ML20071H685
| ML20071H685 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 07/08/1994 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20071H683 | List: |
| References | |
| RTR-NUREG-1433 NUDOCS 9407190443 | |
| Download: ML20071H685 (256) | |
Text
{{#Wiki_filter:EDWIN 1. HATCH NUCLEAR PLANT IMPROVED TECHNICAL SPECIFICATIONS REVISION INSERTION INSTRUCTIONS REVISION B Paae Instruction Cover sheet (U1 Improved Sa'ecifications) Discard 3.6-13 through 3.6-15 Replace 3.8-5 through 3.8-9 Replace 3.8-15 Reolace 3.8-31 Replace 3.8-39 Replace Cover sheet (Unit 1 ImDroved Bases) Discard B 3.3-41 Replace B 3.3-42A Add B 3.3-43 Replace B 3.6-1 Repl ace B 3.6-7 Replace B 3.6-27 Replace B 3.6-69 Replace B 3.7-1 Replace B 3.8-19 Repl ace B 3.8-20A Add B 3.8-23 Replace B 3.8-24A Add B 3.8-25 Replace Blank with p. B 3.8-26 on back Add (following p. B 3.8-25A) B 3.8-55 Replace B 3.8-61 Repl ace B 3.8-63 Replace B 3.8-64A Add B 3.8-65 Repl ace Blank with p. B 3.8-66 on back Add (following p. B 3.8-65A) Cover sheet (U1 CTS MarkUD & DOC)N Discard CTS 3.1-7 (8 of 15) Replace CTS 3.1-9 (10 of 15) Replace 9 (DOC ITS 3.3.1.1) Replace CTS 3.3-5 (8 of 9) Repl ace 3A (DOC ITS 3.3.2.1) (following page 3) Add CTS 3.2-22 (1 of 9) Replace CTS 3.2-22a Replace CTS 3.2-42 (8 of 10) Replace CTS 3.2-43 (9 of 10) Replace CTS 3.2-42 (3 of 6) Repl ace
- a. In replacing each CTS page, reference the upper right corner for appropriate ITS section.
9407190443 940708 fDR ADOCK 0500072) PDR
' Revision B Insertion Instructions-(continued) :
P Pace Instruction (U1 CTS Markup & DOC) continued i CTS 3.2-43 (4 of 6) Replace ' CTS 3.2-44 (5 of 6) Replace 2 (DOC ITS 3.3.8.2) Replace , CTS 3.2-42 (3 of 5) Replace CTS 3.2-21 (4 of 6) Replace , CTS 3.2-46 (5 of 6) Replace 3 (DOC ITS 3.6.1.1) Replace 4 (DOC ITS 3.6.1.1) Replace CTS 3.7-10a (12 of 13) Replace , I (DOC ITS 3.6.1.4) Replace
- CTS 3.9-2 (2 of 13) Replace 1 (DOC ITS 3.8.1) Replace CTS 3.9-2c (1 of 5) Replace CTS 3.9-3 (2 of 5) Replace 1 (DOC ITS 3.8.4) Repl ace .
3 (DOC ITS 3.8.4) Replace CTS 3.12-2 (4 of 4) Replace CTS 6-15 (2 of 10) Replace I Cover sheet (U1 No Sia Hazards D' tion) Discard 9 (NSHD IIS 3.3.1.1) Rep 1 ace , 3A & 3B (NSHD ITS 3.3.2.1) Add 4 (NSHD ITS 3.3.8.2) Repl ace 2 (NSHD ITS 3.6.1.1) Repl ace 3 (NSHD ITS 3.8.4) Add Cover sheet (U2 Improved Specifications) Discard 3.6-15 Replace 3.8-5 through 3.8-9 Replace , 3.8-15 Replace j 3.8-31 Replace ! 3.8-39 Replace l 3.8-41 Replace
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Cover sheet (U2 Improved Bhses) Discard B 3.3-41 Replace i B 3.3-42A Add (following p. B 3.3-42) l B 3.3-43 Replace ] B 3.4-31 Replace ' B 3.6-1 Replace , B 3.6-7 Replace j B 3.6-27 Replace B 3.6-69 Replace B 3.8-21 Replace Blank (with p. B 3.8-22 on back) Add (following p. B 3.8-21A) B 3.8-23 through 3.8-25 Replace Blank (with p. B 3.8-26 on back) Add (following p. 3.8-25A) B 3.8-55 Replace B 3.8-56A Add B 3.8-61 Replace 2 e
Revision B Insertion Instructions (continued) Pggg Instruction Cover sheet (U2 Improved Bases) (continued) B 3.8-63 Replace B 3.8-64A Add B 3.8-65 Replace Blank (with p. 3.8-66 on back) Add (following p. 3.8-65A) Cover sheet (U2 CTS MarkuD & DOC)N Discard 9 (900 ITS 3.3.1.1) Replace 2 (DOC ITS 3.3.8.2) Replace 3 and 4 (DOC ITS 3.6.1.1) Replace 5 (DOC ITS 3.6.1.1) Add 5 (DOC ITS 3.6.1.3) Replace CTS 3/4 6-9 (1 of 1) Replace 1 (DOC ITS 3.6.1.4) Replace 1A through IF (DOC ITS 3.6.1.4) Add 1 (DOC CTS 3/4.6.1.4) Replace CTS 3/4 8-3a & 3b (4 & 5 of 11) Replace 1 and 2 (DOC ITS 3.8.1) Replace CTS 3/4 8-14 & 15 (2 & 3 of 5) Replace CTS 3/4 8-6 (5 of 5) Replace 4 (DOC ITS 3.8.4) Replace 2 (DOC ITS 3.8.6) Replace 2A (DOC ITS 3.8.6) Add Cover sheet (U2 No Sic Hazards D' tion) Discard 11 (NSHD ITS 3.3.1.1) Replace 4 (NSHD ITS 3.3.8.2) Replace 2 (NSHD ITS 3.6.1.1) Replace 8 (NSHD ITS 3.6.1.3) Replace 1 (NSHD ITS 3.6.1.4) Add 1A (NSHD ITS 3.6.1.4) Add 1 (NSHD CTS 3/4.6.1.4) Replace 4 (NSHD ITS 3.8.4) Replace 4A (NSHD ITS 3.8.4) Add Cover sheet (NUREG 1433 Comparison - SDeCs) Discard 3.6-19 Replace 3.8-3 Replace INSERT Notes 3.8.1.2 Replace INSERT Notes 3.8.1.5 Replace 3.8-9 Replace 3.8-13 Replace 3.8-25 Replace INSERT SRs 3.8.4.7A & 78 Replace 3.8-27 Replace INSERT SR 3.8.4.8 Add 3.8-37 Replace INSERT A/B 3.8.7 Replace
- a. In replacing each CTS page, reference the upper right corner for appropriate ITS section.
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Revision B Insertion Instructions (continued) Paae Instruction Cover sheet (NUREG 1433 Comparison - Bases) Discard B 3.3-43 Replace INSERT AA for Background, B 3.3.2.1 Add B 3.3-45 Replace B 3.6-1 Replace B 3.6-7 Replace B 3.6-33 Repl ace INSERT A (following p. B 3.6-34) Remove B 3.6-81 Replace B 3.6-95 Replace B 3.7-1 Replace B 3.8-17 Repl ace INSERT SR 3.8.1.5 (continued) Replace INSERT SRs 3.8.1.6 & 3.8.1.7 (Unit 1) Replace INSERT SRs 3.8.1.6 & 3.8.1.7 (Unit 2) Replace B 3.8-21 Replace B 3.8-49 Replace INSERT FOR BACKGROUND BASES 3.8.4 Add B 3.8-53 Replace INSERT Action 3.8.4 A/B (Unit 1) Replace INSERT Action 3.8.4 A/B (Unit 1) (cont) Replace INSERT Action 3.8.4 A/B (Unit 2) Replace INSERT Action 3.8.4 A/B (Unit 2) (cont) Replace INSERT SR 3.8.4.1 Replace INSERT SR 3.8.4.2 Replace INSERT SR 3.8.4.4/5 Replace B 3.8-55 Replace INSERT SR 3.8.4.7A Add B 3.8-57 Replace INSERT SR 3.8.4.7 Replace INSERT SR 3.8.4.8A Add INSERT 3.8.4.8 Note Replace Cover sheet (NUREG 1433 J for Deviation) Discard 3 (ITS 3.6) Replace 9 & 10 (ITS 3.6) Replace 3 & 4 (ITS 3.8) Replace 4A (ITS 3.8) Add 7A & 78 (ITS 3.8) Add 8 (ITS 3.8) Replace l l 4
UNIT 1 IMPROVED TECHNICAL SPECIFICATIONS 1 1 I l O l O i l l < L )
PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.6.1.3.3 ------------------NOTES------------------
- 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
- 2. Not required to be met for DCIVs that are open under administrative control s .
Verify each primary containment manual Prior to isolation valve and blind flange that is entering MODE 2 located inside primary containment and is or 3 from required to be closed during accident MODE 4 if conditions is closed. primary containment was de-inerted-while in MODE 4, if not performed O within the previous 92 days SR 3.6.1.3.4 Verify continuity of the traversing 31 days incore probe (TIP) shear isolation valve explosive charge. SR 3.6.1.3.5 Verify the isolation time of each power In accordance operated and each automatic PCIV is with the within limits. Inservice Testing Program (continued) O HATCH UNIT 1 3.6-13 REVISION A t ;
PCIVs 3.6.1.3 l O SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY , SR 3.6.1.3.6 Verify each automatic PCIV, excluding 18 months EFCVs, actuates to the isolation position on an actual or simulated isolation signal. SR 3.6.1.3.7 Verify each reactor instrumentation line 18 months EFCV actuates to restrict flow to within limits. SR 3.6.1.3.8 Remove and test the explosive squio from 18 months on a each shear isolation valve of the TIP STAGGERED TEST system. BASIS O SR 3.6.1.3.9 Verify leakage rate through each MSIV is -----NOTE----- 5 11.5 scfh when tested at 2 28.0 psig. SR 3.0.2 is not applicable. In accordance with 10 CFR 50, Appendix J, as modified by approved exemptions (continued) O HATCH UNIT 1 3.6-14 REVISION
PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.6.1.3.10 Replace the valve seat of each 18 inch 18 months purge valve having a resilient material seat. SR 3.6.1.3.11 Cycle each 18 inch excess flow isolation 18 months damper to the fully closed and fully open position. O O HATCH UNIT 1 3.6-15 REVISION A
Drywell Pressure ) 3.6.1.4 3.6 CONTAINMENT SYSTEMS 3.6.1.4 Drywell Pressure h . LC0 3.6.1.4 Drywell pressure shall be s 1.75 psig. I APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell pressure not A.1 Restore drywell I hour within limit. pressure to within limit. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.4.1 Verify drywell pressure is within limit. 12 hours O HATCH UNIT 1 3.6-16 REVISION,A' i (
AC Sources - Operating , 3.8.1 ACTIONS (continued) f CONDITION REQUIRED ACTION COMPLETION TIME . D. Two or more required D.1 Declare required 12 hours from ' offsite circuits feature (s) with no discovery of inoperable. offsite power Condition D
- available inoperable concurrent with when the redundant inoperability of required feature (s) redundant are inoperable. required '
feature (s) AN_Q ! D.2 Restore all but one 24 hours , required offsite ! circuit to OPERABLE status. r E. One required offsite ------------NOTE------------- circuit inoperable. Enter applicable Conditions O AND and Required Actions of LC0 3.8.7, " Distribution Systems - Operating," when One required DG Condition E is entered with - inoperable. no AC power source to one ' l 4160 V ESF bus. E.1 Restore required 12 hours offsite circuit to , OPERABLE status, j
.1 E.2 Restore required DG 12 hours to OPERABLE status, i F. Two or more (Unit 1 F.1 Restore all but one 2 hours and swing) DGs Unit I and swing DGs inoperable. to OPERABLE status.
(continued) i HATCH UNIT 1 3.8-5 REVISIONf J
AC Sources - Operating 3.8.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action and G.1 Be in MODE 3. 12 hours Associated Completion Time of Condition A, AND B, C, D, E, or F not met. G.2 Be in MODE 4. 36 hours H. One cr more required H.1 Enter LC0 3.0.3. Immediately offsite circuits and two or more required DGs inoperable. 0_R Two or more required offsite circuits and one required DG inoperable. O 1 t ! HATCH UNIT 1 3.8-6 REVISION A t I
. ._ - . .- _.= . -
AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and 7 days indicated power availability for each required offsite circuit. SR 3.8.1.2 -------------------NOTES-------------------
- 1. Performance of SR 3.8.1.5 satisfies this SR.
- 2. All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.
- 3. A modified DG start involving idling and gradual acceleration to synchronous speed may be used for this SR as recommended by the manufacturer. When modified start procedures are not used, the time, voltage, and frequency tolerances of SR 3.8.1.5.a must be met.
- 4. For the swing DG, a single test will satisfy this Surveillance for both units, using the starting circuitry of Unit 1 and synchronized to 4160 V bus IF for one periodic test, and the starting circuitry of Unit 2 and synchronized to 4160 V bus 2F during the next periodic test.
- 5. DG loadings may include gradual loading as recommended by the manufacturer.
- 6. Starting transients above the upper voltage limit do not invalidate this test.
(continued) O HATCH UNIT 1 3.8-7 REVISION ,A
AC Sources - Operating 3.8.1 i SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY l SR 3.8.1.2 NOTES (continued)
- 7. Momentary transients outside the load range do not invalidate this test.
- 8. This Surveillance shall be conducted on only one DG at a time.
Verify each DG: As specified in
- a. Starts from standby conditions and Table 3.8.1-1 achieves steady state voltage 2 3740 V and s 4243 V and frequency 2 58.8 Hz and s 61.2 Hz; and
- b. Operates for 2 60 minutes at a load 2 1710 kW and s 2000 kW.
SR 3.8.1.3 Verify each day tank contains 2 900 gallons 31 days O of fuel oil. SR 3.8.1.4 Check for and remove accumulated water from 184 days each day tank. (continued) O HATCH UNIT 1 3.8-8 REVISION A
AC Sources - Operating 3.8.1 i SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.5 -------------------NOTES-------------------
- 1. All DG starts may be preceded by an engine prelube period.
- 2. DG loadings may include gradual loading as recommended by the manufacturer.
- 3. Momentary load transients outside the load range do not invalidate this test.
- 4. This Surveillance shall be conducted on l only one DG at a time.
- 5. For the swing DG, a single test will l satisfy this Surveillance for both units, using the starting circuitry of Unit 1 and synchronized to 4160 V bus IF for one periodic test and the starting circuitry of Unit 2 and synchronized to 4160 V bus 2F during the next periodic test.
Verify each DG:
- a. Starts from standby conditions and achieves, in s 12 seconds, voltage a 3740 V and frequency 2 58.8 Hz and after steady state conditions are reached, maintains voltage 2 3740 V and 184 days s 4243 V and frequency 2 58.8 Hz and s 61.2 Hz; and l
- b. Operates for 2 60 minutes at a load )
2 2250 kW and s 2400 kW for DGs IA , and IC, and 2 2360 kW and s 2425 kW for i DG 18. l I (continued) ] O HATCH UNIT 1 3.8-9 REVISION /
i AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.6 ------------------NOTE--------------------- This Surveillance shall not be performed in MODE 1 or 2. However, credit may be taken for unplanned events that satisfy this SR. Verify automatic and manual transfer of 18 months unit power supply from the normal offsite circuit to the alternate offsite circuit. SR 3.8.1.7 ------------------NOTES--------------------
- 1. This Surveillance shall not be performed in MODE 1 or 2, except for the swing DG. For the swing DG, this Surveillance shall not be performed in MODE 1 or 2 using the Unit 1 controls.
Credit may be taken for unplanned events that satisfy this SR.
- 2. For the swing DG, a single test at the O specified Frequency will satisfy this Surveillance for both units.
Verify each DG rejects a load greater than 18 months or equal to the single largest post-accident load, and:
- a. Following load rejection, the frequency is s 65.5 Hz; and
- b. Within 3 seconds following load rejection, the voltage is 2: 3740 V and s 4580 V.
l (continued) O HATCH UNIT 1 3.8-10 REVISION
AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) l SURVEILLANCE FREQUENCY l i SR 3.8.1.13 -------------------NOTES------------------- )
- 1. This Surveillance shall be performed '
within 5 minutes of shutting down the DG after the DG has operated 2 2 hours loaded 2 2565 kW. Momentary transients outside of load range do not invalidate this test.
- 2. All DG starts may be preceded by an engine prelube period.
- 3. For the swing DG, a single test at the specified Frequency will satisfy this Surveillance for both units.
Verify each DG starts and achieves, in s 12 seconds, voltage 2 3740 V and frequency 2 58.8 Hz; and after steady state conditions are reached, maintains voltage 18 months 2 3740 V and s 4243 V and frequency O 2 58.8 Hz and s 61.2 Hz. SR 3.8.1.14 -------------------NOTE-------------------- This Surveillance shall not be performed in MODE 1, 2, or 3. However, credit may be taken for unplanned events that satisfy this SR. Verify each DG: 18 months
- a. Synchronizes with offsite power source while loaded with emergency loads upon a simulated restoration of offsite power;
- b. Transfers loads to offsite pcwer source; and
- c. Returns to ready-to-load operation.
(continued) HATCH UNIT 1 3.8-15 REVISION
AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) ; SURVEILLANCE FREQUENCY SR 3.8.1.15 -------------------NOTE-------------------- This Surveillance shall not be aerformed in MODE 1, 2, or 3. However, creait may be taken for unplanned events thas satis fy this SR. Verify with a DG operating in test mode and 18 months connected to its bus, an actual or simulated ECCS initiation signal overrides the test mode by:
- a. Returning DG to ready-to-load operation; and
- b. Automatically energizing the emergency load from offsite power.
SR 3.8.1.16 ------------------NOTE--------------------- This Surveillance shall not be performed in MODE 1, 2, or 3. However, credit may be h taken for unplanned events that satisfy this SR. Verify interval between each sequenced 18 months load block is within 10% of design interval for each load sequence timing device. (continued) O HATCH UNIT 1 3.8-16 REVISION A
DC Sources -- Operating 3.8.4 () SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.4.7 -------------------NOTES-------------------
- 1. The modified performance discharge test in SR 3.8.4.8 may be performed in lieu of the service test in SR 3.8.4.7.
- 2. This Surveillance shall not be performed in MODE 1, 2, or 3, except for the swing DG battery. However, credit may be taken for unplanned events that satisfy this SR.
Verify battery capacity is adequate to 18 months supply, and maintain in OPERABLE status, the required emergency loads for the design duty cycle when subjected to a battery service test. /T SR 3.8.4.8 -------------------NOTE-------------------- \- / This Surveillance shall not be performed in MODE 1, 2, or 3, except for the swing DG battery. However, credit may be taken for unplanned events that satisfy this SR. Verify battery capacity is 2: 80% of the 60 months manufacturer's rating when subjected to a performance discharge test or a modified AND performance discharge test. (continued) l l 1 \s-)
/
HATCH UNIT 1 3.8-31 REVISIONA'/hp 1
DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4,8 (continued) 12 months when battery shows degradation or has reached 85% of expected life with capacity < 100% of manufacturer's rating AND 24 months when battery has reached 85% of expected life with capacity i 2: 100% of manufacturer's rating SR 3.8.4.9 For required Unit 2 DC Sources, the SRs of In accordance Unit 2 Specification 3.8.4 are applicable. with applicable SRs i l W HATCH UNIT 1 3.8-32 REVISION A
i 1 Distribution Systems - Operating 3.8.7 i 3.8 ELECTRICAL POWER SYSTEMS { 3.8.7.' Distribution Systems - Operating ! LC0 3 8.7 The following AC and DC electrical power distribution subsystems shall be OPERABLE:
- a. Unit 1 Division 1 and Division 2 and the swing bus AC and DC electrical power distribution subsystems; and ,
- b. Unit 2 AC and DC electrical power distribution '
subsystems needed to support equipment required to be OPERABLE by LC0 3.6.4.3, " Standby Gas Treatment (SGT)- System," and LC0 3.8.1, "AC Sources-0perating." APPLICABILITY: MODES 1, 2, and 3. , i ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME ,
/7 A. One or more required A.1 Restore required Unit 7 days
() Unit 2 AC or DC electrical power 2 AC and DC subsystem (s) to subsystems inoperable. OPERABLE status. B. One or more (Unit 1 or B.1 Restore DG DC 12 hours swing bus) DG DC electrical power electrical power distribution AND . distribution subsystem to OPERABLE subsystems inoperable. status. 16 hours from discovery of f failure to meet i LC0 3.8.7.a (continued) i O HATCH UNIT 1 3.8-39 REVISION [
Distribution Systems - Operating 3.8.7 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. One or more (Unit 1 or C.1 Restore AC electrical 8 hours swing bus) AC power distribution electrical power subsystem to OPERABLE AND distribution status. subsystems inoperable. 16 hours from l discovery of failure to meet LC0 3.8.7.a D. One Unit 1 station D.1 Restore Unit 1 2 hours service DC electrical station service DC power distribution electrical power AND subsystem inoperable. distribution subsystem to OPERABLE 16 hours from status. discovery of failure to meet LCO 3.8.7.a O E. Required Action and E.1 Be in MODE 3. 12 hours associated Completion Time of Condition A, AND B, C, or D not met. E.2 Be in MODE 4. 36 hours F. Two or more electrical F.1 Enter LCO 3.0.3. Immediately power distribution subsystems inoperable l that result in a loss of function. l i O HATCH UNIT 1 3.8-40 REVISION,Ag l \
1 UNIT I IMPROVED BASES O l O ;
SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.5 and SR 3.3.1.2.6 (continued) REQUIREMENTS Although the Surveillance could be performed while on IRM Range 3, the plant would not be expected to maintain steady state operation at this power level. In this event, the 12 hour Frequency is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the tima required to perform the Surveillances. SR 3.3.1.2.7 Performance of a CHANNEL CALIBRATION at a Frequency of 18 months verifies the performance of the SRM detectors and associated circuitry. The Frequency considers the plant conditions required to perform the test, the ease of performing the test, and the likelihood of a change in the system or component status. The neutron detectors are , excluded from the CHANNEL CALIBRATION (Note 1) because they cannot readily be adjusted. The detectors are fission chambers that are designed to have a relatively constant
, sensitivity over the range and with an accuracy specified
( for a fixed useful life. Note 2 to the Surveillance allows the Surveillance to be delayed until entry into the specified condition of the Applicability. The SR must be performed in MODE 2 within 12 hours of entering MODE 2 with IRMs on Range 2 or below. The allowance to enter the Applicability with the 18 month Frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels. Although the Surveillance could be performed while on IRM Range 3, the plant would not be expected to maintain steady state operation at this power level. In this event, the 12 hour Frequency is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances. REFERENCES 1. NRC Safety Evaluation Report for Amendment 185, April 30, 1993. HATCH UNIT I B 3.3-41 RL'!ISION A
Control Rod Block Instrumentation B 3.3.2.1 B 3.3 INSTRUMENTATION B 3.3.2.1 Control Rod Block Instrumentation BASES BACKGROUND Control rods provide the primary means for control of reactivity changes. Control rod block instrumentation includes channel sensors, logic circuitry, switches, and relays that are designed to ensure that specified fuel design limits are not exceeded for postulated transients and accidents. During high power operation, the rod block monitor (RBM) provides protection for control rod withdrawal error events. During low power operations, control rod blocks from the rod worth minimizer (RWM) enforce specific control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA). During shutdown conditions, control rod blocks from the Reactor Mode Switch - Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent criticalities. The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint during control rod manipulations. It is assumed to function to block further control rod withdrawal to preclude a MCPR Safety Limit (SL) violation. The RBM supplies a trip signal to the Reactor Manual Control System (RMCS) to appropriately inhibit control rod withdrawal during power operation above the low power range setpoint. The RBM has two channels, either of which can initiate a control rod block when the channel output exceeds the control rod block setpoint. One RBM channel inputs into one RMCS rod block circuit and the other RBM channel inputs into the second RMCS rod block circuit. The RBM channel signal is generated by averaging a set of local power range monitor (LPRM) signals at various core heights surrounding the control rod being withdrawn. A signal from one average power range monitor (APRM) channel assigned to each Reactor Protection System (RPS) trip system supplies a reference signal for the RBM channel in the same trip system. Tnis reference signal is used to determine which RBM range setpoint (low, intermediate, or high) is enabled. If the APRM is indicating less than the low power range setpoint, the RBM is automatically bypassed. The RBM is also automatically bypassed if a peripheral control rod is selected (Ref. 1). A rod block signal is also generated if an RBM Downscale trip or an Inoperable trip occurs. The Downscale trip will occur if the RBM channel signal (continued) HATCH UNIT 1 B 3.3-42 REVISION /I
Control Rod Block Instrumentation B 3.3.2.1 BASES BACKGROUND decreases below the Downscale trip setpoint after the RBM (continued) signal has been normalized. The Inoperable trip will occur during the nulling (normalization) sequence, if: the RBM i channel fails to null, too few LPRM inputs are available, a module is not plugged in, or the function switch is moved to any position other than " Operate." The Bypass Time Delay ensures that the normalized signal is passed to the trip logic within the appropriate time. The delay is between the time the signal is nulled to the reference and the signal is passed to the trip logic. 4 O (continued) HATCH UNIT 1 B 3.3-f3 Q A REVISION [ l
Control Rod Block Instrumentation B 3.3.2.1 BASES BACKGROUND The purpose of the RWM is to control rod patterns during (continued) startup and shutdown, such that only specified con +,rol rod sequences and relative positions are allowed over the operating range from all control rods inserted to 10% RTP. The sequences effectively limit the potential amount and rate of reactivity increase during a CRDA. Prescribed control rod sequences are stored in the RWM, which will initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the stored sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses - feedwater flow and steam flow signals to determine when the reactor power is above the preset power level at which the RWM is automatically bypassed (Ref. 2). The RWM is a single channel system that provides input into both RMCS rod block circuits. With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This Function prevents inadvertent criticality as the result of a control rod withdrawal during MODE 3 or 4, or during MODE 5 when the reactor mode switch is required to be in the O shutdown position. The reactor mode switch has two channels, each inputting into a separate RMCS rod block circuit. A rod block in either RMCS circuit will provide a control rod block to all control rods. APPLICABLE 1. Rod Block Monitor SAFETY ANALYSES, LCO, and The RBM is designed to prevent violation of the MCPR APPLICABILITY SL and the cladding 1% plastic strain fuel design limit that may result from a single control rod withdrawal error (RWE) event. The analytical methods and assumptions used in evaluating the RWE event are summarized in Reference 3. A statistical analysis of RWE events was performed to determine the RBM response for both channels for each event. From these responses, the fuel thermal performance as a function of RBM Allowable Value was determined. The Allowable Values are chosen as a function of power level. Based on the specified Allowable Values, operating limits are established. (continued) HATCH UNIT 1 B 3.3-43 REVISION A l
.J
Control Rod Block Instrumentation 5 3.3.2.1 BASES h APPLICABLE 1. Rod Block Monitor (continued) SAFETY ANALYSES, LCO, and The RBM Function satisfies Criterion 3 of the f;RC Policy APPLICABILITY Statement (Ref. 9). Two channels of the RBM are required to be OPERABLE, with their setpoints within the appropriate Allowable Values, to ensure that no single instrument failure can preclude a rod block from this function. The setpoints are calibrated consistent with applicable setpoint methodology (nominal trip setpoint). Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor power), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived g from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for. The RBM is assumed to mitigate the consequences of an RWE event when operating 2 29% RTP. Below this power level, the consequences of an RWE event will not exceed the MCPR SL and, therefore, the RBM is not required to be OPERABLE (Ref. 3). When operating < 90% RTP, analyses (Ref. 3) have shown that with an initial MCPR 21.70, no RWE event will result in exceeding the MCPR SL. Also, the analyses demonstrate that when operating at 2 90% RTP with MCPR 21.40, no RWE event will result in exceeding the MCPR (continued) HATCH UNIT 1 B 3.3-44 REVISIONf(
Primary Containment B 3.6.1.1 l l B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment BASES l BACKGROUND The function of the primary containment is to isolate and contain fission products released from the Reactor Primary System following a Design Basis Accident (DBA) and to confine the postulated release of radioactive material. The primary containment consists of a steel lined, reinforced concrete vessel, which surrounds the Reactor Primary System and provides an essentially leak tight barrier against an uncontrolled release of radioactive material to the environment. The isolation devices for the penetrations in the primary containment boundary are a part of the containment leak tight barrier. To maintain this leak tight barrier:
- a. All penetrations required to be closed during accident conditions are either:
- l. capable of being closed by an OPERABLE automatic x containment isolation system, or
- 2. closed by manual valves, blind flanges, or de-activated automatic valves secured in their closed positions, except as provided in LC0 3.6.1.3, " Primary Containment Isolation Valves (PCIVs);"
- b. The primary containment air lock is OPERABLE, except as provided in LC0 3.6.1.2, " Primary Containment Air Lock"; and
- c. All equipment hatches are closed.
