5 (: ORE AND COffrAIN.4ENT SYSTEMS C00LI:3G SYSTEMS (Ibte:
Because adjacent unit is cross-connect inoperable at a time capability is not a when operability is short term required, the requirement, a remaining RilR pump compon ent is not and associated heat considered inoperable exchanger on the unit if cross-connect cross-connection and capability can be the associated diesel restored to service generator shall be within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.)
demonstrated to be operable immediately 12.
If one RHR pump or and every 15 days associated heat thereafter until the exchanger located on inoperable pump and the unit cross-associated heat connection in unit 2 exchanger are is inoperable for any returned to normal reason (including
- service, valve inoperability, pipe break, etc.),
the reactor may remain in operation for a per2.od not to exceed 30 days provided the remaining RilR pump and associated diesel 13.
No additional surveil?ance generator are required.
13.
If EliR cross-connection flow or 14.
All recirculation pump heat removal discharge valves shall capability in lost, be tested for operability the unit mai remain during any period of in operation for a reactor cold shutdown period not to exceed exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, 10 days unless suco if operability tests "pa have not been performed e to e during the preceeding 31 days.
14.
All recirculation pump discharge valves shall be operable prior to reactor startup (or closed if permitted p
elsewhere in these 154 Specifications).
f f /* j
3.6/4.6 B AS ES 9 Target Rock And 2 Crosby Valves To meet the safety design basis, thirteen safety-relief valves have been installed on unit 2 with a total capacity of 81.08% of nuclear boiler rated steam flow. The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure of 1293 psig if a neutron fl ux scram is assumed This results in an 82 psig margin of the code allowable over-pressure limit of 1375 psig.
To meet the operational design basis, the total safety-relief capacity of 81.08% of nuclear boiler rated has been divided into 66.88% relief (11 valves) and 14.2% safety (2 valves). The analysis of the plant iso-lation transient (turbine trip with bypass valve f ailure to open) assuming a turbine trip scram is presented in Reference 5 on page 29.
This analysis shows that the 11 relief valves limit pressure at the safety valves to 1218 psig, well below the setting of the safety valves.
Therefore, the safety valves will not open.
This analysis shows that peak system pressure is limited to 1243 psig which is 132 psig below the allowed vessel overpressure of 1375 psig.
/"~
11 Target Rock Valves Only To meet the safety design basis, thirteen safety-relief valves have been installed on unit 2 with a total capacity of 84.2% of nuclear boiler rated steam flow. The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure of 1280 psig if a neutron flux scram is assumed This results in an 95 psig margin of the code allowable over-pressure limit of 1375 psig.
To meet the operational design basis, the total safety-relief capacity of 84.2% of nuclear boiler rated has been divided into 70% relief (11 valves) and 14.2% safety (2 valves). The analysis of the plant iso-lation transient (turbine trip with bypass valve failure to open) assuming a turbine trip scram is presented in Reference 5 on page 29.
This analysis shows that the 11 relief valves limit pressure at the safety valves to 1206 psig, well below the setting of the safety valves. Therefore, the safety valves will not open.
This analysis shows that peak system pressure is limited to 1232 psig which is 143 psig below the allowed I
vessel overpressure of 1375 psig.
/
22s j j43
- 33
3.6/4.6 B AS ES Experience in relief and safety valve operation shows that a testing of 50 percent of the valves per year is adeqs..be to detect f ailures or deteriorations.
The relief and safety valves are benchtested every second operating cycle to ensure that their set points are within the il percent tolerance.
The relief valves are tested in place once per operating cycle to establish that they will open and pass steam.
The requirements established above apply when the nuclear system be pressurized above ambient coaditions.
These requirements can are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed.
However, these transients are much less severe, in terms of pressure, than those starting at rated conditions.
The valves need not be f unctional when the vessel head is removed, since the nuclear system cannot be pressurized.
REFERENC ES 1.