This Specification ensures that the performance of the primary containment, in the event of a DBA, meets the assumptions used in the safety analyses of References 1 and 2. SR 3.6.1.1.1 leakage rate requi rements are in conformance with 10 CFR 50, Appendix .' (Ref. 3), as modified by approved exemptions. (continued) HATCH UNIT 1 B 3.6-1 REVISION A l
1 Primary Containment B 3.6.1.1 l BASES (continued) h APPLICABLE The safety design basis for the primary containment is that l SAFETY ANALYSES it must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate. The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage. Analytical methods and assumptions involving the primary containment are presented in References 1 and 2. The safety analyses assume a nonmechanistic fission product release following a DBA, which forms the basis for determination of offsite doses. The fission product release is, in turn, based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not exceeded. The maximum allowable leakage rate for the primary containment (L is 1.2% by weight of the containment air per 24 hours al) the maximum peak containment pressure (P,) of 53.6 psig (Ref. 1). l Primary containment satisfies Criterion 3 of the NRC Policy Statement (Ref. 4). LC0 Primary containment OPERABILITY is maintained by limiting leakage to less than L., except prior to the first startup after performing a required 10 CFR 50, Appendix J, leakage test. At this time, the combined Type B and C leakage must be < 0.6 L,, and the overall Type A leakage must be < 0.75 L,. Compliance with this LC0 will ensure a primary containment configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analyses. Individual leakage rates specified for the primary containment air lock are addressed in LCO 3.6.1.2. (continued) HATCH UNIT 1 B 3.6-2 REVISION
l Primary Containment Air Lock B 3.6.1.2 BASES I BACKGROUND containment leakage rate to within limits in the event of a l (continued) DBA. Not maintaining air lock integrity or leak tightness 1 may result in a leakage rate in excess of that assumed in the unit safety analysis. l APPLICABLE The DBA that postulates the maximum release of radioactive SAFETY ANALYSES material within primary containment is a LOCA. In the : analysis of this accident, it is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment is designed with a maximum allowable leakage rate (L ) of 1.2% byweightofthecontainmentairper24hoursatIhe calculated maximum peak containment pressure (P,) of 53.6 psig (Ref. 2). This allowable leakage rate forms the l basis for the acceptance criteria imposed on the SRs associated with the air lock. Primary containnient air lock OPERABILITY is also required to minimize the amount of fission product gases that may escape primary containment through the air lock and contaminate and O pressuri7e the secondary containment. The primary containment air lock satisfies Criterion 3 of the NRC Policy Statement (Ref. 4). LC0 As part of primary containment, the air lock's safety function is related to control of containment leakage rates i following a DBA. Thus, the air lock's structural integrity j and leak tightness are essential to the successful mitigation of such an event. - The primary containment air lock is required to be OPERABLE. i for the air lock to be considered OPERABLE, the air lock iaterlock mechanism must be OPERABLE, the air lock must be - in compliance with the Type B air lock leakage test, and : both air lock doors must be OPERABLE. The interlock allows ! cely one air lock door to be opened at a time. TLis , provision ensures that a gross breach of primary containment
- does not exist when primary containment is required to be i
(continued) HATCH UNIT 1 B 3.6-7 REVISION A 1
1 Primary Containment Air Lock B 3.6.1.2 BASES h LC0 OPERABLE. Closure of a single door in each air lock is (continued) sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry and exit from primary containment. APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the primary containment air lock is not required to be OPERABLE in MODES 4 and 5 to prevent leakage of radioactive material from primary containment. ACTIONS The ACTIONS are modified by Note 1, which allows entry and exit to perform repairs of the affected air lock component. If the outer door is inoperable, then it may be easily accessed to repair. If the inner door is the one that is inoperable, however, then a short time exists when the containment boundary is not intact (during access through the outer door). The allowance to open the OPERABLE door, even if it means the primary containment boundary is temporarily not intact, is acceptable due to the low probability of an event that could pressurize the primary containment during the short time in which the OPERABLE door is expected to be open. The OPERABLE door must be immediately closed after each entry and exit. The ACTIONS are modified by a second Note, which ensures appropriate remedial measures are taken, if necessary, if air lock leakage results in exceeding overall containment leakage rate acceptance criteria. Pursuant to LC0 3.0.6, actions are not required, even if primary containment is exceeding its leakage limit. Therefore, the Note is added to require ACTIONS for LCO 3.6.1.1, " Primary Containment," to be taken in this event. i (continued) HATCH UNIT 1 B 3.6-8 REVISION A
PCIVs B 3.6.1.3 BASES REFERENCES 4. 10 CFR 50, Appendix J. (continued)
- 5. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
O O HATCH UNIT 1 B 3.6-27 REVISION A
Drywell Pressure d 3.6.1.4 8 3.6 CONTAINMENT SYSTEMS B 3.6.1.4 Drywell Pressure BASES BACKGROUND The drywell pressure is limited during normal operations to preserve the initial conditions assumed in the accident analysis for a Design Basis Accident (DBA) or loss of coolant accident (LOCA). APPLICABLE Primary containment performance is evaluated for the entire SAFETY ANALYSES spectrum of break sizes for postulated LOCAs (Ref.1). Among the inputs to the DBA is the initial primary containment internal pressure (Ref.1). Analyses assume an initial drywell pressure of 1.75 psig. This limitation ( ensures that the safety analysis remains valid by maintaining the expected initial conditions and ensures that the peak LOCA drywell internal pressure does not exceed the maximum allowable of 62 psig. The maximum calculated drywell pressure occurs during the reactor blowdown phase of the DBA, which assumes an instantaneous recirculation line break. The calculated peak drywell pressure for this limiting event is 53.6 psig l (Ref. 1). Drywell pressure satisfies Criterion 2 of the NRC Policy Statement (Ref. 2). LC0 In the event of a DBA, with an initial drywell pressure s 1.75 psig, the resultant peak drywell accident pressure l will be maintained below the drywell design pressure. APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 (continued) HATCH UNIT I B 3.6-26 REVISIONA
i CAD System ! B 3.6.3.1 G Q BASES (continued) ACTIONS L1 If one CAD subsystem is inoperable, it must be restored to OPERABLE status within 30 days. In this Condition, the remaining OPERABLE CAD subsystem is adequate to perform the oxygen control function. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in reduced oxygen control capability. The 30 day Completion Time is based on the low probability of the occurrence of a LOCA that would generate hydrogen and oxygen in amounts capable of exceeding the flammability limit, the amount of time available after the event for operator action to prevent exceeding this limit, and the availability of the OPERABLE CAD subsystem and other hydrogen mitigating systems. Required Action A.1 has been modified by a Note that indicates that the provisions of LCO 3.0.4 are not applicable. As a result, a MODE change is allowed when one CAD subsystem is inoperable. Th O allowance is provided because of the low probability of the occurrence of a LOCA that would generate hydrogen hnd oxygen in amounts capable of exceeding the flammability limit, the low probability of O- the failure of the OPERABLE subsystem, the amount of time available after a postulated LOCA for operator action to prevent exceeding the flammability limit, and the availability of other hydrogen mitigating systems. B.1 and B.2 With two CAD subsystems inoperable, the ability to perform the hydrogen control function via alternate capabilities must be verified by administrative means within I hour. The alternate hydrogen control cg abilities are provided by the Primary Containment Purge System. The 1 hour Completion Time allows a reasonable period of time to verify that a loss of hydrogen control function does not exist. In ' addition, the alternate hydrogen control system capability must be verified once per 12 hours thereafter to ensure its l continued availability. Both the initial verification and 4 all subsequent verifications may be performed as an administrative check by examining logs or other information (continued) { HATCH UNIT 1 B 3.6-69 REVISION A
CAD System B 3.6.3.1 BASES h ACTIONS B.1 and B.2 (continued) to determine the availability of the alternate hydrogen control system. It does not mean to perform the Surveillances needed to demonstrate OPERABILITY of the alternate hydrogen control system. If the ability to perform the hydrogen control function is maintained, continued operation is permitted with two CAD subsystems inoperable for up to 7 days. Seven days is a reasonable time to allow two CAD subsystems to be inoperable because the hydrogen control function is maintained and because of the low probability of the occurrence of a LOCA that would gererate hydrogen in amounts capable of exceeding the l flammability limit. i With two CAD subsystems inoperable, one CAD subsystem must be restored to OPERABLE status within 7 days. The 7 day Completion Time is based on the low probability of the occurrence of a LOCA that would generate hydrogen in the amounts capable of exceeding the flammability limit, the amount of time available after the event for operator action ! to prevent exceeding this limit, and the availability of other hydrogen mitigating systems. I C.1 If any Required Action cannot be met within the associated Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the I plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging l plant systems. SURVEILLANCE SR 3.6.3.1.1 REQUIREMENTS Verifying that there is a 2000 gallons of liquid nitrogen supply in each Nitrogen Storage Tank will ensure at least 7 days of post-LOCA CAD operation. This minimum volume of liquid nitrogen allows sufficient time after an accident to replenish the nitrogen supply for long term inerting. This is verified every 31 days to ensure that each subsystem is (continued) HATCH UNIT 1 B 3.6-70 REVISION A
i RHRSW System l 1 B 3.7.1 ' B 3.7 PLANT SYSTEMS i B 3.7.1 Residual Heat Removal Service Water (RHRSW) System l l BASES
)
BACKGROUND The RHRSW System is designed to provide cooling water for the Residual Heat Removal (RHR) System heat exchangers, required for a safe reactor shutdown following a Design Basis Accident (DBA) or transient. The RHRSW System is ; operated whenever the RHR heat exchangers are required to e operate in the shutdown cooling mode or in the suppression pool cooling or spray mode of the RHR System. , The RHRSW System consists of two independent and redundant - subsystems. Each subsystem is made up of a header, two i 4000 gpm pumps, a suction source, valves, piping, heat , exchanger, and associated instrumentation. Either of the - two subsystems is capable of providing the required cooling capacity with two pumps operating to maintain safe shutdown conditions. The two subsystems are separated from each other by normally closed motor operated cross tie valves, so that failure of one subsystem will not affect the O OPERABILITY of the other subsystem. The RHRSW System is designed with sufficient redundancy so that no single active component failure can prevent it from achieving its design , function. The RHRSW System is described in the FSAR, ! Section 10.6, Reference 1. i Cooling water is pumped by the RHRSW pumps from the Altamaha River through the tube side of the RHR heat exchangers, and discharges to the circulating water flume. A minimum flow line from the pump discharge to the intake structure . prevents the pump from overheating when pumping against a ; closed discharge valve. The system is initiated manually from the control room. If operating during a loss of coolant accident (LOCA) or a loss of offsite power (LOSP), the system is automatically tripned to allow the diesel generators to automatically power only , that equipment necessary. The system can be manually i started any time the LOCA signal is manually overridden or , clears. The system can be manually started any time after ; the LOSP signal is received. O _(continued) HATCH UNIT 1 B 3.7-1 REVISION A
- - - - ._____________j
RHRSW System B 3.7.1 BASES (continued) h APPLICABLE The RHRSW System removes heat from the suppression pool to SAFETY ANALYSES limit the suppression pool temperature and primary containment pressure following a LOCA. This ensures that the primary containment can perform its function of limiting the release of radioactive materials to the environment following a LOCA. The ability of the RHRSW System to support long term cooling of the reactor or primary containment is discussed in the FSAR, Sections 5.2 and 14.4.3 (Refs. 2 and 3, respectively). These analyses explicitly assume that the RHRSW System will provide adequate cooling support to the equipment required for safe shutdown. These analyses include the evaluation of the long term primary containment response after a design basis LOCA. The safety analyses for long term cooling were performed for various combinations of RHR System failures. The worst case single failure that would affect the performance of the RHRSW System is any failure that would disable one subsystem of the RHRSW System. As discussed in the FSAR, Section 14.4.3 (Ref. 3) for these analyses, manual initiation of the OPERABLE RHRSW subsystem and the associated RHR System is assumed to occur 10 minutes after a DBA. The RHRSW flow assumed ir. the analyses is 4000 gpm per pump with two pumps operating in one loop. In this case, ! the maximum suppression chamber water temperature and pressure are approximately 210 F and 15 psig, respectively, well below the design temperature of 281 F and maximum l allowable pressure of 62 psig. The RHRSW System satisfies Criterion 3 of the NRC Policy Statement (Ref. 4). l LC0 Two RHRSW subsystems are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming the worst case l single active failure occurs coincident with the loss of ! offsite power. l l l l (continued) HATCH UNIT 1 B 3.7-2 REVISION
AC Sources-0perating B 3.8.1 b v BASES SURVEILLANCE SR 3.8.1.2 REQUIREMENTS (continued) This SR helps to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and maintain the unit in a safe shutdown condition, and verifies that the DGs are capable of proper startup, synchronizing, and accepting a load approximately 50% of the continuous load rating. This demonstrates DG capability while minimizing the mechanical stress and wear on the engine. A minimum run time of 60 minutes is required to stabilize engine temperatures, while minimizing the time that the DG is connected to the offsite source. Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.8 lagging and 1.0. The 0.8 value is the design rating of the machine, while 1.0 is an operational limitation. To minimize the wear on moving parts that do not get lubricated when the engine is not running, this SR has been modified by a Note (Note 2) to indicate that all DG starts (~% for this Surveillance may be preceded by an engine prelube () period and followed by a warmup prior to loading. For the purposes of this testing, the DGs are started from standby conditions. Standby conditions for a DG mean that the diesel engine coolant and oil are being continuously circulated and temperature is being maintained consistent with manufacturer recommendations. In order to reduce stress and wear on diesel engines, the DG manufacturer recommends a modified start in which the starting speed of DGs is limited, warmup is limited to this lower speed, and the DGs are gradually accelerated to synchronous speed prior to loading. These start procedures are the intent of Note 3. SR 3.8.1.5.a requires that, at a 184 day Frequency, the CG starts from standby conditions and achieves required voltage and frequency within 12 seconds. The 12 second start requirement supports the assumptions in the design basis LOCA analysis of FSAR, Chapter 6 (Ref. 4). The 12 second start requirement is not applicable to SR 3.8.1.2 (see Note 3), when a modified start procedure as described above is used. If a modified start is not used, the O (continued) O HATCH UNIT 1 B 3.8-19 REVISION A
1 i AC Sources-Operating l B 3.8.1 BASES SURVEILLANCE SR 3.8.1.2 (continued) REQUIREMENTS 12 second start voltage and frequency requirements of SR 3.8.1.5.a apply. Since SR 3.8.1.5.a does require a 12 second start, it is more restrictive than SR 3.8.1.2, and it may be performed in lieu of SR 3.8.1.2. This procedure is the intent of Note 1. To minimize testing of the swing DG, this SR is modified by a note (Note 4) to allow a single test (instead of two tests, one for each unit) to satisfy the requirements for both units, using the starting circuitry of one unit for one periodic test and the starting circuitry of the other unit during the next periodic test. This is allowed since the main purpose of the Surveillance, to ensure DG OPERABILITY, is still being verified on the proper frequency, the starting circuits historically have a very low failure rate, as compared to the DG itself, and that, while each starting circuit is only being tested every second test (due to the staggering of the tests), some portions of the starting circuits are common to both units. If the swing DG fails one of these Surveillance, the DG should be considered inoperable on both units, unles.: the cause of the failure can be directly related to only one unit. Note 5 modifies this Surveillance to indicate that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized. Note 6 modifies the Surveillance by stating that starting transients above the upper voltage limit do not invalidate this test. Note 7 modifies this Surveillance by stating that momentary l load transients because of changing bus loads do not invalidate this test. Note 8 indicates that this Surveillance is required to be conducted on only one DG at a time in order to avoid common l cause failures that might result from offsite circuit or I grid perturbations. l l (continued) 1 HATCH UNIT I B 3.8-20 REVISIONA
AC Sources-Operating B 3.8.1
) BASES SURVEILLANCE SR 3.8.1.2 (continued)
REQUIREMENTS The normal 31 day Frequency for SR 3.8.1.2 (see Table 3.8.1-1, " Diesel Generator Test Schedule") is consistent with Regulatory Guide 1.108 (Ref. 10). This Frequency provides adequate assurance of DG OPERABILITY, while minimizing degradation resulting from testing. O l l (continued) 1 HATCH UNIT I B 3.8-20A REVISION B l
l l AC Sources-Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.5 (continued) I REQUIREMENTS Note 3 mcdifies this Surveillance by stating that momentary voltage or load transients because of changing bus loads do not invalidate this test. Note 4 indicates that this Surveillance is required to be l conducted on only one DG at a time in order to avoid common cause fcilures that might result from offsite circuit or grid perturbations. To minimize testing of the swing DG, Note 5 allows a single l test (instead of two tests, one for each unit) to satisfy the requirements for both units, with the DG started using the starting circuitry of one unit and synchronized to the ESF bus of that unit for one periodic test and started using the starting circuitry of the other unit and synchronized to the ESF bus of that unit during the next periodic test. This is allowed since the main purpose of the Surveillance, to ensure DG OPERABILITY, is still being verified on the proper frequency, and each unit's starting circuitry and breaker control circuitry, which is only being tested every second test (due to the staggering of the tests), ,gD historically have a very low failure rate. If the swing DG V fails one of these Surveillances, the DG should be considered inoperable on both units, unless the cause of the failure can be directly related to only one unit. SR 3.8.1.6 Transfer of each 4.16 kV ESF bus power supply from the normal offsite circuit to the alternate offsite circuit demonstrates the OPERABILITY of the alternate circuit distribution network to power the shutdown loads. The 18 month Frequency of the Surveillance is based on engineering judgment taking into consideration the plant conditions required to perform the Surveillance, and is intended to be consistent with expected fuel cycle lengths. Operating experience has shown that these components usually pass the SR when performed on the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
,r x (continued)
HATCH UNIT 1 REVISION B3.8-2[3
AC Sources-Operating B 3.8.1 BASES h SURVEILLANCE SR 3.8.1.6 (continued) REQUIREMENTS This SR is modified by a Note. The reason for the Note is that, during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, plant safety systems. Credit may be taken for unplanned events that satisfy this SR. This Surveillance tests the applicable logic associated with the Unit I swing bus. The comparable test specified in the Unit 2 Technical Specifications tests the applicable logic associated with the Unit 2 swing bus. Consequently, a test must be performed within the specified Frequency for each unit. The Note specifying the restriction for not performing the test while the unit is in MODE 1 or 2 does not have applicability to Unit 2. As the Surveillance represents separate tests, the Unit 1 Surveillance should not te performed with Unit 1 in MODE I or 2 and the Unit 2 test should not be performed with Unit 2 in NODE 1 or 2. SR 3.8.1.7 Each DG is provided with an engine overspeed trip to prevent damage to the engine. Recovery from the transient caused by the loss of a large load could cause diesel engine overspeed, which, if excessive, might result in a trip of the engine. This Surveillance demonstrates the DG load response characteristics and capability to reject the largest single load without exceeding predetermined voltage and frequency and while maintaining a specified margin to the overspeed trip. The largest single load for DGs IA and IC is a core spray pump at rated flow (1275 bhp). For DG 1B, the largest single load is a residual heat removal service water pump at rated flow (1225 bhp). This Surveillance may be accomplished by either a.) tripping the DG output breaker with the DG carrying greater than or equal to the largest single load while paralleled to offsite power or while solely supplying the bus, or b.) tripping the largest single load with the DG solely supplying the bus. Although Plant Hatch Unit 1 is not committed to IEEE-387-1984 (Ref. 12), this SR is consistent with the IEEE-387-1984 requirement that states the load rejection test is acceptable if the increase in diesel speed does not exceed 75% of the difference between synchronous speed and the (continued) HATCH UNIT 1 B3.8-2/4 REVISION [
l AC Sources-Operating B 3.8.1 l t s BASES SURVEILLANCE SR 3.8.1.7 (continued) REQUIREMENTS overspeed trip setpoint, or 15% above synchronous speed, whichever is lower. For all DGs, this represents 65.5 Hz, equivalent to 75% of the difference between nominal speed and the overspeed trip setpoint. O (continued) HATCH UNIT 1 B 3.8-24A REVISION A'
AC Sources-Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.7 (continued) REQUIREMENTS The voltage and frequency specified are consistent with the nominal range for the DG. SR 3.8.1.7.a corresponds to the maximum frequency excursion, while SR 3.8.1.7.b is the voltage to which the DG must recover following load rejection. The 18 month Frequency is consistent with the recommendation of Regulatory Guide 1.108 (Ref.10). This SR is modified by two Notes. The reason for Note 1 is that, during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, plant safety systems. Credit may be taken for unplanned events that satisfy this SR. In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as possible, testing is performed with only the DG providing power to the associated 4160 V ESF bus. The DG is not synchronized with offsite power. To minimize testing of the swing DG, Note 2 allows a single q test (instead of two tests, one for each unit) to satisfy the requirements for both units. This is allowed since the main purpose of the Surveillance can be met by performing the test on either unit (no unit specific DG components are l being tested . If the swing DG fails one of these the DG should be considered inoperable on l Surveillances),less both units, un the cause of the failure can be di related to only one unit. SR 3.8.1.8 This Surveillance demonstrates the DG capability to reject a full load without overspeed tripping or exceeding the i predetermined voltage limits. The DG full load rejection l may occur because of a system fault or inadvertent breaker tripping. This Surveillance ensures proper engine generator load response under the simulated test conditions. This test simulates the loss of the total connected load that the DG experiences following a full load rejection and verifies that the DG does not trip upon loss of the load. These l f (continued) HATCH UNIT 1 B3.8-2/3 REVISIONp' bL
,. ~
m AC Sources-Operating B 3.8.1 BASES h SURVEILLANCE SR 3.8.1.8 (continued) REQUIREMENTS acceptance criteria provide DG damage protection. While the DG is not expected to experience this transient during an event, and continues to be available, this response ensures that the DG is not degraded for future application, L O (continued) HATCH UNIT 1 8 3.8-2p S , REVISION [
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AC Sources-Operating B 3.8.1 BASES h SURVEILLANCE SR 3.8.1.8 (continued) REQUIREMENTS including reconnection to the bus if the trip initiator can be corrected or isolated. In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as possible, testing must be performed using a power factor s 0.88. This power factor is chosen to be representative of the actual design basis inductive loading that the DG would experience. The 18 month Frequency is consistent with the recommendation of Regulatory Guide 1.108 (Ref.10) and is intended to be , consistent with expected fuel cycle lengths. 1 This SR is modified by four Notes. Note 1 states that momentary transients due to changing bus loads do not invalidate this test. The reason for Note 2 is that during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that would challenge continued steady state operation and, as a result, plant safety systems. Credit may be taken for unplanned events that satisfy this SR. Note 3 is provided in recognition that if the offsite electrical power distribution system is lightly loaded (i.e., system voltages are high), it may not be possible to raise voltage without creating an overvoltage condition on the ESF bus. Therefore, to ensure the bus voltage, supplied ESF loads, and DG are not placed in an unsafe condition during this test, the power factor limit does not have to be met if grid voltage or ESF bus loading does not permit the power factor limit to be met when the DG is tied to the grid. When this occurs, the power factor should be maintained as close to the limit as practicable. To minimize testing of the swing DG, Note 4 allows a single test (instead of two tests, one for each unit) to satisfy the requirements for both units. This is allowed since the main purpose of the Surveillance can be met by performing the test on either unit (no unit specific DG components are being tested). If the swing DG fails one of these Surveillances, the DG should be considered inoperable on both units, unless the cause of the failure can be directly related to only one unit. (continued) HATCH VNIT 1 B 3.8-26 REVISION A
I Diesel Fuel Oil and Transfer, Lube Oil, and Starting Air ) B 3.8.3 BASES SURVEILLANCE SR 3.8.3.7 (continued) REQUIREMENTS system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for manual fuel transfer are OPERABLE. Since the fuel oil transfer pumps are being tested on a 31 day Frequency in accordance with SR 3.8.3.5, the 18 month Frequency has been determined to be acceptable based on engineering judgement and operating experience. REFERENCES 1. FSAR, Section 8.4.
- 2. FSAR, Chapters 5 and 6.
- 3. FSAR, Chapter 14.
- 4. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
O V A V HATCH UNIT 1 B 3.8-55 REVISION A
DC Sources - Operating B 3.8.4 8 3.8 ELECTRICAL POWER SYSTEMS 8 3.8.4 DC Sources - Operating h BASES BACKGROUND The DC electrical power system provides the AC emergency power system with control power. It also provides both motive and control power to selected safety related equipment. As required by 10 CFR 50, Appendix A, GDC 17 (Ref.1), the DC electrical power system is designed to have sufficient independence, redundancy, and testability to perform its safety functions, assuming a single failure. The DC electrical power system also conforms to the recommendations of Regulatory Guide 1.6 (Ref. 2) and IEEE-308 (Ref. 3). The station service DC power sources provide both motive and control power to selected safety related and nonsafety related equipment. Each DC subsystem is energized by one 125/250 V station service battery and three 125 V battery chargers (two normally inservice chargers and one standby charger). Each battery is exclusively associated with a single 125/250 VDC bus. Each set of battery chargers exclusively associated with a 125/250 VDC subsystem cannot be interconnected with any other 125/250 VDC subsystem. The normal and backup chargers are supplied from the same AC load groups for which the associated DC subsystem supplies the control power. The loads between the redundant 125/250 VDC subsystem are not transferable except for the Automatic Depressurization System, the logic circuits and valves of which are normally fed from the Division 1 DC system. The diesel generator (DG) DC power sources provide control and instrumentation power for their respective DG and their respective offsite circuit supply breakers. In addition, DG 1A power source provides circuit breaker control power for the respective Division I loads on 4160 VAC buses IE and IF, and DG IC power source provides circuit breaker control power for the respective Division II loads on 4160 VAC buses IF and IG. Each DG DC subsystem is energized by one 125 V battery and two 125 V battery chargers (one normally inservice charger and one standby charger). During normal operation, the DC loads are powered from the respective station service and DG battery chargers rith the batteries floating on the system. (continued) HATCH UNIT 1 REVISION A B3.8-59h
DC Sources-Operating i B 3.8.4 i BASES ACTIONS _(,d (continued) If one of the required DC electrical power subsystems is inoperable (e.g., inoperable battery, inoperable battery charger (s), or inoperable battery charger and associated ' inoperable battery), the remaining DC electrical power subsystems have the capacity to support a safe shutdown and to mitigate an accident condition. Since a subsequent postulated worst case single failure could result in the lors of minimum necessary DC electrical subsystems to , mitigate a postulated worst case accident, continued power operation should not exceed 2 hours. The 2 hour Completion Time is based on Regulatory Guide 1.93 (Ref. 7) and reflects a reasonable time to assess unit status as a function of the , inoperable DC electrical power subsystem and, if the DC electrical power subsystem is not restored to OPERABLE , status, to prepare to effect an orderly and safe unit shutdown. D.1 and D.2 O If the DC electrical power subsystem cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not 1 apply. To achieve this status, the unit must be brought to at least MODE 3 within 12 hours and to MODE 4 within - 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. The Completion Time !
~
to bring the unit to MODE 4 is consistent with the time required in Regulatory Guide 1.93 (Ref. 7). { L1 , Condition E corresponds to a level of degradation in the DC electrical power subsystems that causes a required safety function to be lost. When more than one DC source is lost, and this results in the loss of a required function, the plant is in a condition outside the accident analysis. ; Therefore, no additional time is justified for continued l LC0 3.0.3 must be entered immediately to operation. commence a controlled shutdown. i O (continued) HATCH UNIT 1 B 3.8-61 REVISION A
OC Sources - Operating 8 3.8.4 BASES (continued) SURVEILLANCE SR 3.8.4.1 REQUIREMENTS Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended function. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery (or battery cell) and maintain the battery (or a battery cell) in a fully charged state. Voltage requirements are based on the nominal design voltage of the battery and are consistent with the initial voltages assumed in the battery sizing calculations. The voltage requirement for battery terminal voltage is based on the open circuit voltage of a lead-calcium cell of nominal 1.215 specific gravity. Without regard to other battery parameters, this voltage is indicative of a battery that is capable of performing its required safety function. The 7 day Frequency is consistent with manufacturer's recommendations and IEEE-450 (Ref. 8). SR 3.8.4.2 l Visual inspection to detect corrosion of the battery cells and connections, or measurement of the resistance of each inter-cell, inter-rack, inter-tier, and terminal connection, h l provides an indication of physical damage or abnormal deterioration that could potentially degrade battery per#ormance. fne connection resistance limits are established to maintain connection resistance as low as reasonably possible to minimize the overall voltage drop across the battery and the l possibility of battery damage due to heating of connections. i The resistance values for each battery connection are l located in the Technical Requirements Manual (Reference 9). l The Frequency for these inspections, which can detect conditions that can cause power losses due to resistance heating, is 92 days. This Frequency is considered acceptable based on operating experience related to detecting corrosion trends. 1 SR 3.8.4.3 Visual inspection of the battery cells, cell plates, and battery racks provides an indication of physical damage or (continued) HATCH UNIT 1 B3.8-6h REVISIONgg 1
DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.3 (continued) REQUIREMENTS abnormal deterioration that could potentially degrade battery performance. The 18 month Frequency of the Surveillance is based on engineering judgment, taking into consideration the desired plant conditions to perform the Surveillance. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. SR 3.8.4.4 and SR 3.8.4.5 Visual inspection and resistance measurements of inter-cell, inter-rack, inter-tier, and terminal connections provides an indication of physical damage or abnormal deterioration that could indicate degraded battery condition. The anti-corrosion material is used to help ensure good electrical connections and to reduce terminal deterioration. The visual inspection for corrosion is not intended to ' require removal of and inspection under each terminal connection. The removal of visible corrosion is a preventive maintenance SR. The presence of visible corrosion does not necessarily represent a failure of this SR, provided visible corrosion is removed during performance of this Surveillance. The connection resistance limits are established to maintain connection resistance as low as reasonably possible to minimize the overall voltage drop across the battery and the possibility of battery damage due to heating of connections. The resistance values for each battery connection are located in the Technical Requirements Manual (Reference 9). The 18 month Frequency of the Surveillances is based on engineering judgment, taking into consideration the desired plant conditions to perform the Surveillance. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. (continued) HATCH UNIT 1 B 3.8-63 REVISION A
DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.6 REQUIREMENTS (continued) Battery charger capability requirements are based on the design capacity of the chargers (Ref. 4). According to Regulatory Guide 1.32 (Ref. 10), each battery charger supply is required to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences. The minimum required amperes and duration ensures that these requirements can be satisfied. The Frequency is acceptable, given the unit conditions required to perform the test and the other administrative controls existing to ensure adequate charger performance curing these 18 month intervals. In addition, this Frequency is intended to be consistent with expected fuel cycle lengths. SR 3.8.4.7 A battery service test is a special test of the battery's capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The h discharge rate and test length corresponds to the design duty cycle requirements as specified in Reference 4. The Frequency of 18 months is consistent with the recommendations of Regulatory Guide 1.32 (Ref. 10) and Regulatory Guide 1.129 (Ref.11), which state that the battery service test should be performed during refueling operations or at some other outage, with intervals between tests not to exceed 18 months. This SR is modified by two Notes. Note 1 allows the performance of a modified performance discharge test in lieu of a service test. The modified performance discharge test is a simulated duty cycle consisting of just two rates: the 1 minute rate published for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test, both of which envelope the duty cycle of the service test. Since the ampere-hours removed by a rated (continued) HATCH UNIT 1 B3.8-6% REVISION Q
DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.7 (continued) REQUIREMENTS 1 minute discharge represent a very small portion of the battery capacity, the test rate can be changed to that for the performance test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test should remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test. ; A modified performance discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity. Initial - conditions for the modified performance discharge test i should be identical to those specified for a service discharge test. The reason for Note 2 is that performing the Surveillance would remove a required DC electrical power subsystem from - O service, perturb the electrical distribution system, and challenge safety systems. Credit may be taken for unplanned events that satisfy the Surveillance. The swing DG DC battery is exempted from this restriction, since t i (continued) ! HATCH _ UNIT 1 B3.8-6,9%A REVISION
DC Sources - Operating B 3.8.4
) BASES SURVEILLANCE SR 3.8.4.7 (cor.tinued)
REQUIREMNETS it is required by both units' LCO 3.8.4 and cannot be performed in the manner required by the Note without resulting in a dual unit shutdown. SR 3.8.4.8 A battery performance discharge test is a constant current capacity test to detect any change in the capacity determined by the acceptance test. Initial conditions consistent with IEEE-450 need to be met prior to the performing of a battery performance discharge test. The test results reflect the overall effects of usage and age. A battery modified performance discharge test is described in the Bases for SR 3.8.4.7. Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.8; however, only the modified performance discharge test may be used to satisfy SR 3.8.4.8, while satisfying the requirements of SR 3.8.4.7 at the same time. The acceptance criteria for this Surveillance is consistent with IEEE-450 (Ref. 8) and IEEE-485 (Ref. 12). These
" references recommend t hat the battery be replaced if its capacity is below 80% of the manufacturer's rating.
Although there may be ample capacity, the battery rate of deterioration is rapidly increasing. The Frequency for this test is normally 60 months. If the battery shows degradation, or if the battery has reached 85% of its expected application service life and capacity is s 100% of the manufacturers ratin the Surveillance Frequency is reduced to 12 months.g, However, if the battery shows no degradation but has reached 85% of its expected application service life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity 2: 100% of the manufacturer's rating. Degradation is i indicated, according to IEEE-450 (Ref. 8), when the battery capacity drops by more than 10% of rated capacity from its capacity on the previous performance test or is more than 10% below the manufacturer's rating. All these Frequencies are consistent with the recommendations in IEEE-450 (Ref. 8). This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required DC electrical power subsystem from service perturb the electrical distribution system, and chailenge safety l [ l O . g (continued) HATCH UNIT 1 B 3.8-Joh3 REVISION _ [$
DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.8 (continued) REQUIREMNETS systems. Credit may be taken for unplanned events that satisfy the Surveillance. The swing DG DC battery is exempted from this restriction, since it is required by both l l l l (continued) h HATCH UNIT 1 B 3.8-J1 6 5 A REVISIONgg
DC Sources - Operating B 3.8.4 BASES h SURVEILLANCE SR 3.8.4.8 (continued) REQUIREMENTS units' LC0 3.8.4 and cannot be performed in the manner required by the Note without resulting in a dual unit shutdown. SR 3.8.4.9 With the exception of this Surveillance, all other Surveillances of this Specification (SR 3.8.4.1 through SR 3.8.4.8) are applied only to the Unit 1 DC sources. This Surveillance is provided to direct that the appropriate Surveillances for the required Unit 2 DC sources are governed by the Unit 2 Technical Specifications. Performance of the applicable Unit 2 Surveillances will satisfy both any Unit 2 requirements, as well as satisfying this Unit 1 Surveillance Requirement. The Frequency required by the applicable Unit 2 SR also governs performance of that SR for both Units. REFERENCES 1. 10 CFR 50, Appendix A, GDC 17.
- 2. Regulatory Guide 1.6.
- 3. IEEE Standard 308 - 1971.
- 4. FSAR, Section 8.5.
- 5. FSAR, Chapters 5 and 6.
- 6. FSAR, Chapter 14. !
- 7. Regulatory Guide 1.93, December 1974.
- 8. IEEE Standard 450 - 1987.
- 9. Technical Requirements Manual .
- 10. Regulatory Guide 1.32, February 1977.
- 11. Regulatory Guide 1.129, December 1974.
(continued) h HATCH UNIT 1 B 3.8-66 REVISION A
___2,a, a a .,A a.2-43 a _.- ---- -- -a O UNIT 1 MARKUP OF CURRENT TECHNICAL SPECIFICATIONS AND DISCUSSION OF CHANGES N 1 i a O I e f I i O _ + - - _ _ _
O O . 3 7.I.1-)
~4 Table 4rki z- Reactor Protection System (RPSl lnstrumentation Functional Test, Functional
, Test Minimum Frequency, and Calibration Minimum Frequency
" y23 '). 3. l.l.'l, 3. 3. I Ik 5fd. 3.l I 80 E ID .bt l 3. 7.I M 3.3.1 I.% J 3 3././. D. 3.3.1.1.1? <
Z tram ,g Source of Scram Trip Signal
) Instrument Check Mnimum Frequency Instrument Functional Test Mnimum Frequency Instrument Calibration Minimum Frequency Nurt$er- i Group 1 (n)h (b) (E;)JN I!
- 1 - -
.7 X)O Mode Switch in SHUTDOWN A i NA Once/ Operating Cycle -17 Not Applicable ~
2 11 Manuel Scram A i NA >,. Not Applicable f**tt&t b Once/ week A> uf- 2.k 3 9 3. 3.1 []) Jr 1,9 1RM High High Flux C (&Si "' Once/ Week -4 Once/ Operating Cycle -)[ t
- SR 3. Al l.L 4.Srt7 3.it.7 a ., page %
S Inoperative NA Once NA 3 R. 3.3.l. l.54 43 Reactor Vessel Steam Dome (I S*I Every 3 months -1 Once/ Operating Cycle - O Pressure - High r f 5b Drywell Pressure High 0 S -I Every 3 months -i Once/ Operating Cycle 6r' 9 Reactor Vessel Water Level - 0 S-I Every 3 months ') - 9 Once/ Operating Cycle - 13 Low (Level 3) ' 7 Scram Discharge Volume High High i d b 4. 4. Float Switches Thermal Level Sensors B A NA NA (Every?ln6EThs) Every 3 months - 9 hOnce/-GOperating Cycle -%.1 L ,7~
@ g ,
,8 2 C- APRM Fixed High-High Flux -B l S-f Every 3 months fs kce/ Week (p)S A-10 4, \
$ N 5(23.3.1l.lo >
C. Inoperable B NA Every 3 mont ~.J NA %y3.1 l.Z. 2: o d Downscale B NA Once/ Week (e) ' , NA gg),3,,,g t \p [3. ( Flow Reference Simulated B S -I Every 3 months v -1 Qnce/ Week ( SA - /c .p , Thermal Power Monit ' ' Sa m
$* 6 15% Flux SI Once in *) . nce/We gugrq) I 1-f ' l f B NA M Every(C Effective gg g 6' Full Powe ours .W g 8 Q.i L _
3 31 ) B .w
; g.u y 3 a.i.is sea.ugj zy P-u # d.9
~-
V - v 2 3.s.m
% Notes for Table 4.3-1 (Cont'd) -4 m
- k. The electrohydraulic control oil pressure sensors shall be set to trip atg600 psig control oil pressure. Y M fe 1 ~b dd N INIWoble Y'al"e (l. 'Pettorm withinh(hours of starkif not performed wb the previous 7 dag
- m. When changing from the Run Mode to the Start and Hot Standby Mode, perform the required surveil lance within , 4 h b N ).3,f.f,QwJ 'g 3.1/,/,y 12 hours ef ter entering the Start and Hot Standby Mode gs peMarmed witNq the previout I days ( g
%t 3 . 3. 5 l . i.
- n. ThekPRM IRMSnd SRM channels shall be compared for overlap during each startup, if not performed within g 3_ 3,g,g, q the previous 7 days. y,
- p. This calibration shall consist of the adjustment of the APRM channel to conform to the power values
- 3' y, [ 7 b
calculated Adjust the APRM by achannelif heat balance during the absolute the Run>2% difference Mode when thermal power 125% of rated thermal power ( q. {fP x 3 M A fu 3/ 3. 3. IJ. ) -h This calibration channel to conformshall consist of to a calibrated the flow adjustment of the APRM flow referenced simulatedT thermal signal. power f ' Ns 3, 2 g) Y - .. . -.
, (dfo$ed 6 k 3 3.l.).li d $ d 3 J.J.l g m
3= 9 to o
- 2.
B to o c+
.o g@
b D K:\wp\techsp\h\3&4U1 TAB b f
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% W >\ _
1 _ _ _ ._ _ _ _ . _ . . _ _ l i DISCUSSION OF CHANGES ITS: SECTION 3.3.1.1 - RPS INSTRUMENTATION TECHNICAL CHANGE - LESS RESTRICTIVE (continued) L.8 The CHANNEL FUNCTIONAL TEST (CFT) requirement for the float type switches has been extended from quarterly to once per 18 months. This new Frequency will reduce radiation exposure to plant personnel performing this Surveillance. The NRC issued this change as Amendment 193 to the Unit 1 TS by letter dated March 15, 1994. Analysis has also been performed (GENE-770-25-1092) that shows a negligible impact on safety with the Surveillance being performed every 18 months instead of the current 3 months. Since the CFT is part of a CHANNEL CALIBRATION (per the definition), and a CHANNEL CALIBRATION requirement is specified every 18 months (proposed SR 3.3.1.1.13), an actual CFT SR is not provided. O F 1 O HATCH UNIT 1 9 REVISION
$b) 3).2.)
llMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.3.F. Doeration with a Limitino Control 4.3.F. Doeration with a limitino Control Rod Pattern (for Rod Withdrawal Rod Pattern (for Rod Withdrawal Error. RWE) Error. RWE)
--- - , ,, g A Limiting Rod Pattern for RWE Duri operdtion naL%iting ,
u exists when the MCPR is less to rol Roc Patte for 19fE exi s V N than the value provided in the a only 'ne RBM hannel/is ' g M M Core Operating Limits Report. operabi an in rument/funct nal {.4 3 festo he RBM shall be perf reed ,' During operation with a Limiting P prior o withdfawal of the ntrol ! Control Rod Pattern for RWE and 4% rod 4. A L ern f / RWE (isdefine%iting)todPa d by Specific tion /
\whencorethermalpoweris1 either:
l- [ , Lg 3, J, .2.g 1. Both rod block monitor (RBM) g ,p,g4 channels shall be operable, or I If only one RBM channel is f#M 2. hcP b OPERABLE, drawal shall control rod with-be/Dioctea within b vh6 es nours. or - -
- 3. If neither RBM channel is OPER-
@ ABLE, control rod withdrawal (b shall be blocked G. Rod Worth Minimiter (RWM) l G. Rod Worth Minimizer (RWM) l
- 1. Doerability 1. Ooerability l l Whenever the reactor is in the a. RWM shall be demon-
') . 3 k ' \ Start & Hot Standby
- cr Run Mode strated OPERABLE in the d below 10% rated thermal power, Start and Hot Standby MO 3's#,t.i 4
the RWM shall be OPERABLE. Mode prior to withdrawal pu> L -)
'~~
of control rods for the $23J.2.12
- a. With the RWM inoperable purpose of making the before the first 12 control reactor critical and in g 3 g ,7,g,3 rods are withdrawn on a the Run Mode when the RWM startup, one startup per is initiated during control hc5154 L calendar year may be per- rod insertion when reducing formed provided that "E"'"" J control rod movement and pfope:,c) <
compliance with the pre-g Verifyingprope 1(} scribed BPWS control rod (.S(23 3.2.d nnunciation of t pattern are verified by a \ s ection error of A -( second licensed operator lea one control . or qualified member of the m4 which iolates the pre- t h plant technical staff. scribe ithdrawal h kCd b. With the RWM inoperable sequence oaded into (Lag o ths RWM, a after the first 12 control gpod " rods have been fully with-drawn on a startup, opera-l2) Verifying the od block function of the WM by tion may continue provided attempting to mo a that control rod movement control rod that and compliance with the olates the prescri d wi drawal sequence load ,into the RWM. N L g sa 3 3.212. Entry into the Start and Hot Standby Mode and withdrawal of selected control rods is permitted
*q for the purpose of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.