Nuclear System Pressure Relief System (BFNP FSAR Subsection
- 4. 4) r fd t
225a
ENCLOSURE 2 li43 135
NUCLE AR ENERGY BUSINESS GROUP e GENER AL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GENER AL $ ELECTRIC APPLICABLE TO:
PUBLICATION NO.
ERun And MDmDA NEDO-24199 T. L E. NO.
79 NED-281 SHEET Browns Ferry Unit 3 i
TIT LE Reload 2 Supplemental 9/25/79 DATE Licensine Submittal NO TE: Correct allcopies of the apphcable ISSUE DATE June 1974 publication as spectfied below.
REFERENCES INSTRUCTIONS
,j5,ECTION, PAG (CORRECTIONS AND ADDITIONS)
ITEM pg NE) 1.
Appendix A Replace with attached Appendix A.
PAGE 1143 136
NEDO-24199 APPENDIX A Fuel Loading Error LHGR:
16.02 kW/ft Safety / Relief Valve Capacity at Setpoint (No./%):
11/70 Spring Safety Valve Capacity at Setpoint (No./%): 2/14.2 ANALYSES FOR ALTERNATE SAFEIT/ RELIEF VALVES For the purpose of monitoring valve performance, two 6R10 wrosby Safety Relief Valves (SRVs) will be installed for cycle 3 operation. The two Crosby SRVs 1150 psig will replace two Target Rock valves set at 1125 psig in set at locations G and H which are not automatic depressurization system (ADS) locations.
The Crosby SRV is a simple, direct-acting, spring-loaded valve with an external pneumatic piston. Safety valve action occurs when the inlet pres-sure f orces exceed the spring load and force the valve disc of f of its seat.
For manual actuation, the external pneumatic piston is capable of opening the the f orce of the spring at any steam pressure down to O psig.
valve against The pneumatic operator is so arranged that if it malfunctioned it would not the valve disc f rom lif ting if steam inlet pressure reached the spring prevent set pressure.
Since the Target Rock valves on Browns Ferry Unit No. 3 have had their throats enlarged to provide increased capacity, the capacity of each of the two Crosby replacement valves is 94.3% of each of the modified Target Rock valves when compared at the same inlet pressure.
The SRV change is indicated by the following table:
TO FROM Current BF3/C3 Proposed SRV
- Setpoint, SRV
- Setpoint, SRV psig
- of Target Rock psig
- of TR
,# of Crosby 1105+1%
4 1105+1%
4 0
1115+1%
4 1115+1%
4 0
1125+1%
3 1125+1%
1 0
1150+1%
0 2
A-1 hbf
NEDO-24199 The total SRV capacity at setpoint used in these analyses was 66.88%.
For Browns Ferry 3 cycle 3, transient analyses for the limiting transients have been performed to evaluate the impact of using the 2 Crosby Safety Relief Valves set at 1150 psig instead of 2 high set Target Rock SRVs, and the plant responses are indicated in Figure A-1 and A-2 which are similar to previous
'the analyses presented used preliminary Figures 3 and 7, respectively.
These analyses are estimates of the capacity of the Crosby SRVs as 75.5%.
conservative when applied with the actual value of the valve capacity (94.3%).
The first portion of the table in Section 9 changes as follows:
Core P
aC?R Plant Power Flow 4
Q/A F,1 y
Transient Ex p sure (2)
(2) (% NBR) (% NBR) (psig) (psig) 8x8 8xBR F8x8R Response Load Rejection BOC-EOC 104.5 100 229.5 109.3 1213 1243 0.15 0.15 0.16 Figure A-1 without Bypass The table in Section 12 becomes OVERPRESSUR1ZATION ANALYSIS
SUMMARY
(5.3)
P Plant Power Core Flow Psl y
(%)
(%)
(psig)
(psig)
Response
MS1V Closure 104.5 100 1260 1293 Figure A-2 (Flux Scram) it is still The peak vessel bottom pressure increased 13 psi to 1293 psig, but well below the limit of 1375 psig for vessel overpressure protection.
There is no effect on the peak heat flux because the SRVs with setpoints of 1125 psig or greater open after the time of peak heat flux and therefore do not impact th' ACPR.
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