HATCH - UNIT 1 3.3-5 Amendment No. N, 38, 42, 52, ME, H2, MB, 180 8d9
1 DISCUSSION OF CHANGES ITS: SECTION 3.3.2.1.- CONTROL R0D BLOCK INSTRUMENTATION l TECHNICAL CHANGE - LESS RESTRICTIVE ,
" Specific" !
(continued) L.4 The proposed change will delete the requirement to test the other RBM channel when one channel is inoperable. Generically, the requirement for demonstrating operability of the redundant subsystems was originally chosen because there was a lack of plant operating history ard a lack of sufficient equipment failure data. Since that time, plant operating experience has demonstrated that testing of the redundant subsyst. ems when . one subsystem is inoperable is not necessary to provide adequate assurance l of system operability. This change will allow credit to be taken for normal periodic l surveillances as a demonstration of operability and availability of the ' remaining RBM channel. The periodic frequencies specified to demonstrate operability of the remaining channel has been shown to be adequate to ensure equipment operability. As stated in NRC Generic Letter 87-09, "It is overly conservative to assume that systems or components are inoperable when a surveillance requirement has not been performed. The opposite is in fact . the case; the vast majority of surveillances demonstrate the
/7 systems or components in fact are operable." Therefore, reliance on the V specified surveillance intervals does not result in a reduced level of confidence concerning the equipment availability. Also, the current 1
Standard Technical Specifications (STS) and, more specifically, all the Technical Specifications approved for recently licensed BWR's accept the philosophy of system operability based on satisfactory performance of monthly, quarterly, refueling interval, post maintenance or other specified performance tests without requiring additional testing when another system is inoperable (except for diesel generator testing, which is not being changed). O ; HATCH UNIT 1 MA REVISION /"
(~'
\ (~ p/
x w 2o g.3.3.3 3
- 33 7 J'l y f(cfoxd ggg [g g. g g 5*AW Table &24t-N:- i 8 INSTRUMENTATION WHICH PROVIDES SURVEILLANCE INFORMATION Required
- z Ref. Operable
[ No. Instrument Instrument ype and 1 (b) Channets Ranoe Action Remerk s Au hen _ ' Fuae E
/
b 'b Reactor Vessel Water Level (1] O cdd I ecorder -150" to + 60* (c) { (d) ecator -150" to + 60" 2 (c) - (d) 2.4 Shroud Water Level 1 R order -317" to -17* (c) (d) L Ind stor -317" to -17" (c) (d) { 8' Reactor Pressure 1 Reco er O to 1500 psig (c) il 2 indice r O to 1500 psig (c) () 44 Drywell Pressure 2 Recorde -10 to + 90 psig (c) (c i 3
}U Y Drywell Tempereture Recorder to 500oF (c) (d Suppression er Air Temperature Recorder O t 500eF (d I,I Ofd 7 q -V Suppression Chamber Water Temperature 2( g Recorder O to 50cF (c) (d){.
N Suppression Chamber Water Level Indicator O to ! 3 4l b 2 2 ( Recorder O to 3
- O* (c)
(c)(e) (d) (d) , hN _ Socorh Chember PressureN Recorder -10 to 90 psig (d) N Rod Posihenjeformation SystemM Q 28 Volt Indicating 'ghts (d) h 1 Hydrogen and Oxygen Analyzer Recorder O to 5% (d) c:I Post LOCA Radiation Monitoring System 1 Recorder (c) ' (d)
]p Indicator 1 to 10C Re (c) (d) na g @
f e. /h_bo., U
*g ,(
.No a de d fu
*= 13 a) Saf / Relief Valve Position Primary RV Indicating Light at 85 ps a (f) I W' It%8cc.L b b Indicat 3.7. A 4.c.,_ g b) Safety /Reh Valve Position Secondary 1 Recorder O to 600cF (f) I le k s,- 443 *
[ fndicator Q_ y JT3 %gp. Q ~.$. ' 3; q 1 R.i
jN /~} - N, 3.3.3.I-) y TABLE-&? T1 (Continued) W
$ INSTRUMENTATION WH!CH PROVIDES SURVEILLANCE INFORMATION Required 2 Operable G Ref. Operable No. Instrument Instrument ~
~
R (b) Channels vpe and Renae A ction I 44 Drywell High Range Pressure 2 Re rder o to 250 ps;g (c) di g 46 Drywell High Range Radiation '2 4'b Indice r 1 to 107R/Hr (g) 2 Record 1 to 107R/Hr (g) Main Stack Post- Accident Effluent Monit . 1 Recorder 10~3 to I, (g) (h) 1x105pCi/ 17 ctor Building Vent Plenum 1 Recorder 5x1 '3 to (g) (h) Post- cident Effluent Monitor 1x105 Ci/cc { R/ P v
~
pra p s k % 2.c,L.A, % Lw!h G!tO2,; c and uc+es a-b f er k _._ a n C1 a L N W
.h 2 e w s
w O 03
I
,,- 7- % r V) U) f x_/
8
^,
# m
).].lZ#}
h
~ P(ppied M ) k M ,
) Table 4r2-fr
$ kQ u g'(t v%t S Check. Functional Test, and Calibration Minimum Frequency for Radiation
, Monitoring Systems Which Limit Radioactivity Release S(t 3 34 2 1 $g3 34.2. 3 c
- z- Ref. Instrument Check Instrument Functional Test Instrument Cahbration
[ No. Minimum Frequency Minimum Frequency Minimum Frequency 1 Instrument (b) (c) (d) 1 Off s Post Treatment Once/ day Once/ month (f) Every 3 months Radiation Monitors __ _ l
-f h Refueling Floor Exhaust Vent Radiation Monitors Once/ day I e/ quarter (D Every 3 months l
A.7
#3 Reactor Building Exhaust Once/ day L0nce/ quarter Every 3 months Vent Radiation Monitors l h5 Control Room Intake Radiation Monitors Once/dey Once/ quarter (f) Every 3 m l
5 Main Steam Line None once/ week (f) Every 3 months (g) - adiation Monitors f YC* 3.3 L.L-l Notes for Table 4.t8 l (Ct b[k I"
- d
> I
- a. mn entitled "Ref. No." is only for nvenience so that a one-to-one relationship can be (
established een items in Table 4.2 8 and item Table 3.2-8. k '
%d e 3 3. ~1.1 I TI: L 3.g j P instrument checks are not r uiredTv7ie~HThese instruments are not quired to be operable or are AM b trippe owever, if instrument ch . s are missed, they shall be performe rior to returning the '#
- I instrument to a& operable staine -
A l "b 2 a b
/
ro
= +
Q. t
! W
= ~
2: N o tc hw l 8$
/~%
] N 3 3.l 1-1
--4 --
Notes for Table 4.44 (Cont'd) Instrument functional tests are not required when t 'nstruments are not required to be operable or er ripped. However, if instrument functional tests are rru ed, they shall be performed prior
- 2: G(1.3 0- l to returns the instrument to an operable status.
- d. Instrument calib 'ons are not required when the instruments are not r ired to be operable or are tnpped. However, if ins. ment cahbrations are missed, they shall be perform prior to return-g the instrument to en operab status.
- e. Deleted.
l rA b [f. This instrumentation is exempted from the instrument functional test definition. This instrumenD )<- t Dig g g functional test will consist of injecting a simulated electrical signalinto the measurement
- W h TTgrj,3 4y channels. '7' W CTS: QQ I 2 Standard current source used which provides an instrument channel alignment. Calibration using a ~ ws a b , Qep , O radiation source shall be made once per operating cycle. i* M 3 30-5(D.3-4 L C -
Logic system functional tests [mcLsimulated i!rtrternatic ectuatior(shall be performed once each operating
- UN cycle for the following:
B , m 1. Secondary Containment Actuation ?>
- o. E 3
m @ e U \ v3 a 5
% .--. co 0 U3
em <m /n l % Tab)e 2-8 l @
--e i
I i b heck Functio i Test, end Calibration Mini Monitcri Jystems Which t.imit Radion 'vityEnlaa=a Frequency for Radiation i Instrument Functional Test Instrument Ca!>bration E Ref. Instrument Check Minimum Frequency Minimum Frequency Mirdmum Frequency [ N o. Instrument (b) (c) (di 1 Once/ month (f) Every 3 monthb f1 Off-gas Post Treatment Once/ day l f Radiation Monitors Once/ day Once/ quarter (f) Every 3 months 2 Refueling Floor Exhaust l Vent Radiation Monitors once/ day Once/ quarter (f) Every 3 months 3 Reactor Building Exhaust Vent Radiation Monitors SgJ. J zf,3 S(dL 3 7.).1 R 3. M . L 2. - 4 Control Room Intake Once/ day (6nceWig,Oa ' Every 1 monthe Radiation Monitors %M fl __ None Once/ week (f) Every 3 months (g) 5 Main Steam Line s Radiation Mog ru Notes for Table 4.2-8) ' e
#a' e column entitled 'Ref. No. 's only for convenience so that a one-t one relationship can be kk4res of entablish _hniyveen items in Table -8 and items in Table 3.2-8. D: 3Ul
- b. nstrument checks are not require when these instruments are not require to be operable or a _
Q,Q 3.O.1 % S te. w/ .Q i tripp ~ However. if instr sment checks at 'ssed, they shall be performed prior retuming the y Im Mc. b*' J.fg) . instrurrwnt to g operable status. a e r= 2. to P M $ o z -d W bd .O O ]
't?
vi
sm sm /~ y
-4 Ne f:: 'd': 4.2 S (C..a'di h '
Mtrument functional tests are not reque. d when the instruments are not required to be operable or i e tripped. However,if instrument functions sts are missed, they shall be performed prior C ScQ 3.c.1 z -
~
to reti 'ng the instrument to an operable status. w
- d. Instrument c ibrations are not required when the instrums are not required to be operable or are tnpped. However, ' instrument calibrations are missed, they she, e performed prior to return-ing the instrument to en o rable status.
- e. Deleted .
l w I f. s instru tation is exempted from th ument functional test definition. This instrument Q ,3 4 w { functio test will consist of injecting a simulated el rical signalinto the measuremen ( chan i
- g. Standard current source used which provides an instrument channel alignment. Calibration using a det D[kaWow d
.. k'S 4( C% 8 reding, enorca nhall be made once per operating cycle. W g; g m ) . 3 7.t. 4 acy f d g3
- p:::::~a Log *c system functional tests g_n_&<uquiateo autwaatic actuatiop shall be performed e _ A.T I
for the following: , E3
@ 1. Secondary Contait ment Actuation :fy
- o. P an e-
$ he SC s5 5,a./ oF d Ie $
c+ z W ' M .L.a '
?
c; - I4 %h Se,(h. [ b- c c co U1
(w.
%s Y h --A Notes for Table 4.2-8 (Cont'd) n Z
s
- 2. Standby Gas Treatment System Actustion Z
w
-4
- 3. Steam Jnet Air Ejector Off-gas Actuation Primary Containment Purge and Vent Valve Closure j 4'4. -
- het g f MWi d of l(%CS
& IT S O. 3 6 1 mD. A 6. L.
5. SR b3 7I 4 MCRECS Control Room Pressurization Mode Actuation
% A c Ts : 3/9 4 g/,s g
1
) 0 a g
/
- 6. (Deleted)
." r 7 . Mechanical Vacuum Pump Isolation i:
e logic system functional tes allinclude a calibration of time lay relays and timers necessary 1 for pro 4 functioning of the trip syste (_ 2 g 2 e
- s e
n O. P E3 & n 2 o w W T i
t DISCUSSION OF CsANCES O ITS: SECTION 3.3.8.2 - RPS ELECTRIC POWER MONITORING ' TECHNICAL CHANGE - LESS RESTRICTlyl (continued) L.2 The allowed out of service time for one inoperable assembly is extended to 72 hours and for two inoperable assemblies is extended to one hour to provide sufficient time for the plant personnel to take corrective actions. With one assembly inoperable, the other assembly is fully capable of providing the necessary protection, thus, 72 hours provides time to repair the inoperable assembly and decreases the potential for a unit upset (that could result when power supplies are shifted, since power is initially lost to the RPS trip system and either RPS bus powered components). The time extension for two inoperable assemblies is minimal but necessary to allow consideration of plant conditions, available personnel and the appropriate actions. L.3 A Note has been added to this Surveillance such that the Surveillance is only required to be performed when the unit is in MODE 4 2 24 hours. Thus, the 184 day Frequency would not have to be met until a shutdown to MODE 4 2 24 hours occurs. The performance of this Surveillance could result in hal f-scrams, actual valve isolations, and other plant perturbations, since if the assembly opens, power is lost. The test requirement has been changed to allow it to be performed while shutdown to minimize the impact of this Surveillance on plant operation. This is O consistent with many of the of the more recently licensed BWRs and the BWR Standard Technical Specifications, NUREG 1433. L.4 The time delay setting for the undervoltage trip has been extended from zero to s 4 seconds. In addition, a time delay setting has been provided for the overvoltage and underfrequency trips. The NRC issued this change as Amendment 191 to the Unit 1 TS by letter dated November 24, 1993. O sms om 1 2 RmSION 4
^ r (R
( w (O J I f.1 g
-4 W Table 4.2-8 2- Check, Functional Test, and Cah tion Mmimum Frequency for Radiation
, Monitoring Systems Whic imit Radioactivity Release C
z Raf. Instrument Check Instrument Functional Test Instrun nt Calibration No. Mmimum Frequency Minimum Frequency Mmimun requency laL Instru t (b) (c) (d 1 Off-gas Post Treat nt Once/ day Onces 3onth (f) Every 3 months / Radiation Monitors # l ( -- 77 Refueling Floor Exhaust once/ day Once/ quarter (f) Every 3 months Vent Radiation Monitors l 3 Reactor Building Exhaust Once/ day Once/ quarter (f) Every 3 months Vent Radiation Monitors l 4 Control Room Intake Once/ day Once/ quarter (f) Every 3 month L u nadiation Monitare ~ l 5 eam e None week (f very 3 m (g) W \ \ M A i N bM ' %s5fo. cf N Notes for Table 4.2-8 ' 7
- a. The column entitled "Ref. No." is only for convenie e so that a one-to-one relationship can be GA.d'-
([(C gz jp;l-y a. - l est fished between items in Table 4.2-8 and items in Tab 3.2-8. saq 3O a' ,W - r 1 TS* 3. 0 4, , g Q
- b. Instrume checks are not required when these instruments are no quired to be operable or are jq
[ b-} tripped. Howe if instrument checks are missed, they shall be perform rior to retuming the g instrument to en oper le status, 2 c .
-b f
> ?
2 t m x<
- s
- r+ N Z
O f v3
- L ***
co
y . Table 3.2-10
-4.
O INSTRUMENTATION WHICH MO ORS LEAKAGE INTO THE DRYWELL i E R 3 Required Operable o Q (a) Instrument (c) Channels per System Settina_ Remarks 1 Q .g eil Equipment Drai ump 1 d) Tec Spec The imiting Conditions for Flow In. retor 't R operati of the Leakage g, Detection stem are provided Drywell Floor Drain Sump 1(b)(d) - Tech Spec - in Section 3. { gg Flow Integrator 3.6.G.1. 3 Scinti!!ation Detector for i m 1 [d) (e
$2
! Monitoring Air Particulates 4 L b 2-Scintillation Detector for [,.- I lid) (e) 3d f Monitoring Radioiodine , l 5 GM Tubes for Monitoring Noble Gases Jd) (e) t W e
- a. Th lumn entitled "Ref. No." is only for e nience so that a one-to-one relatidsyship can be established
- Fu items i ahl *t 9 t ri an_d items in Table 4.2-1 -
, b.1. Whenever the systems are required to be operable, there shall be one operable system. If this cannot htT/44 8 be met, the indicated action shall be taken, g l b.2. One instrument channel may be inoperable for up to G hours to perform required surveillances prior to entering h/cy M kffj other applicable actions. (1 C 3
- c. (holso finbraararnre_ onebebipment d_ rain sumo and th7other for the floor drain sump, comprise one basic
{}O instrument systemgwo s sum-ioaice scinuiiauvo unou i., urm ior niunavnng air' rticuistes ano one for inv.a 7,Q .(. n. onng soioome, comprise t basic instrument systems. eta sensitive GM detect for monitoring noble
> ,/ gases co 'ses a fourth basic ins ment system. An alternate tem to determine the 1 age flow is a manual 2 ;
Q system wher the time between su ump starts is monitored. 's time interval will date ;ne the leakage floy C.
- s becan=afH t i . ^* % - - i=knnwn. _7 M (NedministrativMrmation: performs nhqntret function P r+ _
g, 2 High setpoint ala ill be set three times above back nd radiation. Failure alarm will be et below background A. 2 O r tion. Specific valu ill be established during syste artup. p '
-3 '
4, M f-w e- s w O U1
e f
- l. LA-L
/ ,
y - Table 4.2-10
--I
$ Check, Functional Test, and Calibratio inimum Frequency for instrumentation
, Which Monitors Leaka into the Drywe!! l z Instrument Check in ment Func*ional Test ins ment Calibration
,Z Ref. N . Minimum Frequency ' mum Frequency Mirw m Frequency Ja) _
fnstrument _ (b) c) d) 1 r 11 Equipment Dh O O nth Ever months u,j} f SumpF Integrator 2 Drywell Floor Drain Sump (%e/ Once/ month $R 3. gg,1 "Every months l Flow Integrator ! [ OnceMey ' '" 3 '~ Scintillation Detector for Morytoring Air Partic-L. IM% Once/ month Q "J.Q. 5,1 3 d.53 ' everv *i moath-lI r ulates ' - d l 4 Scintillation Detector Oneenley Once/ month SR 3 S of2 ' Every & months R AW6 T l for monitoring Radioiodine M 3-(5.\ t 5 GM Tubes for A j Once/ month Sg 3,Qg Every a months) ,l ro i
^
g Notes for Table -10 column entitled *Re . " is only for convenience so that a one- s' 2
~
one relationship can be ablished between . ite in Table 4_L10 and ita in Tahta 3.2-10. l
- b. Instrument checka are not required when these instruments are not required to be operable ri i However, g 3, g g if instrument checks are missed, they shall be performed prior to returning the instrument to en operable status,
- c. Instrument functional tests are not required when the instruments are not required to be operabl [
However, if instrument functional tests are missed, they shall be performed prior to retuming the instrument to i an operable status. Q 3. Q. } l B i m M n "l3 e t n 'N z F > o :P " e . -
%,1 V.g - -
N we V iM ! t h _ _ _ _ _ _ . _ _ _ . _ _ . _ _ . . _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _m.. - - - --- *,: - e = , - . . - +---i _ _ _ _ _ _ _ _ _ _ _ _ _
T DISCUSS:0N OF CHANGES l {q} ITS: SECTION 3.6.1.1 - PRIMARY CONTAINMENT TECHNICAL CHANGES - LESS RESTRICTIVE LC.1 (continued) addressed by plant operational procedures and policies. Therefore, the continuous leak rate monitor, and associated actions are removed from the Technical Specification.
" Specific" L.1 The value for P, is being lowered to 53.6 psig. P, is defined in the l Technical Specifications as the peak containment internal pressure that is used for 10 CFR 50 Appendix J (leakage testing) purposes. The peak containment internal pressure, as related to 10 CFR 50 Appendix J, has traditionally been the calculated maximum pressure following a large break, design basis Loss of Coolant Accident (LOCA). For Hatch Unit 1, this break also results in the highest Final Safety Analysis Report (FSAR) analyzed accident pressures. The current Mark I Containment Long-Term Program analyses regarding the containment temperature and pressure responses following a LOCA are documented in Unit 1 FSAR Section 14.4.3.3.2. In addition, a more recent analysis, which increased the containment normal operating pressure limit from 0.75 psig to 1.75 psig is
(] y documented in GE-NE-A00-05873-02, dated April 1994. The Hatch Unit 1 containment pressure response, due to a postulated design basis LOCA, was re-evaluated as part of the Mark I Containment Long-Term Program and is documented in NED0-24570. The purpose of the Mark I Containment Long-Term Program was to " perform a complete reassessment of the suppression chamber (torus) design..." according to Appendix A of NUREG-0661. As a part of this complete reassessment, the Mark I Containment Long-Term Program included plant unique analyses of the containment LOCA pressure response using the Homogeneous Equilibrium Model (HEM) for vessel blowdown described in NED0-21052 and the containment response model described in NED0-10320. These plant-unique analyses and results were provided to the NRC in Georgia Power Company's letter dated January 26,1983 (with later supplements) and approved by the NRC in a Safety Evaluation Report dated January 25, 1984. These approved analyses resulted in significantly lower containment peak pressures than submitted in the original FSAR. Subsequent to NRC approval, the Hatch Unit 1 FSAR was updated to reflect the new analyses and their results. Since the Georgia Power Company Mark I Containment Long-Term Program submittal, revisions have been made to certain parameters used in the model to account for the Extended Operating Domain Analyses with reduced feedwater temperature. This revision has resulted in slightly higher peak containment LOCA analyses pressures from those presented in the 1983 submittal. Through the 10 CFR 50.59 safety evaluation process, the FSAR ( was updated to reflect these results. The current LOCA analyses, provided HATCH UNIT 1 3 REVISION,Ah
q y DISCUSSION OF CHANGES ITS: SECTION 3.6.1.1 - PRIMARY CONTAINMENT TECHNICAL CHANGES - LESS RESTRICTIVE L.1 (continued) in the FSAR section referenced above, result in peak containment internal pressures of 51.6 psig for Unit 1. As indicated in NED0-24570, the peak containment pressure calculations for a design basis LOCA assumed an initial pressure of 0.75 psig. Also, the peak containment LOCA pressure is higher than the analyzed peak containment pressure for a Main Steam Line Break or small break LOCA inside containment. As indicated in GE-NE-A00-05873-02, the containment initial pressure was evaluated to be increased to 1.75 psig. The evaluation addressed the following issues:
- Short-term DBA-LOCA containment pressure and temperature
- Long-term DBA-LOCA containment pressure and temperature
= wCA containment hydrodynamic loads
- Safety / relief valve loads
- Appendix J containment leakage requirements Other issues not related to this P, change.
Based on the result of these evaluations, it was determined by the Mark I Containment Long-Term Program should be increased by 2 psig to 53.6 psig. Therefore, peak containment internal pressure value of 53.6 psig for l Unit 1 forms an acceptable basis for structural integrity as identified in the proposed Bases of the Technical Specifications. This pressure is significantly less than the containment design pressure of 56 psig and the ASME Code allowable of 62 psig. L.2 A 1 hour time is proposed to allow restoration of the primary containnient prior to requiring a unit shutdown. The new out of service time is based on engineering judgement of the relative risks associated with: 1) the probability of an event during the 1 hour requiring the primary containment; and 2) the plant transient and potential challenge of safety systems experienced by requiring a unit shutdown. The new time is consistent with the current Unit 2 Technical Specifications and with the BWR Standard Technical Specifications, NUREG 1433. I
\
CJ l l HATCH UNIT 1 4 REVISION j
hC[k"$c4 3.(.f. 3 _ , i LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS l 3.7.A.7. Primary ContajnmJtn1 4.7.A.7. Primary Containment Purae System Purae System
- a. LWhen primary containment is a. In addition to the requirements required, all drywell and of Specification 4.7.D, each A.' .
suppression chamber 18 inch purge supply and exhaust isolation drywell and suppression chamber 18 inch pu e supply and M M l* y* ) Lyalvee chall be enerable in l- exhaust iso atton valve shall . the fully closed position except w be verified to be closed at I g ].(,g.).) when required for inerting, de- propned least monthly. ; inerting, or pressure contro gg mg b. Each refueling outage each
$h4 e Each dr 1 and suppre ion drywell and suppression chamber M 6.l 3./D hamber 18 h purge su y an g 18 inch purge supply and >
g e aust isola n valve sh hav p g,p exhaust isolation valve with a a1 ge rate a specified resilient material seat shall 4.7.A. . be demonstrated operable by having its valve seat replaced 3.b.13 c. [The drywell and suppression ri- gsu ieda ,~7 chamber 18 inch excess flow a s wi n its li -- . 3 .6 isoiation dampers shaii be . __ : operable at all times when th p") c. At least once per3 ynts the Lk 3 - U dampers will be Nisuail bnitIprimarycontai _teority is recuired n the 18 Qnspected and,tycle gdoverifh N3 k l'E N h isolatioTvat1TH to the tRe o.mpn a nave no damage ; in M dry 1 or suppression c ber are which renders them incapable of ' onen. - perfoming their designj i functinn - If these requirements cannot be met, close the drywell and ; suppression chamber 18 inch purge ' Acpid supply and exhaust isolation O g valve (s) or otherwise isolate the penetration (s) within 4 hours or fulfill the requirements of i Specification 3.7.A.8. ,
- 8. Shutdown Recuirements If Specification 3.7.A. cannot be met, an orderly shutdown shall be initiated and M0 d the reactor shall be brought to Hot i g, Shutdown within 12 hours and shall be in >
the Cold Shutdown condition within the t following 24 hours. i e f i e 6
. . g g g g chl3
i l l DISCUSSION OF CHANGES l ITS: SECTION 3.6.1.4 - DRYWELL PRESSURE l TECHNICAL CHANGE - MORE RESTRICTIVE M.1 A Specification requiring drywell pressure to be s 1.75 psig is being added. This is required since accident analyses, documented in the GE-NE-A00-05873-02, dated April 1994, assume this pressure at the start of an accident. Appropriate ACTIONS and Surveillance Requirements are also added. This is consistent with the BWR Standard Technical Specifications, NUREG 1433. O i HATCH UNIT 1 1 REVISION ,A
-- - ~ - -. - . _ - . . ~ . -- . . - - , - .. . . . - _ - 5 Sged L haa 3.2.1 tlMITING COPOITIONS FOR OPERATION O SURVEILLANCE RE00lRDefr5 3.9.A.2. Standby AC Power Sunoly (Diesel 4.9.A.2. Standby AC Power Supoly (Diesel Generators IA. IB. and IC) Generators IA. 18. and IC) * (Continued) (Continued) fof'5' M 6**.
- f. % e5 % st ).P.l.2 l Doerabilit a. Doerability L diesel rator itself/ g'3 ard,its auxili ies are 1 Iachdieselgeneratorshall ;
operable. be manually started and M'l-@ loaded to demonstrate @
- d M 2.
operational readiness b SR 3 9.t.2. in accortlance with the t frequency specified in Table t, (4.9-1 a v - area - i erify that each l I arts from ar@ient (condition /fyradually load w , _ .. tor to 1710-2000 g ,jo 16t** and operate for 2 60 inute stead -state of 4160 4- o ts ' and a steady-state frequency t, .l of 60 21.2 Hz will be
' .Laaintaineds erify the pres *g dieselairstar] t receivers to be 2 225 psi p ' QP*'*I See PiS&W **
(k%g f uW -
-pagef ,
3 g3' ;s do 2.lAtleastonceper184 days, f M each diesel generator shall bM' be started and verified to reach synchronous speed in b 512 seconds, loaded to an > indicated 2250-2400 kW** for b IA and It and 2360-2425 kW** /repa cI , for IB R 120 Necond 7 ccle 2
> d g ,).f f
operated for 1 60 minutes. The test will verify the ['3 diesel generator will achieve and maintain a stead -state voltage of 4160 20 volts and a steady-state requency r ( of 60 i 1.2 Hz.* g.to i MO 'For the IB (swing) diesel, a single test will satisfy the requirements for 54 PAI Unit 1 Specification 4.9.A.2.a.1 and Unit 2 Specification 4.8.1.1.2.a.4, with the diesel connected to one unit's emergency bus for one periodic test and connected to the emergency bus in the other unit during the next i periodic test. A single 6-month (IB4-day) test for the IB diesel will satisfy the 4 "c requirements for Unit 1 Specification 4.9.A.2.a.2 and Unit 2 A 6 W3 tM Specification 4.8.1.1.2.b. The 6-month test will be performed using .dh il .i the starting circuitry and emergency bus from one unit. The next 6-month test will be performed using the starting circuitry and OpM emergency bus from the other unit. f!s t' g ** Momentary variations outside this band shall not invalidate the test.
'ra 4Wb N g3948 HATCH - UNIT 1 3.9-2 Ameruhent No. M, M7,178 1 413
DISCUSSION OF CHANGES ITS: SECTION 3.8.1 - AC SOURCES - OPERATING ADMINISTRATIVE A.] Reformatting and renumbering requirements are in accordance with the BWR Standard Technical Specifications, NUREG 1433. As a result the Technical r Specifications should be more readily readable, and therefore, ; understandable by plant operators as well as other users. During this ' reformatting and renumbering process, no technical changes (either actual or interpretational) to the Technical Specifications were made unless they were identified and justified. In the specific case of the Auxiliary Electrical Systems section, the new section number is 3.8. These general paragraphs in this section have been deleted since they provide general guidance that is discussed in other parts of the Technical Specifications or Bases. A.2 Proposed Notes 1, 3, 5, 6, and 8 to SR 3.8.1.2, and Note 4 to SR 3.8.1.5 l , have been added. Note 1 to SR 3.8.1.2 allows SR 3.8.1.5 to satisfy SR ' 3.8.1.2, since it is more restrictive than SR 3.8.1.2. Note 3 to SR - 3.8.1.2 allows the engine to be warmed up and gradually started. These methods are currently employed, and have been specifically added for clarity. Note 5 to SR 3.8.1.2 allows gradual loading. Note 6 to SR 3.8.1.2 allows for voltage transients prior to establishing steady state operation. Note 8 to SR 3.8.1.2 and Note 4 to SR 3.8.1.5 only allow a SR to be performed on one DG at a time. All of these are currently being performed, and have been specifically added for clarity. All of these O changes are considered administrative. A.3 The existing limitation on 18-month Surveillances to perform them "during shutdown" is more specifically presented in the proposed Surveillances. Each proposed SR contains a specific Note limiting the performance in certain MODES. While these limitations vary from SR to SR, each is consistent with the BWR Standard Technical Specifications, NUREG 1433, presentation (or bracketed option allowed based on plant specific - justification) which defines the intent of "during shutdown" for each SR, and with the guidance of Generic Letter 91-04. Additionally, the Note clearly presents the practice taking credit for unplanned events, provided the necessary data is obtained. In addition, since the swing DG is common , to both units, SRs that allow one SR performance to satisfy both units' requirements are allowed to be performed while one unit is not shutdown, provided the SR is being performed from the other unit. Since this is i only a change in presentation, this change is considered administrative. l l l HATCH UNIT 1 1 REVISION [k l
1 h L Q c y r w __4x
- bhS 3' SLRVEILLANCE p10UIREMWT5 LIMITING COP 01TIONS FOR OPERATim 3.9.A.2. Standby AC Power Suoolv (Diesel 4.9.A.2. Stan&v AC Power Suoulv (Diesel Generators IA. 18. and IC) Generators IA. 18. and IC)
(Continued) (Continued) to 4 O V b. 1 c.o - Diesel Battery (125 Voll) Izh- it mm1 Wam/ M b. Diesel Battery (125 Volti Each 125-volt diesel battery g propwJ capable f 1 g y?t.4. ) shall be subjected to the same periodic surveillance as 3,,gg 7> J .8 4.6 3 fupplyingthe ired 1 . W
.5(L the plant batteries in Speci- 33, Mg L' 847 fication 4.9.A.3. 3,g.d,g
- c. 3attery Charner Chd /- Battery Chatner , f.N!s' f L,c. 3.g.q.s N operable Dh tery chn ner 11) dCh bl Indicators I be prVvided _
)
available/3ta battery Eharsjer -10 to monitor the atus of 7TTiave adequa capacity to 4~ OOM battery cha r supply. d res n its battery full > s instnnentati shall cha within 24 hours rom a inc indication o output-discha condition whi g cu t and output vo _.) f/WE
. carrying DC load SE3.B M
- d. Diesel Fuel 7 *
[d. Diesel Fuel I There shall be a minim a of 1. The quantity of diesel l 99,000 gallons of acceptable l fuel available in eachy diesel fuel in the diesel fuel storaoe tan % uel oay tant ihall bel orage_ tanks _andJ mmina or svu gau5ns l
-- - - ,4 e s wuiu.o n each diesel fuel day tank.
~
I concurrently with the operability test speci-fled for the diesel in LSpecification 4.9.A.2.a.1 bt e. Me os,w of ck.y ( _
- 2. At least once per 92 days Cor .Tw. 3.% i.3 44b.J by verifying that a sanple of diesel fuel from the fuel See bss roact <bs 3e3 storage tank, obtained in
/ -fiv- Its t 3.g.3, J., %3 accordance with ASIM-0270-65, seg , is within the acceptable
! limits specified in Table 1 of ASTM 0975-74 when checked for viscosity, water and sediment. __
- e. Fuel Oil Transfer Pinos e. Fuel Oil Transfer Ptses A fuel oil transfer puup shall be operable and capable 1. The operation of the diesel of transferring fuel oil fuel oil transfer ptsms to frum the storage system to transfer fuel from the the day tank. storage systen to the day tank shall be demonstrated concurrent with the oper-ability test specified for that diesel in Specification j 4.9.A.2.a.1.
/
- 2. The operation of the diesel fuel oil transfer ptams to <
transfer fuel from each ! associated fuel storage j ! tank to the day tank of f ; each diesel, via the 1 installed cross connection l lines, shall be demonstrated-at least once per 18 months i during shutdown. (V HATCH - UNIT 1 3.9-2c kerdnent No.178 low
i Q seer ) f M se n k 8 'bAb'O'Y LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS ; I 3.9.A.3. 125/250 volt DC Emergency Power 4.9.A.3. 125/250 Volt DC Emergency Power System (Plant Batteries 1 A and System (Plant Batteries 1A and g See b ssi. of O g Lco7 8.4.b Eth\25/Zb9 oli plant b teries a. Weekly Surveillance c% <.c w
'N, ,
TM.JnlNiff sha be operab and .) Every week the specific gravit h ! sh'al q v n ope le batter ind the voltaae of the nilot ceU g.b charger gd ntilat system 4 6 erall battery voltage shall available fbe e h. C 5e be measured and recorded. Each pg,w J 7 9A.) 125 volt battery shall have a , p minimum of volts at the bat-3 , g.t4. c tery terminals to be considered operable. I 33 P D {b. Monthiv Survei11ance Every month measurements shall b m '2 r b ' 7' made of voltage of each cell to 1 pro p% h .* O-, g 4,9 ' VSA% the nearest 0.1 volt and the spe-y24$ cific gravity of each cell. These 3 .g.4 5 measurements shall be recorded. ;
) 1.iquid level shall be checked v 9 Ah g q,7 visually. _
g.g40
- c. Refuelino Outaae Surveillance g During (ei?E.QTreduled ref uelinD
~ utage J he batteries shall be j
,3 g ,3 subjected to a rated load dis-
- charge test. jThe specific gravity $
he3XWu,wofckD,3 -Q,c Fromed 4nd v5Tiage of each cell shall be r IB 340,i.2 nis Sedoa.\ s Q 3.3.4.'l , tJete io determined after the discharge i [ ' and recorded. i O 4 Emeroency 4160 Volt Buses (1E. 1F. and 1G)
- 4. Emeroency 4160 volt Buses (1E.
1F. and 1G) i The emergency 4160 volt buses (1E, The emergency 4160 volt buses (1E, 1F, and 1G) shall be energized and 1F, and 1G) shall be monitored to I one able. the extent that they are shown to be ready and capable of trans-mitting the emergency load.
- 5. Emeroency 600 Volt Buses (1C 5. Emeroency 600 Volt Buses (1C and 1D) ;
and 1_D) 1 The emergency 600 volt buses (1C The emergency 600 volt buses (1C l and 1D) shall be energized and and 10) shall be monitored to the operable. extent that they are shown to be ready and capable of. transmitting , the emergency load.
- 6. Emeroent,y_250 Volt DC to 600 Volt 6. Emeroency 250 Volt DC to 600 Volt !
AC Inverters AC Inverters i The emergency 2.50 volt DC to 600 a. The emergency 250 volt DC/600 volt AC inverters shall be ener- volt AC inverters shall be moni-gized and operable. tored to the extent that they are shown to be ready and capable of 1 transmitting the emergency load. - O "" * - % D k% W Sch 3.r HATCH - UNIT 1 3.9-3 Amendment No. 27, 4B 2eif
DISCUSSION OF CHANGES
, ITS: SECTION 3.8.4 - DC SOURCES - OPERATING ADMINISTRATIVE A.1 The proposed Technical Specifications present the station service and DG battery hardware components (battery and charger) in the DC sources LCO (proposed LC0 3.8.4). The battery cell parameters and DC Distribution buses are in separate LCOs (proposed LCOs 3.8.6 and 3.8.7, respectively.)
A.2 The Frequency of "each scheduled refueling outage" has been modified to be "18 months," since 18 months is a normal refueling outage schedule. A.3 This general paragraph has been deleted since it provides general guidance that is discussed in other parts of the Technical Specifications or Bases. A.4 The format of the proposed Technical Specifications would allow multiple Conditions to be simultaneously entered. Two or more DC sources could be inoperable, ACTIONS being taken in accordance with the Specification, and LCO 3.0.3 entry conditions not met. To preserve the existing intent for a unit shutdown, ACTION E is proposed. TECHNICAL CHANGE - MORE RESTRICTIVE H.1 Surveillances have been added consistent with the BWR Standard Technical O Specifications. Proposed SRs 3.8. 4. 2, 3.8. 4. 3, 3.8.4.4, 3.8.4. 5, 3.8. 4. 6, ( and 3.8.4.8 have beer. added for both the station service and DG batteries. SR 3.8.4.2 ensures the connection resistance is within limits or that no corrosion at the bittery terminals is present every 92 days. SR 3.8.4.3 ensures the batter) cells show no visual indication of physical damage or abnormal deteriorat ion. SR 3.8.4.4 removes visible corrosion and coats the connections witt anti-corrosion material. Both are required every 18 months. SR 3.8.4.6 ensures the connection resistance is within limits every 18 months. SR 3.8.4.6 verifies battery charger capability every 18 months. SR 3.8.4.8 requires a battery performance discharge test or a modified performance discharge test every 60 months. Also, a discharge test will be required every 12 months when a battery shows degradation or has reached 85% of expected life with capacity < 100% of manufacturer's rating and every 24 months when a battery has reached 85% of expected life with capacity 2- 100% of manufacturer's rating. These new Surveillances are additional restrictions on plant operation. M.2 Certain equipment needed to meet Unit I accident analysis is powered from Unit 2 DC sources. Currently, the Unit 2 DG DC sources are required since Unit I definition of OPERABILITY requires the necessary electrical power to be OPERABLE. To make the Technical Specifications more user friendly, i the Unit 2 required sources have been added, similar to the already l required Unit I sources. Since Unit 2 sources are now described, the current LC0 and ACTIONS for Unit I sources has been modified to explicitly O V 1 HATCH UNIT 1 1 REVISION
l l l DISCUSSION OF CHANGES ITS: SECTION 3.8.4 - DC SOURCES-0PERATING IECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" LA.1 The details relating to system design and purpose have been relocated to the Bases. The design features and system operation are also described in the FSAR. Thus, the LC0 has been written to require the Division 1 and 2 station service DC subsystem and the DG battery subsystems, as described in comment A.1 above. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in Chapter 5 of the Technical Specifications. Changes to the FSAR will be controlled by the provisions of 10 CFR 50.59. " Specific" L.1 The time to reach MODE 4, Cold Shutdown has been edended from 24 hours to 36 hours to provide the necessary time to shut down and cool down the plant in a controlled and orderly manner that is within the capabilities of the unit, assuming the minimum required equipment is OPERABLE. This extra time reduces the potential for a unit upset that could challenge safety systems. This time is consistent with the BWR Standard Technical Specifications, NUREG 1433.
L.2 Proposed ACTIONS A and B have been added to provide clarity as to what actions to take when a DG DC source is inoperable. The Completion Time for ACTION B (12 hours) is consistent with the proposed time (assuming both a DG and an offsite circuit are inoperable when the DC DG source is inoperable) in proposed LC0 3.8.1. ACTION A allows 7 days to restore the swing DG DC source, if it is inoperable due to SR 3.8.4.7 or SR 3.8.4.8. These two SRs result in the inoperability of the swing DG DC battery, which is common to both units. Therefore, 7 days is provided to perform the SR and return the DG to OPERABLE status. Vithout this allowance, a dual unit shutdown would be required to perform the SR. L.3 A Note has been added to proposed SR 3.8.4.7 to allow the modified performance discharge test to be performed in lieu of the service test of ' SR 3.8.4.7. As stated in the BWR Standard Technical Specifications Bases, NUREG 1433 (proposed by NUREG change package NRC-15), this substitution is acceptable because SR 3.8.4.8 represents a more severe test of battery capacity than SR 3.8.4.7. O HATCH UNIT 1 3 REVISION h
spu h ton c.x , \ LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.12 A.2. Performance Requirements 4.12.A.2. Filter Testing O a. The results of the in-place DOP and halogenated hydro-
- a. The tests and analysis shall be performed at carbon tests at design flows least once per operating 57'O on HEPA filters and charcoal f,J 7 cycle, not to exceed 18 f,f7.f absorber banks shall show months, or af ter every 199-percent DOP removal and 720 hours of system opera-199-percent halogenated tion or following painting, hydrocarbon removal, respec- fire or chemical release in tively when tested in accordance any ventilation zone communi -
with(ANSI N510-1975.} cating with the system.
- b. The results of laboratory p,1 b. 00P testing shall be per-carbonsampjjt. analysis formed after each complete shall show >90-percen radio-l or partial replacement of f~. .f,7. C active meth(yi todice f,$. 7 the HEPA filter bank or removal when test e_d in after any structural accordance withjRDI-Mrb-li maintenance on the system 25'C, 95-percent R.H. ) . l housing,
- c. Fans shall be shown to c. Halogenated hydrocarbon operate within +10-percent l testing shall be performed 6,6,7, d design flow when tested after each complete or f.f,7 in accordance wit ANSI Q510-19M. ,
partial replacement of the charcoal adsorber bank of f af ter any structural b- maintenance on the system housing. I O [. $, ], t. A b Y yb $ob,'Oe. pLu %'T!*n VA$UU dM L 'fb f. f. 7 ] B, Isolation Valve Operability and B. Isolation Valve Testina Closing Time (Deleted) (Deleted) - O HATCH - UNIT 1 3.12-2 Amendment No. 22. EJ, 156 1 Y l t
.. . . - . . -. ~
l ADMINISTRATIVE CONTROLS -. A .lL : I ANNUAL REPORTS (Continued 6.9.1.5 Reports required on an annual basis shall include: , I
- a. A tabulation on an annual basis of the number of station, utility and other personnel, including contractors, receiving exposures
'{} greater than 100 mrem /yr and their associated man rem exposure according to work and job functions.* e.g., reactor operations i f and surveillance, inservice inspection, routine maintenance, special ;
, maintenance (describe maintenance), waste processing, and refueling. The dose assignment to various duty functions may be ; estimates based on pocket desimeter TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, b' at least of the total whole body dose received from external sources hal be assigned to specific major work functions. i f,6 f4 b Documentation of all challenges to safety / relief valves The results of specific activi analysis in which the primary coolant exceeded the limits of ecification 3.6.F.1. The a following information shall be in uded: (1) Reactor power i history starting 48 hours prior to he first sample in which the ' imit was exceeded; (2) Results of t last isotopic analysis for r ioiodine performed prior to exceedt the limit, results of analysis while limit was exceeded and re ults of one analysis after the h dioiodine activity was reduced to 1 s than limit. Each rib g resul should include date and time of samp ing and the radiciodine concent tions; (3) Cleanup system flow hist y starting 48 hours prior to he first sample in which the limit w s exceeded; (4) Graph , of the I-I concentration and one other radioi dine isotope l concentrati in microcuries per grarr, as a funct n of time for the i O duration of i specific activity above the stead state level; and (5) The time du tion when the specific activity a the primary coolant exceeded e radiciodine limit. AJ d. other u% unique re%s requireqv.mQbdy y f, 4, 2, U ANNUAL RADIOLOGICAL ENVIRONMENTAL 1.RVEILLANCE REPORTN is l 6.9.1.6 The Annual Radiological' Environmental Surveillance Report covering : the radiological environmental surveillance activities related to the plant - during the previous calendar year shall be submitted before Ma of each- i
,f,6. 7.- year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the i objectives outlined in the ODCH and Sections IV.B.2, IV.B.3,-and IV.C of l Appendix 1to10CFRPart50.g 42.
{.INSERTIf ! N a. A single submittal may be made for a multiple-unit station. The f,6 d submittal should combine those sections common to all units at the station. E la. 'This tabulation supplements the requirements of 20.2206 of 10 CFR Part 20. l l O HATCH - UNIT 1 6-15 Amendment No. 65, 4M, 440, 449,190 l l 2 vi IC) l I
UNIT 1 NO SIGNIFICANT HAZARDS DETERMINATION O l O
q NO SIGNIFICANT HAZARDS DETERMINATION Q ITS: SECTION 3.3.1.1 - RPS INSTRUMENTATION L.8 CHANGE The No Significant Hazards Determination evaluation is provided in Georgia Power Company letter from J.T. Beckham, Jr. to the NRC, dated September 20, 1993. The NRC issued this change as Amendment 193 to the Unit 1 Technical Specifications by letter dated March 15, 1994. O HATCH UNIT 1 9 REVISIONp'k
1 I NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.3.2.1 - CONTROL R0D BLOCK INSTRUMENTATION ) L.4 CHANGE 1 In accordance with the criteria set forth in 10 CFR 50.92, Georgia Power Company l has evaluated this proposed Technical Specifications change and determined it ' does not involve a significant hazards consideration based on the following
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
This change does not result in any hardware or operating procedure changes. The RBM is not assumed to be an initiator of any analyzed event. This change redefines the method for demonstrating OPERABILITY of the remaining channel when a channel is declared inoperable. Since the other channel remains OPERABLE, redefining the method by which the channel is demonstrated OPERABLE does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be installed) or changes (3 in parameters governing normal plant operation. The proposed change will (/ only redefine the method by which the remaining channel is verified OPERABLE when the other channel is declared inoperable. Redefining the , method by which a channel is demonstrated OPERABLE does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
This change allows credit to be taken for normal periodic surveillances as t a demonstration of OPERABILITY and availability of the remaining channel. - Thus, this change eliminates the requirement to perform surveillances on a channel when the other channel is declared inoperable. The periodic frequencies specified to demonstrate OPERABILITY of the remaining components have been shown to be adequate to ensure equipment OPERABILITY. As stated in NRC Generic letter 87-09, "It is overly conservative to assume that systems or components are inoperable when a surveillance O HATCH UNIT _I 7 1'M REVISION B
. ... . - . . . _ . .- . ..- ...-~..- -- - . - .
I NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.3.2.1 - CONTROL R00 BLOCK INSTRUMENTATION L4 CHANGE (continued, requirement has not been performed. The opposite is in fact the case; the vast majority of surveillances demonstrate the systems or components in fact are operable." Therefore, reliance on the specified surveillance ' intervals does not result in a reduced level of confidence concerning the equipment availability. l Therefore, reliance on the normal surveillance requirement is judged to be an equivalent testing program as compared to the requirements being deleted. Thus, this change does not involve a significant reduction in a margin of safety. , O O HATCH UNIT 1 ,2'M REVISION B
(gj NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.3.8.2 - RPS ELECTRIC POWER MONITORING L.4 CHANGE The No Significant Hazards Determination evaluation is prosided in GPC letter from J.T. Beckham, Jr. to the NRC, dated October 19, 1993. The NRC issued this change as Amendment 191 to the Unit 1 Technical Specifications by letter dated November 24, 1993. !O i 1 HATCH UNIT 1 4 REVISION ,A' l !
1 . . __ _ _ - _ _ . ._ . _ . _ _ _ _ _ . NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.1.1 - PRIMARY CONTAINMENT L.1 CHANGE (continued) r
- 3. Does this change involve a significant reduction in a margin of safety?
The proposed change does not involve a significant reduction in the margin of safety because leakage testing and structural limits will continue to be met based on the peak containment pressure resulting from a design basis accident. The peak containment internal pressure of 53.6 psig l continues to be within the containment internal maximum allowable pressure of 62 psig. There is no requirement for the test pressure to be higher , than the peak accident pressure. The proposed change to P, will not change the accident analyses and resultant radiological consequences for a postulated LOCA. The radiological consequences continue to be within the requirements of 10 CFR 100. The use of the revised P w leakage rate is measured and calculated appropriately.ill ensure that the O t I I O i HATCH UNIT 1 2 REVISION [
. _- . . - - . . . ~ . . . NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.8.4 - DC SOURCES-0PERATING L.3 CHANGE In accordance with the criteria set forth in 10 CFR 50.92, Georgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? .
The DC electrical power sources are used to support mitigation of the consequences of an accident; however, they are not considered the initiator of any previously analyzed accident. As such, the performance of a modified performance discharge test in lieu of a service discharge test will not increase the probability of any accident previously evaluated. The proposed SR continues to provide adequate assurance of OPERABLE batteries, since the modified performance discharge test represents a more severe test of battery capacity than does a service discharge test. Therefore, the proposed change does not involve an increase in the consequences of any accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not introduce a new mode of plant operation and does not involve physical modification to the plant. Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not created.
- 3. Does this change involve a significant reduction in a margin of safety?
This change does not involve a significant reduction in a margin of safety, since the proposed substitution of the modified performance discharge test for the service discharge test continues to provide adequate indication that the battery is capable of performing its design function. HATCH UNIT 1 3 REVISION [
O UNIT 2 IMPROVED TECHNICAL SP'sCIFICATIONS i l O O
PCIVs l 3.6.1.3 { SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.6.1.3.10 Verify leakage rate through each MSIV is ------NOTE----- s 100 scfh, and a combined maximum SR 3.0.2 is not i pathway leakage s 250 scfh for all four applicable main steam lines, when tested at --------------- l 2: 28.8 psig. However, the leakage rate acceptance In accordance criteria for the first test following with 10 CFR 50, l discovery of leakage through an MSIV not Appendix J, as j meeting the 100 scfh limit, shall be modified by i s 11.5 scfh for that MSIV. approved exemptions l SR 3.6.1.3.11 Replace the valve seat of each 18 inch 18 months purge valve having a resilient material seat. SR 3.6.1.3.12 Cycle each 18 inch excess flow isolation 18 months damper to the fully closed and fully open position. } l l l a HATCH UNIT 2 3.6-15 REVISION A
Drywell Pressure 3.6.1.4 3.6 CONTAINMENT SYSTEMS 3.6.1.4 Drywell Pressure h LC0 3.6.1.4 Drywell pressure shall be s 1.75 psig. l APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell pressure not A.1 Restore drywell I hour within limit. pressure to within limit. i B. Required Action and B.1 Be in MODE 5. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours l SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY l l SR 3.6.1.4.1 Verify drywell pressure is within limit. 12 hours i HATCH UNIT 2 3.C-16 REVISION [
I AC Sources - Operating 3.8.1 r~ ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. Two or more required D.1 Declare required 12 hours from offsite circuits feature (s) with no discovery of inoperable. offsite power Condition D available inoperable concurrent with when the redundant inoperability of required feature (s) redundant are inoperable. required feature (s) AND D.2 Restore all but one 24 hours required offsite circuit to OPERABLE status. E. One required offsite ------------NOTE------------- m circuit inoperable. Enter applicable Conditions / and Required Actions of V) AND LC0 3.8.7, " Distribution Systems - Operating," when One required DG Condition E is entered with inoperable. no AC power source to one 4160 V ESF bus. E.1 Restore required 12 hours offsite circuit to OPERABLE status. 0_E E.2 Restore required DG 12 hours to OPERABLE status. F. Two or more (Unit 2 F.1 Restore all but one 2 hours and swing) DGs Unit 2 and swing DGs inoperable. to OPERABLE status. (continued) HATCH UNIT 2 3.8-5 REVISION 1
AC Sources - Operating 3.8.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action and Associated Completion G.1 Be in MODE 3. 12 hours Time of Condition A, B, C, D, E, or F not AND met. G.2 Be in MODE 4. 36 hours H. One or more required H.1 Enter LCO 3.0.3. Immediately offsite circuits and two or more required DGs inoperable. DE Two or more required offsite circuits and one required DG inoperable. O l 1 i 1 O HATCH UNIT 2 3.8-6 REVISION A
l AC Sources - Operating l 3.8.1 l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY , SR 3.8.1.1 Verify correct breaker alignment and 7 days i indicated power availability for each required offsite circuit. , SR 3.8.1.2 -------------------NOTES------------------- i I. Performance of SR 3.8.1.5 satisfies this SR.
- 2. All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.
- 3. A modified DG start involving idling ,
and gradual acceleration to synchronous speed may be used for this > SR as recommended by the manufacturer. When modified start procedures are not used, the time, voltage, and frequency ! tolerances of SR 3.8.1.5.a must be i met.
- 4. For the swing DG, a single test will satisfy this Surveillance for both units, using the starting circuitry of Unit 2 and synchronized to 4160 V bus 2F for one periodic test,-and the !
starting circuitry of Unit I and < synchronized to 4160 V bus IF during i the next periodic test.
- 5. DG loadings may include gradual loading as recommended by the manufacturer.
- 6. Starting transients-above the upper voltage limit do not invalidate this test.
(continued) i O HATCH UNIT 2 3.8-/ [ REVISION ((' ; i 1
-. - ~ ,_. - - - - _ _ _ _ _ _ _ _ _ - - . - - - . _ _ _ _ _ _ - - - _ - _ _ _ _ . _ - - . . . - _ . -
AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.2 NOTES (continued)
- 7. Momentary transients outside the load range do not invalidate this test.
- 8. This Surveillance shall be conducted on only one DG at a time.
Verify each DG: As specified in
- a. Starts from standby conditions and Table 3.8.1-1 achieves steady state voltage 2 3740 V and s 4243 V and frequency 2 58.8 Hz and s 61.2 Hz; and
- b. Operates for 2 60 minutes at a load 21710 kW and s 2000 kW.
SR 3.8.1.3 Verify each day tank contains 2 900 gallons 31 days of fuel oil. SR 3.8.1.4 Check for and remove accumulated water from 184 days each day tank. (continued) O HATCH UNIT 2 3.8-8 REVISION A
.- - .-.- . - - - - -- -.- . - - ..-- - -= - . . - . - _ .
AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.5 -------------------NOTES-------------------
- 1. All DG starts may be preceded by an engine prelube period.
- 2. DG loadings may include gradual loading as recommended by the manufacturer.
- 3. Momentary load transients outside the load range do not invalidate this test.
- 4. This Surveillance shall be conducted l on only one DG at a time.
- 5. For the swing DG, a single test will l satisfy this Surveillance for both units, using the starting circuitry of Unit 2 and synchronized to 4160 V bus 2F for one periodic test and the
, starting circuitry of Unit 1 and synchronized to 4160 V bus IF during the next periodic test.
I i Verify each DG:
- a. Starts from standby conditions and achieves, in s 12 seconds, voltage 2 3740 V and frequency 2 58.8 Hz and 184 days after steady state conditions are reached, maintains voltage 2 3740 V and s 4243 V and frequency 2 58.8 Hz 1 and s 61.2 Hz; and
- b. Operates for a 60 minutes at a load 2 2764 kW and s 2825 kW for DG 2A, a 2360 kW and s 2425 kW for DG 18, and 2 2742 kW and s 2825 kW for DG 2C.
(continued) HATCH UNIT 2 3.8-9 REVISION [
l AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.6 ------------------NOTE--------------------- This Surveillance shall not be performed in MODE 1 or 2. However, credit may be taken for unplanned events that satisfy this SR. Verify automatic and manual transfer of 18 months unit power supply from the normal offsite circuit to the alternate offsite circuit. SR 3.8.1.7 ------------------NOTES--------------------
- 1. This Surveillance shall not be performed in MODE 1 or 2, except for the swing DG. For the swing DG, this Surveillance shall not be performed in MODE 1 or 2 using the Unit 2 controls.
Credit may be taken for unplanned events that satisfy this SR.
- 2. For the swing DG, a single test at the specified Frequency will satisfy this Surveillance for both units.
Verify each DG rejects a load greater than 18 months or equal to the single largest post-accident load, and:
- a. Following load rejection, the frequency is s 65.5 Hz; and
- b. Within 3 seconds followir.g load rejection, the voltage is 2 3740 V and s 4580 V.
(continued) l O HATCH UNIT 2 3.8-)1' lO REVISIONA
. . ~ -- .. .- _ . . . -
1 I AC Sources - Operating ! 3.8.1 l l SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY j SR 3.8.1.13 -------------------NOTES-------------------
- 1. This Surveillance shall be performed within 5 minutes of shutting down the DG after the DG has operated a 2 hours loaded 2 2565 kW. Momentary transients outside of load range do not invalidate this test.
- 2. All DG starts may be preceded by an engine prelube period.
- 3. For the swing DG, a single test at the specified Frecuency will satisfy this Surveillance for both units.
Verify each DG starts and achieves, in s 12 seconds, voltage 2 3740 V and frequency 2 58.8 Hz; and after steady state 18 months conditions are reached, maintains voltage 2 3740 V and s 4243 V and frequency 2 58.8 Hz and s 61.2 Hz. SR 3.8.1.14 -------------------NOTE-------------------- This Surveillance shall not be performed in MODE 1, 2, or 3. However, credit may be taken for unplanned events that satisfy this SR. Verify each DG: 18 months
- a. Synchronizes with offsite power source while loaded with emergency loads upon a simulated restoration of offsite power;
- b. Transfers loads to offsite power source; and
- c. Returns to ready-to-load operation.
(continued) HATCH VNIT 2 3.8-15 REVISION ,
AC Sources -- Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) ll SURVEILLANCE FREQUENCY SR 3.8.1.15 -------------------NOTE-------------------- This Surveillance shall not be performed in MODE 1, 2, or 3. However, credit may be taken for unplanned events that satisfy this SR. Verify with a DG operating in test mode and 18 months connected to its bus, an actual or simulated ECCS initiation signal overrides the test mode by:
- a. Returning DG to ready-to-load operation; and
- b. Automatically energizing the emergency load from offsite power.
SR 3.8.1.16 ------------------NOTE--------------------- This Surveillance shall not be performed in MODE 1, 2, or 3. However, credit may be taken for unplanned events that satisfy this SR. Verify interval between each sequenced 18 months load block is within 10% of design interval for each load sequence timing device. (continued) O HATCH UNIT 2 3.8-16 REVISION A
4 DC Sources - Operating , 3.8.4 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.4.7 -------------------NOTES-------------------
- 1. The modified performance discharge test ,
in SR 3.8.4.8 may be performed in lieu i of the service test in SR 3.8.4.7. L 2. This Surveillance shall not be ' performed in MODE 1, 2, or 3, except for the swing DG battery. However, credit may be taken for unplanned I events that satisfy this SR. Verify battery capacity is adequate to 18 months supply, and maintain in OPERABLE status, the required emergency loads for the design duty cycle when subjected to a br.ttery service test. I O SR 3.8.4.8 -------------------NOTE-------------------- This Surveillance shall not be performed in MODE 1, 2, or 3, except for the swing DG battery. However, credit may be taken for . , unplanned events that satisfy this SR. . Verify battery capacity is a 80% of the 60 months manufacturer's rating when subjected to a performance discharge test or a modified AND performance discharge test. (continued) O HATCH UNIT 2 3.8-31 REVISION,A'[ j
DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY - l SR 3.8.4.8 (continued) 12 months when battery shows l degradation or has reached 85% of expected life with capacity < 100% of manufacturer's - rating AND 24 months when battery has reached 85% of expected life with capacity 2 100% of manufacturer's rating SR 3.8.4.9 For required Unit 1 DC Sources, the SRs of In accordance Unit 1 Specification 3.8.4 are applicable. with applicable SRs O HATCH UNIT 2 3.8-32 REVISION A
Battery Cell Parameters 3.8.6 ( Table 3.8.6-1 (page 1 of 2) Battery Cell Parameter Requirements CATEGORY A: CATEGORY B: CATEGORY C: LIMITS FOR EACH LIMITS FOR EACH LIMITS DESIGNATED PILOT CONNECTED CELL FOR EACH PARAMETER CELL CONNECTED CELL Electrolyte > Minimum level > Minimum level Above top of Level indication mark, indication mark, plates, and not and s % inch above and s % inch above overflowing maximum level maximum level indication mark (a) indication mark (a) Float Voltage 2: 2.13 V 2: 2.13 V > 2.07 V Float Charging (b) (b) (b) ,rq Current V (a) It is acceptable for the electrolyte level to temporarily increase above the specified maximum level during equalizing charges provided it is not overflowing. (b) As applicable to each battery. O HATCH UNIT 2 3.8-39 REVISION A
Distribution Systems - Operating 3.8.7 3.8 ELECTRICAL POWER SYSTEMS 3.8.7 Distribution Systems - Operating LC0 3.8.7 The following AC and DC electrical power distribution cubsystems shall be OPERABLE:
- a. Unit 2 Division 1 and Division 2 and the swing bus AC and DC electrical power distribution subsystems; and
- b. Unit 1 AC and DC electrical power distribution subsystems needed to support equipment required to be OPERABLE by LCO 3.6.4.7, " Standby Gas Treatment (SGT)
System-Operating," LC0 3.7.4, " Main Control Room Environmental Control (MCREC) System," LCO 3.7.5,
" Control Room Air Conditioning (AC) System," and LC0 3.8.1, "AC Sources-0perating."
APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore required Unit 7 days Unit 1 AC or DC 1 AC and DC electrical power subsystem (s) to subsystems inoperable. OPERABLE status. B. One or more (Unit 2 or B.1 Restore DG DC 12 hours swing bus) DG DC electrical power electrical power distribution AND distribution subsystem to OPERABLE subsystems inoperable. status. 16 hours from discovery of failure to meet LC0 3.8.7.a (continued) O HATCH UNIT 2 3.8-4/O REVISIONp'
Distribution Systems - Operating 3.8.7 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. One or more (Unit 2 or C.1 Restore AC electrical 8 hours swing bus) AC power distribution electrical power subsystem to OPERABLE A_ND distribution status. subsystems inoperable. 16 hours from l discovery of failure to meet LCO 3.8.7.a ; l D. One Unit 2 station D.1 Restore Unit 2 2 hours service DC electrical station service DC power distribution electrical power AND subsystem inoperable. distribution subsystem to OPERABLE 16 hours from status. discovery of failure to meet LC0 3.8.7.a l E. Required Action and E.1 Be in MODE 3. 12 hours , associated Completion 4 Time of Condition A, MD B, C, or D not met. E.2 Se in MODE 4. 36 hours F. Two or more electrical F.1 Enter LC0 3.0.3. Immediately power distribution l subsystems inoperable l that result in a loss of function. O HATCH UNIT 2 3.8-4/_I_. REVISIONf
l Distribution Systems - Operating l 3.8.7 SURVEILLANCE REQUIREMENTS O' SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct breaker alignments and 7 days i voltage to required AC and DC electrical power distribution subsystems. O O HATCH UNIT 2 3.8-42 REVISION A
UNIT 2 IMPROVED BASES I O . l O
l l SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.5 and SR 3.3.1.2.6 (continued) REQUIREMENTS Although the Surveillance could be performed while on IRM Range 3, the plant would not be expected to maintain steady state operation at this power level. In this event, the 12 hour Frequency is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances. SR 3.3.1.2.7 Performance of a CHANNEL CALIBRATION at a frequency of 18 months verifies the performance of the SRM detectors and : associated circuitry. The Frequency considers the plant conditions required to perform the test, the ease of performing the test, and the likelihood of a change in the system or component status. The neutron detectors are excluded from the CHANNEL CALIBRATION (Note 1) because they cannot readily be adjusted. The detectors are fission , chambers that are designed to have a relatively constant sensitivity over the range and with an accuracy specified O for a fixed useful life. ; Note 2 to the Surveillance allows the Surveillance to be delayed until entry into the specified condition of the . Applicability. The SR must be performed in MODE 2 within 12 hours of entering MODE 2 with IRMs on Range 2 or below. The allowance to enter the Applicability with the 18 month Frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to perform the Surveillance while at higher power , level s. Although the Surveillance could be performed while on IRM Range 3, the plant would not be expected to maintain - steady state operation at this power level. In this event, the 12 hour Frequency is reasonable, based on the SRMs being , otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances. i REFERENCES 1. NRC Safety Evaluation Report for Amendment 125, April 30, 1993. O HATCH UNIT 2 B 3.3-41 REVISION A l I i l i
i Control Rod Block Instrumertatior. B 3.3.2.1 B 3.3 INSTRUMENTATION i t 8 3.3.2.1 Control Rod Block Instrumentation l l l BASES , I l BACKGROUND Control rods provide the primary means for control of I reactivity changes. Control rod block instrumentation includes channel sensors, logic circuitry, switches, and relays that are designed to ensure that specified feel design limits are not exceeded for postulated transients and accidents. During high power operation, the rod block monitor (RBM) provides protection for control rod withdrawal error events. During low power operations, control rod l blocks from the rod worth minimizer (RWM) enforce specific l control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA). During shutdown conditions, control rod blocks from the Reactor Mode Switch - Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent criticalities. The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint during control rod manipulations. It is assumed to function
- to block further control rod withdrawal to preclude a MCPR l Safety Limit (SL) violation. The RBM supplies a trip signal l to the Reactor Manual Control System (RMCS) to appropriately l inhibit control rod withdrawal during power operation above i the low power range setpoint. The RBM has two channels, I either of which can initiate a control rod block when the channel output exceeds the control rod block setpoint. One RBM channel inputs into one RMCS rod block circuit and the other RBM channel inputs into the second RMCS rod block circuit. The RBM channel signal is generated by averaging a set of local power range monitor (LPRM) signals at various core heights surrounding the control rod being withdrawn. A signal from one average power range monitor (APRM) channel assigned to each Reactor Protection System (RPS) trip system supplies a reference signal for the RBM channel in the same trip system. This reference signal is used to determine which RBM range setpoint (low, intermediate, or high) is enabled. If the APRM is indicating less than the low power range setpoint, the RBM is automatically bypassed. The RBM is also automatically bypassed if a peripheral control rod is selected (Ref. 1). A rod block signal is also generated I
if an RBM Downscale trip or an Inoperable trip occurs. The Downscale trip will occur if the RBM channel signal (continued) HATCH UNIT 2 B 3.3-42 REVISION A l l
Control Rod Block Instrumentation B 3.3.2.1 BASES BACKGROUND decreases below the Downscale trip setpoint after the RBM (continued) signal has been normalized. The Inoperable trip will occur during the nulling (normalization) sequence if: the RBM channel fails to null, too few LPRM inputs are available, a module is not plugged in, or the function switch is moved to any position other than " Operate." The Bypass Time Delay ensures that the normalized signal is passed to the trip logic within the appropriate time. The delay is between the time the signal is nulled to the reference and the signal is passed to the trip logic. l (continued) HATCH UNIT 2 B3.3-4/j/7 REVISION A A : I I
Control Rod Block Instrumentation B 3.3.2.1 BASES BACKCROUND The purpose of the RWM is to control rod patterns during (continued) startup and shutdown, such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods inserted to 10% RTP.
- The sequences effectively limit the potential amount and ,
rate of reactivity increase during a CRDA. Prescribed ' control rod sequences are stored in the RWM, which will , initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the stored sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses i feedwater flow and steam flow signals to determine when the ! reactor power is above the preset power level at which the i RWM is automatically bypassed (Ref. 2). The RWM is a single I channel system that provides input into both RMCS rod block circuits. With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This Function prevents inadvertent criticality as the result of a " control rod withdrawal during MODE 3 or 4, or during MODE 5 ' when the reactor mode switch is required to be in the. O shutdown position. The reactor mode switch has two channels, each inputting into a separate RMCS rod block circuit. A rod block in either RMCS circuit will provide a , control rod block to all control rods. APPLICABLE 1. Rod Block Monitor SAFETY ANALYSES, ; LCO, and The RBM is designed to prevent violation of the MCPR , APPLICABILITY SL and the cladding 1% plastic strain fuel design limit that . may result from a single control rod withdrawal error (RWE) event. The analytical methods and assumptions'used in evaluating the RWE event are summarized in Reference 3. A , statistical analysis of RWE events was performed to i determine the RBM response for both channels for each event. From these responses, the fuel thermal performance as a , function of RBM Allowable Value was determined. The Allowable Values are chosen as a function of power level. Based on the specified Allowable Values, operating limits are established. (continued) HATCH UNIT 2 8 3.3-43 REVISION A
Control Rod Block instrumentation B 3.3.2.1 BASES APPLICABLE 1. Rod Block Monitor (continued) SAFETY ANALYSES, LC0, and The RBM Function satisfies Criterion 3 of the NRC Policy APPLICABILITY Statement (Ref. 9). Two channels of the RBM are required to be OPERABLE, with their setpoints within the appropriate Allowable Values, to l ensure that no single instrument failure can preclude a rod block from this Function. The setpoints are calibrated consistent with applicable setpoint methodology (nominal trip setpoint). Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values i between successive CHANNEL CALIBRATIONS. Operation with a I trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor power), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived I from the analytic limits, corrected for calibration, l process, and some of the instrument errors. The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived l in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for. The RBM is assumed to mitigate the consequences of an RWE event when operating a 29% RTP. Below this power level, the consequences of an RWE event will not exceed the MCPR SL and, therefore, the RBM is not required to be OPERABLE l (Ref. 3). When operating < 90% RTP, analyses (Ref. 3) have shown that with an initial MCPR 21.70, no RWE event will result in exceeding the MCPR SL. Also, the analyses demonstrate that when operating at a 90% RTP with MCPR 21.40, no RWE event will result in exceeding the MCPR (continued) h HATCH UNIT 2 B3.3-4/4 REVISIONgg
RCS Specific Activity B 3.4.6 BASES O ACTIONS A.1 and A.2 (continued) restore the activity level provides a reasonable time for temporary coolant activity increases (iodine spikes or crud bursts) to be cleaned up with the normal processing systems. A Note to the Required Actions of Condition A excludes the MODE change restriction of LC0 3.0.4. This exception allows entry into the applicable MODE (S) while relying on the ACTIONS even though the ACTIONS may eventually require plant-shutdown. This exception is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operation. B.1. B.2.1. B.2.2.1. and B.2.2.2 If the DOSE EQUIVALENT I-131 cannot be restored to s 0.2 Ci/gm within 48 hours, or if at any time it is > 4.0 + Ci/gm, it must be determined at least once every 4 hours and all the main-steam lines must be isolated within O 12 hours. Isolating the main steam lines precludes the possibility of releasing radioactive material to the environment in an amount that is not well within the requirements of 10 CFR 100 during a postulated MSLB accident. Alternatively, the plant can be placed in MODE 3 within 12 hours and in MODE 4 within 36 hours. This option is provided for those instances when isolation of main steam lines is not desired (e.g., due to the decay heat loads). . In MODE 4, the requirements of the LC0 are no longer , applicable. The Completion Time of once every 4 hours is the time needed to take and analyze a sample. The 12 hour Completion Time is reasonable, based on cperating experience, to isolate the main steam lines in an orderly manner and without challenging plant systems. Also, the allowed Completion Times for Required Actions B.2.2.1 and B.2.2.2 for placing the unit in MODES 3 and 4 are reasonable, based on operating experience, to achieve the required plant conditions fr'sm i (continued) HATCH UNIT 2 B 3.4-31 REVISION A
RCS Specific Activity B 3.4.6 BASES ACTIONS B.l. B.2.1. B.2.2.1. and B.2.2.2 (continued) full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This Surveillance is performed to ensure iodine remains within limit during normal operation. The 7 day Frequency is adequate to trend changes in the iodine activity level. This SR is modified by a Note that requires this Surveillance to be performed only in MODE 1 because the level of fission products generated in other MODES is much less. REFERENCES 1. 10 CFR 100.11.
- 2. FSAR, Section 15.1.40. I
- 3. NEDE-240ll-P-A-9-US, "GE Standard Application for Reactor Fuel," Supplement for United States, September 1988.
- 4. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
O HATCH UNIT 2 B 3.4-32 REVISION A'
Primary Containment B 3.6.1.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment BASES BACKGROUND The function of the primary containment is to isolate and contain fission products released from the Reactor Primary System following a Design Basis Accident (DBA) and to confine the postulated release of radioactive material. The primary containment consists of a steel lined, reinforced concrete vessel, which surrounds the Reactor Primary System and provides an essentially leak tight barrier against an uncontrolled release of radioactive material to the environment. The isolation devices for the penetrations in the primary containment boundary are a part of the containment leak tight barrier. To maintain this leak tight barrier:
- a. All penetrations required to be closed during accident conditions are either:
- 1. capable of being closed by an OPERABLE automatic containment isolation system, or
- 2. closed by manual valves, blind flanges, or de-activated automatic valves secured in their closed positions, except as provided in LC0 3.6.1.3, " Primary Containment Isolation Valves (PCIVs);"
- b. The primary containment air lock is OPERABLE, except as provided in LC0 3.6.1.2, " Primary Containment Air Lock"; and
- c. All equipment hatches are closed.
This Specification ensures that the performance of the primary containment, in the event of a DBA, meets the assumptions used in the safety analyses of References 1 and 2. SR 3.6.1.1.1 leakage rate requirements are in conformance with 10 CFR 50, Appendix J (Ref. 3), as modified by approved exemptions. (continued) HATCH VNIT 2 B 3.6-1 REVISION A
Primary Containment B 3.6.1.1 BASES (continued) APPLICABLE The safety design basis for the primary containment is that SAFETY ANALYSES it must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate. The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage. Analytical methods and assumptions involving the primary containment are presented in References 1 and 2. The safety analyses assume a nonmechanistic fission product release following a DBA, which forms the basis for determination of offsite doses. The fission product release is, in turn, based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not exceeded. The maximum allowable leakage rate for the primary containment (L ) is 1.2% by weight of the containment air per 24 hours al the maximum peak containment pressure (P,) of 48.7 psig (Ref.1). l Primary containment satisfies Criterion 3 of the NRC Policy Statement (Ref. 4). LCO Primary containment OPERABILITY is maintained by limiting leakage to less than L,, except prior to the first startup after performing a required 10 CFR 50, Appendix J, leakage test. At this time, the combined Type B and C leakage must be < 0.6 L,, and the overall Type A leakage must be < 0.75 L,. Compliance with this LC0 will ensure a primary containment configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analyses. Individual leakage rates specified for the primary containment air lock are addressed in i.00 3.6.1.2. (continued) HATCH UNIT 2 B 3.6-2
, REVISION ((
Primary Containment Air Lock B 3.6.1.2 BASES BACKGROUND containment leakage rate to within limits in the event of a (continued) DBA. Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in the unit safety analysis. APPLICABLE The DBA that postulates the maximum release of radioactive SAFETY ANALYSES material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment is designed with a maximum allowable leakage rate (L ) of 1.2% byweightofthecontainmentairper24hoursatthe calculated maximum peak containment pressure (P,) of 48.7 psig (Ref. 2). This allowable leakage rate forms the l basis for the acceptance criteria imposed on the SRs associated with the air lock. Primary containment air lock OPERABILITY is also required to minimize the amount of fission product gases that may escape ,q primary containment through the air lock and contaminate and V pressurize the secondary containment. The primary containment air lock satisfies Criterion 3 of the NRC Policy Statement (Ref. 4). . LC0 As part of primary containment, the air lock's safety function is related to control of containment leakage rates j following a DBA. Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event. The primary containment air lock is required to be OPERABLE. For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be , in compliance with the Type B air lock leakage test, and I both air lock doors must be OPERABLE. The interlock allows I only one air lock door to be opened at a time. This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be l (continued) HATCH UNIT 2 B 3.6-7 REVISIONg
Primary Containment Air Lock B 3.6.1.2 BASES l l 2 LC0 OPERABLE. Closure of a single door a each air lock is I (continued) sufficient to provide a leak tight .arrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry and exit from primary containment. ' APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the primary containment air lock is not required to be OPERABLE in MODES 4 and 5 to prevent leakage of radioactive material from primary containment. ACTIONS The ACTIONS are modified by Note 1, which allows entry and exit to perform repairs of the affected air lock component. If the outer door is inoperable, then it may be easily - accessed to repair. If the inner door is the one that is inoperable, however, then a short time exists when the & containment boundary is not intact (during access through W the outer door). The allowance to open the OPERABLE door, even if it means the primary containment boundary is temporarily not intact, is acceptable due to the low probability of an event that could pressurize the primary containment during the short time in which the OPERABLE door is expected to be open. The OPERABLE door must be immediately closed after each entry and exit. The ACTIONS are modified by a second Note, which ensures appropriate remedial measures are taken, if necessary, if air lock leakage results in exceeding overall containment leakage rate acceptance criteria. Pursuant to LCO 3.0.6, actions are not required, even if primary containment is exceeding its leakage limit. Therefore, the Note is added to require ACTIONS for LC0 3.6.1.1, " Primary Containment," to be taken in this event. (continued) HATCH UNIT 2 B 3.6-8 REVISION A
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.10 (continued) REQUIREMENTS The Frequency is required by 10 CFR 50, Appendix J, as modified by approved exemptions; thus, SR 3.0.2 (which allows Frequency extensions) does not apply. SR 3.6.1.3.11 The valve seats of each 18 inch purge valve (supply and exhaust) having resilient material seats must be replaced every 18 months. This will allow the opportunity for repair before gross leakage failure develops. The 18 month Frequency is based on engineering judgment and operational experience which shows that gross leakage normally does not occur when the valve seats are replaced on an 18 month Frequency. SR 3.6.1.3.12 The Surveillance Requirement provides assurance that the
^ excess flow isolation dampers can close following an isolation signal. The 18 month Frequency is based on vendor recommendations and engineering judgment. 0perating experience has shown that these dampers usually pass the Surveillance when performed at the 18 month Frequency.
Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. FSAR, Chapter 15.
- 2. Technical Requirements Manual.
I
- 3. FSAR, Section 15.1.39. l 1
- 4. FSAR, Section 6.2.
- 5. 10 CFR 50, Appendix J.
- 6. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
O HATCH UNIT 2 B 3.6-27 REVISION A
l Drywell Pressure l B 3.6.1.4 8 3.6 CONTAINMENT SYSTEMS B 3.6.1.4 Drywell Pressure l BASES l BACKGROUND The drywell pressure is limited during normal operations to l preserve the initial conditions assumed in the accident analysis for a Design Basis Accident (DBA) or loss of coolant accident (LOCA). l APPLICABLE Primary containment performance is evaluated for the entire SAFETY ANALYSES spectrum of break sizes for postulated LOCAs (Ref. 1). Among the inputs to the DBA is the initial primary containment internal pressure (Ref.1). Analyses assume an initial drywell pressure of 1.75 psig. This limitation l ensures that the safety analysis remains valid by maintaining the expected initial conditions and ensures that l the peak LOCA drywell internal pressure does not exceed the I maximum allowable of 62 psig. I The maximum calculated drywell pressure occurs during the reactor blowdown phase of the DBA, which assumes an W instantaneous recirculation line break. The calculated peak drywell pressure for this limiting event is 48.7 psig l (Ref. 1). Drywell pressure satisfies Criterion 2 of the NRC Policy Statement (Ref. 2). l LC0 In the event of a DBA, with an initial drywell pressure ! s 1.75 psig, the resultant peak drywell accident pressure l will be maintained below the drywell design pressure. APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of l radioactive material to primary containment. In MODES 4 l (continued) HATCH UNIT 2 B 3.6-28 REVISION ,A
Primary Containment Hydrogen Recombiners .I B 3.6.3.1 BASES (continued) LCO Two primary containment hydrogen recombiners must be l OPERABLE. This ensures operation of at lea _st one primary containment hydrogen recombiner subsystem in the event of a worst case single active failure. Operation with at least one primary containment hydrogen , recombiner subsystem ensures that the post-LOCA hydrogen concentration can be prevented from exceeding the flammability limit. APPLICABILITY In MODES 1 and 2, the two primary containment hydrogen recombiners are required to control the hydrogen concentration within primary containment below its flammability limit of 4.0 v/o following a LOCA, assuming a l worst case single failure. In M0DE 3, both the hydrogen production rate and the total hydrogen produced after a LOCA would be less than that ! calculated for the DBA LOCA. Also, because of the limited , time in this MODE, the probability of an accident requiring : p the primary containment hydrogen recombiner is low. g' Therefore, the primary containment hydrogen recombiner is l not required in MODE 3. , In MODES 4 and 5, the probability and consequences of a LOCA are low due to the pressure and temperature limitations in these MODES. Therefore, the primary containment hydrogen recombiner is not required in these. MODES. ACTIONS A.1 With one primary containment hydrogen recombiner inoperable,- i the inoperable recombiner must be restored to OPERABLE . status within 30 days. In this Condition, the remaining 0PERABLE recombiner is adequate to perform the hydrogen control function. However, the overall reliability is reduced because a single failure in the OPERABLE recombiner could result in reduced hydrogen control capability. The 30 day Completion Time is based on the low probability of the occurrence of a LOCA that would generate hydrogen in amounts capable of exceeding the flammability limit, the amount of time available after the event for operator action l (Continued) HATCH UNIT 2 B 3.6-69 REVISION A
Primary Containment Hydrogen Recombiners l B 3.6.3.1 i i l BASES i ACTIONS A.1 (continued) to prevent exceeding this limit, and the low probability of failure of the OPERABLE primary containment hydrogen recombiner. Required Action A.1 has been modified by a Note indicating that the provisions of LC0 3.0.4 are not applicable. As a result, a MODE change is allowed when one recombiner is inoperable. This allowance is provided because of the low probability of the occurrence of a LOCA that would generate hydrogen in amounts capable of exceeding the flammability limit, the low probability of the failure of the OPERABLE subsystem, and the amount of time available after a postulated LOCA for operator action to prevent exceeding the flammability limit. B.1 and B.2 With two primary containment hydrogen recombiners inoperable, the ability to perform the hydrogen control function via alternate capabilities must be verified by administrative means within I hour. The alternate hydrogen control capabilities are provided by the Primary Containment Purge System or the Nitrogen Inerting System. The 1 hour Completion Time allows a reasonable period of time to verify that a loss of hydrogen control function does not exist. In addition, the alternate hydrogen control system capability must be verified once per 12 hours thereafter to ensure its l continued availability. Both the initial verification and all subsequent verifications may be performed as an administrative check by examining logs or other information to determine the availability of the alternate hydrogen control system. It does not mean to perform the Surveillances needed to demonstrate OPERABILITY of the alternate hydrogen control system. If the ability to perform the hydrogen control function is maintained, continued operation is permitted with two hydrogen recombiners inoperable for up to 7 days. Seven days is a reasonable time to allow two hydrogen recombiners to be inoperable because the hydrogen control function is maintained and because of the low probability of the occurrence of a LOCA that would generate hydrogen in amounts capable of exceeding the flammability limit. (continued) A HATCH UNIT 2 B 3.6-70 REVISION A',P w
AC Sources - Operating 8 3.8.1 O BASES O SURVEILLANCE SR 3.8.1.2 (continued) REQUIREMENTS Note 6 modifies the Surveillance by stating that starting transients above the upper voltage limit do not invalidate this test. Notes 7 modifies this Surveillance by stating that momentary load transients because of changing bus loads do not invalidate this test. Note 8 indicates that this Surveillance is required to be conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations. The normal 31 day Frequency for SR 3.8.1.2 (see Table 3.8.1-1, " Diesel Generator Test Schedule") is consistent with Regulatory Guide 1.108 (Ref. 9). This Frequency provides adequate assurance of DG OPERABILITY, while minimizing degradation resulting from testing. SR 3.8.1.3 This SR provides verification that the level of fuel oil in the day tank is at or above the level at which fuel oil is automatically added. The level is expressed as an equivalent volume in gallons, and is selected to ensure adequate fuel oil for a minimum of I hour of DG operation at full load plus 10%. The actual amount required to meet the SR (900 gallons) will provide approximately 3.5 hours of DG operation at full load. The 31 day Frequency is adequate to ensure that a sufficient supply of fuel oil is available, since low level alarms are provided and operators would be aware of any large uses of fuel oil during this period. j SR 3.8.1.4 l Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the fuel oil day tanks once every 184 days eliminates the necessary environment for bacterial survival. (continued) HATCH UNIT 2 B 3.8-21 REVISION ,A'
~
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.4 (continued) REQUIREMENTS This is a means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment e (continued) HATCH UNIT 2 REVISION [ B3.8-27 1A
l AC Sources - Operating i B 3.8.1 BASES g SURVEILLANCE SR 3.8.1.4 (continued) REQUIREMENTS in the fuel oil during DG operation. Water in the day tank may come from condensation, rain water, contaminated fuel oil, and breakdown of the fuel oil by bacteria. Checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. The Surveillance Frequency is based on engineering judgment and has shown to be acceptable through operating experience. This SR is for preventive maintenance. The presence of water does not necessarily represent a failure of this SR provided that accumulated water is removed during performance of this Surveillance. SR 3. 8.1.li This SR helps to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and maintain the unit in a safe shutdown condition. This Surveillance verifies that the DGs are capable of a " fast cold" start, synchronizing, and accepting a load more closely simulating accident loads. A minimum run time of 60 minutes is required to stabilize engine temperatures, while minimizing the time that the DG is connected to the offsite source. SR 3.8.1.5 requires that, at a 184 day Frequency, the DG starts from standby conditions and achieves required voltage and frequency within 12 seconds. The 12 second start requirement supports the assumptions in the design basis LOCA analysis of FSAR, Chapter 6 (Ref. 4). For the purposes of this testing, the DGs are started from standby conditions. Standby conditions for a DG mean that the diesel engine coolant and oil are being continuously circulated and temperature is being maintained consistent with manufacturer recommendations. Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.8 lagging and 1.0. The 0.8 value is the design rating of the machine, while 1.0 is an operational limitation. (continued) HATCH UNIT 2 B 3.8-22 REVISION A
l AC Sources - Operating ; B 3.8.1 ' (9 BASES v SURVEILLANCE SR 3.8.1.5 (continued) REQUIREMENTS The 184 day Frequency for SR 3.8.1.5 is a reduction in cold testing consistent with Generic Letter 84-15 (Ref. 7). This Frequency provides adequate assurance of DG OPERABILITY, while minimizing degradation resulting from testing. To minimize the wear on moving parts that do not get lubricated when the engine is not running, this SR has been modified by a Note (Note 1) to indicate that all DG starts for this Surveillance may be preceded by an engine prelube period and followed by a warmup prior to loading. Note 2 modifies this Surveillance to indicate that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized. Note 3 modifies this Surveillance by stating that momentary voltage or load transients because of changing bus loads do not invalidate this test. p/ Note 4 indicates that this Surveillance is required to be l U conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations. To minimize testing of the swing DG, Note 5 allows a single l test (instead of two tests, one for each unit) to satisfy the requirements for both units, with the DG started using the starting circuitry of one unit and synchronized to the ESF bus of that unit for one periodic test and started using the starting circuitry of the other unit and synchronized to the ESF bus of that unit during the next periodic test. This is allowed since the main purpose of the Surveillance, to ensure DG OPERABILITY, is still being verified on the proper frequency, and each unit's starting circuitry and breaker control circuitry, which is only being tested every second test (due to the staggering of the tests), historically have a very low failure rate. If the swing DG ! fails one of these Surveillances, the DG should be considered inoperable en both units, unless the cause of the failure can be directly related to only one unit. f i (continued) HATCH UNIT 2 B 3.8-2/ % REVISION A'
/J !t
AC Sources - Operating B 3.8.1 BASES h SURVEILLANCE SR 3.8.1.6 REQUIREMENTS (continued) Transfer of each 4.16 kV ESF bus power supply from the normal offsite circuit to the alternate offsite circuit demonstrates the OPERABILITY of the alternate circuit distribution network to power the shutdown loads. The 18 month Frequency of the Surveillance is based on engineering judgment taking into consideration the plant conditions required to perform the Surveillance, and is intended to be consistent with expected fuel cycle lengths. Operating experience has shown that these components usually pass the SR when performed on the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. This SR is modified by a Note. The reason for the Note is that, during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, plant safety systems. Credit may be taken for unplanned events that satisfy this SR. This Surveillance tests the applicable logic associated with the Unit 2 swing bus. The comparable test specified in the Unit 1 Technical Specifications tests the applicable logic associated with the Unit 1 swing bus. Consequently, a test must be performed within the specified Frequency for each unit. The Note specifying the restriction for not performing the test while the unit is in MODE 1 or 2 does not have applicability to Unit 1. As the Surveillance represents separate tests, the Unit 2 Surveillance should not be performed with Unit 2 in MODE 1 or 2 and the Unit I test should not be performed with Unit 1 in MODE 1 or 2. SR 3.8.1.7 Each DG is provided with an engine overspeed trip to prevent damage to the engine. Recovery from the transient caused by the loss of a large load could cause diesel engine overspeed, which, if excessive, might result in a trip of the engine. This Surveillance demonstrates the DG load response characteristics and capability to reject the largest single load without exceeding predetermined voltage and frequency and while maintaining a specified margin to (continued) HATCH UNIT 2 B 3.8-24 REVISION A
AC Sources - Operating B 3.8.1 [ BASES SURVEILLANCE SR 3.8.1.7 (continued) REQUIREMENTS the overspeed trip. The largest single load for each DG is a residual heat removal service water pump at rated flow (1225 bhp). This Surveillance may be accomplished by either a.) tripping the DG output breaker with the DG carrying greater than or equal to the largest single load while paralleled to offsite power or while solely supplying the ; bus, or b.) tripping the largest single load with the DG i solely sup:: lying the bus. Although Plant Hatch Unit 2 is not committad to IEEE-387-1984, (Ref. 11), this SR is consistent with the IEEE-387-1984 requirement that states the load rejection test is acceptable if the increase in diesel speed does not exceed 75% of the difference between synchronous speed and the overspeed trip setpoint, or 15% above synchronous speed, whichever is lower. For all DGs, this represents 65.5 Hz, equivalent to 75% of the difference between nominal speed and the overspeed trip setpoint. The voltage and frequency specified are consistent with the nominal range for the DG. SR 3.8.1.7.a corresponds to the , maximum frequency excursion, while SR 3.8.1.7.b is the voltage to which the DG must recover following load O rejection. The 18 month Frequency is consistent with the recommendation of Reg 11 atory Guide 1.108 (Ref. 9). This SR is modified by two Notes. The reason for Note 1 is that, during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, plant safety systems. Credit may be taken for unplanned events that satisfy this SR. In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as possible, testing is performed with only the DG providing power to the associated 4160 V ESF bus. The DG is not synchronized with offsite power. 1 To minimize testing of the swing DG, Note 2 allows a single test (instead of two tests, one for each unit) to satisfy the requirements for both units. This is allowed since the main purpose of the Surveillance can be met by performing O (continued) G HATCH UNIT 2 B3.8-2p6 REVISION A, A
l AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.7 (continued) REQUIREMENTS the test on either unit (no unit specific DG components are being tested). If the swing DG fails one of these Surveillances, the DG should be considered inoperable on both units, unless the cause of the failure can be directly related to only one unit. O (continued) HATCH UNIT 2 B 3.8-27~fi,
. REVISION ,A ^
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.8 REQUIREMENTS (continued) This Surveillance demonstrates the DG capability to reject a full load without overspeed tripping or exceeding the predetermined voltage limits. The DG full load rejection may occur because of a system fault or inadvertent breaker tripping. This Surveillance ensures proper engine generator load response under the simulated test conditions. This test simulates the loss of the total connected load that the DG experiences following a full load rejection and verifies that the DG does not trip upon loss of the load. These acceptance criteria provide DG damage protection. While the DG is not expected to experience this transient during an event, and continues to be available, this response ensures that the DG is not degraded for future application, including reconnection to the bus if the trip initiator can be corrected or isolated. In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as possible, testing must be performed using a power factor s 0.88. This power factor is chosen to be representative of the actual design basis inductive loading that the DG would g experience. W The 18 month Frequency is consistent with the recommendation of Regulatory Guide 1.108 (Ref. 9) and is intended to be consistent with expected fuel cycle lengths. This SR is modified by four Notes. Note I states that momentary transients due to changing bus loads do not invalidate this test. The reason for Note 2 is that during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that would challenge continued steady state operation and, as a result, plant safety systems. Credit may be taken for unplanned events that satisfy this SR. Note 3 is provided in recognition that if the offsite electrical power distribution system is lightly loaded (i.e., system voltage is high), it may not be possible to raise voltage without creating an overvoltage condition on the ESF bus. Therefore, to ensure the bus voltage, supplied ESF loads, and DG are not placed in an unsafe condition during this test, the power factor limit does not have to be met if grid voltage or ESF bus loading does not permit the power factor limit to be met when the DG is tied to the (continued) HATCH UNIT 2 B 3.8-26 REVISION A l
)
b
Diesel Fuel Oil and Transfer, Lube Oil, and Starting Air B 3.8.3 BASES SURVEILLANCE SR 3.8.3.7 (continued) REQUIREMENTS system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for manual fuel transfer are OPERABLE. Since the fuel oil transfer pumps are being tested on a 31 day Frequency in accordance with SR 3.8.3.5, the 18 month Frequency has been determined to be acceptable based on engineering judgement and operating experience. REFERENCES 1. FSAR, Section 9.5.4.
- 2. FSAR, Chapter 6.
- 3. FSAR, Chapter 15.
- 4. NRC No. 93-102, " Final Policy Statement of Technical Specification Improvements," July 23, 1993.
I o [ \ HATCH UNIT 2 B 3.8-55 REVISION A
DC Sources - Operating B 3.8.4 9 3.8 ELECTRICAL POWER SYSTEMS B 3.8.4 DC Sources - Operating BASES BACKGROUND The DC electrical power system provides the AC emergency power system with control power. It also provides both motive and control power to selected safety related equipment. As required by 10 CFR 50, Appendix A, GDC 17 (Ref.1), the DC electrical power system is designed to have sufficient independence, redundancy, and testability to perform its safety functions, assuming a single failure. The DC electrical power system also conforms to the recommendations of Regulatory Guide 1.6 (Ref. 2) and IEEE-308 (Ref. 3). The station service DC power sources provide both motive and control power to selected safety related and nonsafety related equipment. Each DC subsystem is energized by one 125/250 V station service battery and three 125 V battery chargers (two normally inservice chargers and one standby charger). Each battery is exclusively associated with a single 125/250 VDC bus. Each set of battery chargers exclusively associated with a 125/250 VDC subsystem cannot be interconnected with any other 125/250 VDC subsystem. The normal and backup charcers are supplied from the same AC load groups for which the associated DC subsystem supplies the control power. The loads between the redundant 125/250 VDC subsystem are not transferable except for the Automatic Depressurization System, the logic circuits and valves of which are normslly fed from the Division 1 DC system. The diesel generator (DG) DC power sources provide control and instrumentation power for their respective DG and their respective offsite circuit supply breakers. In addition, DG 2A power source provides circuit breaker control power for the respective Division I loads on 4160 VAC buses 2E and 2F, and DG 2C power source provides circuit breaker control power for the respective Division II loads on 4160 VAC buses 2F and 2G. Each DG DC subsystem is energized by one 125 V battery and two 125 V battery chargers (one normally inservice charger and one standby charger). (continued) HATCH UNIT 2 REVISIONh B3.8-5/[" _a
DC Sources - Operating B 3.8.4 BASES BACKGROUND During normal operation, the DC loads are powered from the (continued) respective station service and DG battery chargers with the batteries floating on the system. 1 I i O i l l (continued) HATCH UNIT 2 B 3.8,6( 6/C k REVISION A'; 2 =2
l DC Sources-Operating l B 3.8.4 BASES ACTIONS L1 (continued) If one of the required DC electrical power subsystems is inoperable (e.g., inoperable battery, inoperable battery charger (s), or inoperable battery charger and associated inoperable battery), the remaining DC electrical power subsystems have the capacity to support a safe shutdown and to mitigate an accident condition. Since a subsequent f postulated worst case single failure could result in the loss of minimum necessary DC electrical subsystems to mitigate a postulated worst case accident, continued power operation should not exceed 2 hours. The 2 hour Completion Time is based on Regulatory Guide 1.93 (Ref. 7) and reflects a reasonable time to assess unit status as a function of the inoperable DC electrical power subsystem and, if the DC electrical power subsystem is not restored to OPERABLE status, to prepare to effect an orderly and safe unit shutdown. D.1 and D.2 If the DC electrical power subsystem cannot be restored to y OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LC0 does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 12 hours and to MODE 4 within l 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. The Completion Time to bring the unit to MODE 4 is consistent with the time required in Regulatory Guide 1.93 (Ref. 7). El Condition E corresponds to a level of degradation in the DC electrical power subsystems that causes a required safety i function to be lost. When more than one DC source is lost, and this results in the loss of a required function, the plant is in a condition outside the accident analysis. Therefore, no additional time is justified for continued operation. LC0 3.0.3 must be entered immediately to commence a controlled shutdown. O' (continued) HATCH UNIT 2 B 3.8-61 REVISION A
DC Sources - Operating B 3.8.4 BASES (c.ontinued) SURVEILLANCE SR 3.8.4.1 REQUIREMENTS Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended function. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery (or battery cell' , maintain the battery (or a battery cell) in a fully L . . state. Voltage requirements are based on the nominal de .,.. voltage of the battery and are consistent with the initial voltages assumed in the battery sizing calculations. The voltage requirement for battery terminal voltage is based on the oper circuit voltage of a lead-calcium cell of nominal 1.215 specific gravity. Without regard to other battery r > meters, this voltage is indicative of a battery se? 1 capable of performing its required safety function . 7 day Frequency is consistent with manufacturer's recommendations and IEEE-450 (Ref. 8). SR 3.8.4.2 Visual inspection to detect corrosion of the battery cells and connections, or measurement of the resistance of each inter-cell, inter-rack 'ter-tier, and terminal connection, provides an indication hysical damage or abnormal deterioration that coulo vo tentially degrade battery performance. The connection resistance limits are established to maintain connection resistance as low as reasonably possible to minimize the overall voltage drop across the battery and the possibility of battery damage due to heating of connections. The resistance values for each battery connection are located in the Technical Requirements Manual (Reference 9). The Frequency for these inspections, which can detect conditions that can cause power losses due to resistance heating, is 92 days. This Frequency is considered acceptable based on operating experience related to detecting corrosion trends. SR 3.8.4.3 Visual inspection of the battery cells, cell plates, and battery racks provides an indication of physical damage or (continued) h HATCH UNIT 2 B3.8-6% REVISION h-
DC Sources - Operating 8 3.8.4 BASES . SURVEILLANCE SR 3.8.4.3 (continued) REQUIREMENTS abnormal deterioration that could potentially degrade battery performance. The 18 month Frequency of the Surveillance is based on engineering judgment, taking into consideration the desired plant conditions to perform the Surveillance. Operating experience has shown that these components usually pass the
- SR when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
SR 3.8.4.4 and SR 3.8.4.5 Visual inspection and resistance measurements of inter-cell, ' inter-rack, inter-tier, and terminal connections provides an indication of physical damage or abnormal deterioration that could indicate degraded battery condition. The . anti-corrosion material is used to help ensure good ; electrical connections and to reduce terminal deterioration. The visual inspection for corrosion is not intended to O. require removal of and inspection under each terminal connection. The. removal of visible corrosion is a preventive maintenance ! SR. The presence of visible corrosion does not necessarily i represent a failure of this SR, provided visible corrosion - is removed during performance of this Surveillance. , The connection resistance limits are established to maintain ' connection resistance as low as reasonably possible to minimize the overall voltage drop across the battery and the possibility of battery damage due to heating of connections. The resistance values for each battery connection are i located in the Technical Requirements Manual (Reference 9).
- The 18 month Frequency of the Surveillances is based on engineering judgment, taking into consideration the desired plant conditions to perform the Surveillance. Operating '
experience has shown that these components usually pass the SR when performed at the 18 month Frequency.. Therefore, the . Frequency was concluded to be acceptable from a reliability standpoint. i (continued) HATCH UNIT 2 B 3.8-63 REVISION A .
DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.6 REQUIREMENTS (continued) Battery charger capability requirements are based on the design capacity of the chargers (Ref. 4). According to Regulatory Guide 1.32 (Ref. 10), each battery charger supply is required to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences. The minimum required amperes and duration ensures that these requirements can be satisfied. The Frequency is acceptable, given the unit conditions required to perform the test and the other administrative controls existing to ensure adequate charger performance during these 18 month intervals. In addition, this Frequency is intended to be consistent with expected fuel cycle lengths. SR 3.8.4.7 A battery service test is a special test of the battery's capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length corresponds to the design duty cycle requirements as specified in Reference 4. The Frequency of 18 months is consistent with the recommendations of Regulatory Guide 1.32 (Ref. 10) and Regulatory Guide 1.129 (Ref. 11), which state that the battery service test should be performed during refueling operations or at some other outage, with intervals between tests not to exceed 18 months. This SR is modified by two Notes. Note 1 allows the performance of a modified performance discharge test in lieu of a service test. The modified performance discharge test is a simulated duty cycle consisting of just two rates: the 1 minute rate published for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test, both of which envelope the duty cycle of the service test. Since the ampere-hours removed by a rated I (continued) i
DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.7 (continued) REQUIREMENTS 1 minute discharge represent a very small portion of the battery capacity, the test rate can be changed to that for the performance test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test should remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test. A modified performance discharge test is a test of the , battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition. > to determining its percentage of rated capacity. Initial ; conditions for the modified performance discharge test i should be identical to those specified for a service discharge test. The reason for Note 2 is that performing the Surveillance would remove a required DC electrical power subsystem from O service, perturb the electrical distribution system, and challenge safety systems. Credit may be taken for unplanned events that satisfy the Surveillance. The I f r u (continued) HATCH UNIT 2 B3.8-6% REVISION ,A
DC Sources - Operating B 3.8.4 BASES v SURVEILLANCE SR 3.8.4.7 (continued) REQUIREMENTS swing DG DC battery is exempted from this restriction, since it is required by both units' LCO 3.8.4 and cannot be performed in the manner required by the Note without resulting in a dual unit shutdown. SR 3.8.4.8 A battery performance discharge test is a constant current capacity test to detect any change in the capacity determined by the acceptance test. Initial conditions consistent with IEEE 450 need to be met prior to the performing a battery performance discharge test. The test results reflect the overall effects of usage and age. A battery modified performance discharge test is described in the Bases for SR 3.8.4.7. Either the battery performance discharge test or the modified performance discharge test is
. acceptable for satisfying SR 3.8.4.8; however, only the modified performance discharge test may be used to satisfy SR 3.8.4.8, while satisfying the requirements of SR 3.8.4.7 at the same time.
The acceptance criteria for this Surveillance is consistent with IEEE-450 (Ref. 8) and IEEE-485 (Ref.12). These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer's rating. Although there may be ample capacity, the battery rate of deterioration is rapidly increasing. The Frequency for this test is normally 60 months. If the battery shows degradation, or if the battery has reached 85% of its expected application service life and capacity is s 100% of the manufacturers rating, the Surveillance Frequency is reduced to 12 months. However, if the battery shows no degradation but has reached 85% of its expected application service life, the Surveillance Frequency is or.ly reduced to 24 months for batteries that retain capacity 2: 100% of the manufacturer's rating. Degradation is indicated, according to IEEE-450 (Ref. 8), when the battery capacity drops by more than 10% of rated capacity from its capacity on the previous performance test or is more than 10% below the manufacturer's rating. All these Frequencies are consistent with the recommendations in IEEE-450 (Ref. 8). (continued) HATCH UNIT 2 B3.8-)0'/pD REVISION A'A
' /A
DC Sources - Operating B 3.8.4 l BASES SURVEILLANCE SR 3.8.4.8 (continued) REQUIREMENTS This SR is modified by a Note. The reason for the Note is that performing the Surve11ance would remove a required DC electrical power subsystem from service, perturb the electrical distribution system, and challenge safety systems. Credit may be taken for unplanned events that satisfy the Surveillance. The swing DG DC battery is exempted from this restriction, since it is required by both 9 1 1 l l (continued) HATCH UNIT 2 B3.8,7[/g/% REVISION
DC Sources - Operating ; B 3.8.4 BASES SURVEILLANCE SR 3.8.4.8 (continued) REQUIREMENTS units' LC0 3.8.4 and cannot be performed in the manner required by the Note without resulting in a dual unit shutdown. SR 3.8.4.9 With the exception of this Surveillance, all other Surveillances of this Specification (SR 3.8.4.1 through SR 3.8.4.8) are applied only to the Unit 2 DC sources. This Surveillance is provided to direct that the appropriate Surveillances for the required Unit 1 DC sources are governed by the Unit 1 Technical Specifications. Performance of the applicable Unit 1 Surveillances will satisfy both any Unit I requirements, as well as satisfying this Unit 2 Surveillance Requirement. The Frequency required by the applicable Unit 1 SR also governs performance of that SR for both Units. REFERENCES 1. 10 CFR 50, Appendix A, GDC 17. O
- 2. Regulatory Guide 1.6.
- 3. IEEE Standard 308 - 1971.
- 4. FSAR, Sections 8.3.2.1.1 and 8.3.2.1.2
- 5. FSAR, Chapter 6.
- 6. FSAR, Chapter 15.
- 7. Regulatory Guide 1.93, December 1974.
- 8. IEEE Standard 450 - 1987.
- 9. Technical Requirements Manual .
- 10. Regulatory Guide 1.32, February 1977.
- 11. Regulatory Guide 1.129, December 1974.
(continued) HATCH UNIT 2 8 3.8-66 REVISION A
.& a ed -e,m. m-e-b.5JA-- m-4 --A4%-~+ +ae.. -*4+-- -. m - M-* -> - -. b' --Ea. ,. g u 4. .a UNIT 2 MARKUP OF CURRENT TECHNICAL O seccivicirious ino orscussion or caincts 2 O O
DISCUSSION OF CHANGES ITS: SECTION 3.3.1.1 - RPS INSTRUMENTATION TECHNICAL CHANGE - LESS RESTRICTIVE (continued) L.7 Applicability has been modified to only require RPS functions to be operable in Mode 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies. In addition, proposed ACTION H for MODE 5 only requires action to be initiated to fully insert control rods in core cells containing one or more fuel assemblies. Control rods withdrawn from or inserted into a core cell containing no fuel assemblies have a negligible impact on the reactivity of the core and therefore are not required to be operable with the capability to scram. Provided all rods otherwise remain inserted, the RPS functions serve no purpose and are not required. In this condition the required shutdown margin (LC0 3.1.1) and the required one-rod-out interlock (LC0 3.9.2) ensure no event requiring RPS will occur. The Actions for inoperable equipment in Mode 5 are also revised to be consistent with the proposed Applicability. Since all control rods are required to be fully inserted during fuel movement (LC0 3.9.1), the proposed applicable conditions cannot be entered while moving fuel. The only possible core alteration is control rod withdrawal which is adequately addressed by the proposed action. L.8 The IRMs are added to the current exception to SR 4.0.4 (current Note d) since they are also required in Mode 2, but not in Mode 1, and the 'O required surveillance cannot be performed in Mode 1 (prior to entry in the applicable Mcde 2) without utilizing jumpers or lifted leads. Use of these devices is not recommended since minor errors in their use may significantly increase the probability of a reactor transient or event which is a precursor to a previously analyzed accident. Therefore, time is allowed to conduct the SR after entering the applicable mode. This frequency is consistent with the frequency for the APRMs which have similar function and surveillance requirements. L.9 These surveillance tests are required to be performed periodically (quarterly) while in the applicable MODES. The required periodic Frequency has been determined to be sufficient verification that the APRMs are properly functioning. Performing a reactor startup does not impact the ability of the monitors to perform their required function. Therefore, an additional surveillance required to be performed " prior to a reactor startup" is an extraneous and unnecessary performance of a surveillance. I L.10 The CHANNEL FUNCTIONAL TEST (CFT) requirement for the float type switches has been extended from quarterly to once per 18 months. This new Frequency will reduce radiation exposure to plant personnel performing this Surveillance. The NRC issued this change as Amendment 133 to the Unit 2 TS by letter dated March 15, 1994. Analysis has also been performed (GENE-770-25-1092) that shows a negligible impact on safety with i the Surveillance being performed every 18 months instead of the current 3 months. Since the CFT is part of a CHANNEL CALIBRATION (per the definition), and a CHANNEL CALIBRATION requirement is specified every 18 ' months (proposed SR 3.3.1.2.13), an actual CFT SR is not provided. HATCH UNIT 2 9 REVISION
- . - . _ . - - - . - -. . - . . ~. h DISCUSSION OF CHANGES ITS: SECTION 3.3.8.2 - RPS ELECTRIC POWER MONITORING TECHNICAL CHANGE - LESS RESTRICTIVE L.2 : (continued) ; capable of providing the necessary protection, thus, 72 hours provides ' time to repair the inoperable assembly and decreases the potential for a , unit upset (that could result when power supplies are shifted, since power is initially lost to the RPS trip system and either RPS bus powered components). The time extension for two inoperable assemblies is minimal but necessary to allow consideration of plant conditions, available personnel and the appropriate actions. L.3 A Note has been added to this Surveillance such that the Surveillance-is only required to be performed when the unit is in MODE 4 2 24 hours. Thus, the 184 day Frequency would not have to be met until a shutdown to MODE 4 2 24 hours occurs. The performance of this Surveillance could i result in hal f-scrams, actual valve isolations, and other plant perturbations, since if the assembly opens, pover is lost. The test requirement has been changed to allow it to be perhrmed while shutdown to minimize the impact of this Surveillance on plart operation. This is consistent with many of the of the more recently licensed BWRs and the BWR Standard Technical Specifications, NUREG 1433. L.4 The time delay setting for the undervoltage trip has been extended from i zero to s 4 seconds. In addition, a time delay setting has been provided for the overvoltage and underfraquency trips. The NRC issued this change as Amendment 130 to the Unit 2 T5 oy letter dated November 24, 1993. , i O i HATCH UNIT 2 2 REVISION l
DISCUSSION OF CHANGES ITS: SECTION 3.6.1.1 - PRIMARY CONTAINMENT i TECHNICAL CHANGE - MORE RESTRICTIVE M.1 (continued) This existing action would allow a startup and control rod withdrawal from cold conditions (e.g., < 212 F). Should leakages above limits be discovered while operating, the existing Action is non-specific as to the appropriates action to take. The proposed Actions provide the appropriate operational restriction, which is consistent in limitation and time to the existing LC0 3.0.3. Therefore the proposed presentation and associated Actions for containment leakage rate beyond limits and structural integrity not within limits will result in establishing and maintaining the reactor in a cold shutdown, all-rods-in, condition until the leakage or structural integrity is corrected; resulting in increased safety to the allowances of the existing Action. M.2 A Surveillance Frequency has been added. If this test fails two consecutive times, then it must be performed every 9 months (versus the current 18 months) until the test passes two consecutive times. This is an additional restriction on plant operation. TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" LA.1 Appendix J of 10 CFR 50 delineates certain requirements that must be within Technical Specifications, and others that are allowed to be detailed within -
the Bases of the Technical Specifications. The value of P,, L, and P7 are ones that Appendix J allows to be presented in the Bases. Based on the is proposed to be delineated in the Bases. allowance of the regulation, L and P are not used at Plant P, Hatch for containment tests and will not be placed in the Bases. Future changes to P, would be governed by 10 CFR 50.59 changes to the plant design basis for post-accident peak containment pressure. Refer to comment L.1 below for a description of the change to P,.
" Specific" .
L.1 The value for P, is being lowered to 48.7 psig. P, is defined in the l Technical Specifications as the peak containment internal pressure that is used for 10 CFR 50 Appendix J (leakage testing) purposes. The peak containment internal pressure, as related to 10 CFR 50 Appendix J, has > traditionally been the calculated maximum pressure following a large break, O HATCH UNIT 2 3 REVISION
l DISCUSSION OF CHANGES ITS: SECTION 3.6.1.1 - PRIMARY CONTAINMENT TECHNICAL CHANGE - LESS RESTRICTIVE L.1 (continued) design basis Loss of Coolant Accident (LOCA). For Hatch Unit 2, this break also results in the highest Final Safety Analysis Report (FSAR) analyzed accident pressures. The current Mark I containment Long-Term Program analyses regarding the containment temperature and pressure responses following a LOCA are documented in Unit 2 FSAR Section 6.2.1.4. In addition, a more recent analysis, which increased the containment normal operating pressure limit from 0.75 psig to 1.75 psig is documented in GE-NE-A00-05873-02, dated April 1994. The Hatch Unit 2 containment pressure response, due to a postulated design basis LOCA, was re-evaluated as part of the Mark I Containment Long-Term Program and is documented in NEDO-24569. The purpose of the Mark I Containment Long-Term Program was to " perform a complete reassessment of the suppression chamber (torus) design..." according to Appendix A of NUREG-0661. As a part of this complete reassessment, the Mark I Containment Long-Term Program included plant unique analyses of the containment LOCA pressure response using the Homogeneous Equilibrium Model (HEM) for vessel blowdown described in NED0-21052 and the containment n response model described in NED0-10320. These plant-unique analyses and results were provided to the NRC in Georgia Power Company's letter dated V' January 26,1983 (with later supplements) and approved by the NRC in a l Safety Evaluation Report dated January 25, 1984. These approved analyses I resulted in significantly lower inside containment peak pressures than submitted in the original FSAR. Subsequent to NRC approval, the Hatch Unit 2 FSAR was updated to reflect the new analyses and their results. Since the Georgia Power Company Mark I Containment Long-Term Program submittal, revisions have been made to certain parameters used in the model to account for the Extended Operating Domain Analyses with reduced feedwater temperature. This revision has resulted in slightly higher peak containment LOCA analyses pressures from those presented in the 1983 submittal. Through the 10 CFR 50.59 safety evaluation process, the FSAR was updated to reflect these results. The current LOCA analyses, provided I in the FSAR section referenced above, result in peak containment internal l pressures of 46.7 psig for Unit 2. As indicated in NED0-24569, the peak containment pressure calculations for a design basis LOCA assumed ac n.itial pressure of 0.75 psig. This limit corresponds to the contair u t pressure limit in current Unit 2 Specification 3.6.1.6. Also, the peak containment LOCA pressure is higher than the analyzed peak containment pressure for a Main Steam Line Break or small break LOCA inside containment. O HATCH UNIT 2 4 REVISION 1
DISCUSSION OF CHANGES ITS: SECTION 3.6.1.1 - PRIMARY CONTAINMENT TECHNICAL CHANGE - LESS RESTRICTIVE L.1 (continued) As indicated in GE-NE-A00-05873-02, the containment initial pressure was evaluated to be insreased to 1,75 psig. The evaluation addressed the following issues:
- Short-term DBA-LOCA containment pressure and temperature ,
- Long-term DBA-LOCA containment pressure and temperature
- LOCA containment hydrodynamic loads
- Safety / relief valve loads
- Appendix J containment leakage requirements
- Other issues not related to this P, change.
Based on the result of these evaluations, it was determined that the value of P,, determined by the Mark I containment Long-Term Program, should be - increased by 2 psig to 48.7 psig. Therefore, the peak containment internal pressure value of 48.7 psig for Unit 2 forms an acceptable basis for structural integrity as identified in the Proposed Bases of the Technical Specifications. This pressure is significantly less than the containment design pressure of 56 psig and the ASME Code allowable of 62 psig. i 1 e i G U REVISION A' HATCH UNIT 2 5 A l
l l DISCUSSION OF CHANGES ( ITS: SECTION 3.6.1.3 - PRIMARY CONTAINMENT ISOLATION VALVES TECHNICAL CHANGE - LESS RESTRICTIVE (continued) L.5 The phrase " actual or," in reference to the automatic isolation signal, has been added to the surveillance requirement for verifying that each PCIV actuates on an automatic isolation signal. This allows satisfactory automatic PCIV isolations for other than surveillance purposes to be used to fulfill the surveillance requirements. Operability is adequately demonstrated in either case since the PCIV itself cannot discriminate between " actual" or " simulated" signals. L.6 Comment number not used. L.7 It is proposed that the PCIV position check surveillance for manual isolation valves and blind flanges inside primary containment not be required to be performed each COLD SHUTDOWN unless the primary containment has been de-inerted. Without this exception to the normal requirement for performing this test, the primary containment would be required to be de-inerted solely to perform this test. This scenario would then also require the air lock door seal test be performed within the next 72 hours; creating unnecessary containment entries, cycling of the door seals, and man-power for testing. All these activities are generated to verify the position of valves secured in position in a very controlled area; an area which cannot be entered without major coordination and planning when i inerted (and is almost never entered when inerted). L.8 The allowable leakage limit has been increased to 100 scfh per MSIV and a combined maximum pathway leakage of :s; 250 scfh for all four main steam lines has been added. The NRC issued this change as Amendment 132 to the Unit 2 TS by letter dated March 17, 1994. L.9 An allowance is proposed for intermittently opening, under administrative control, closed primary containment isolation valves (other than the four valves discussed in A.1). The allowance is presented in proposed ACTIONS Note 1, and in Note 2 to SR 3.6.1.3.2 and SR 3.6.1.3.3. Opening of primary containment penetrations on an intermittent basis is required for performing surveillances, repairs, routine evolutions, etc. r% U 5 HATCH UNIT 2 REVISION // LA
f' CONTAINMENT SYSTEtis O ( PRIMARY CONTAINMENT INTERNAL PRESSURE SPeohka 3 6 ' 4 ~
)
LIMITING CONDITION FOR OPERATION
/\
/ \
Llo 3 ' \'O wg }_j g 1.d_ 3.6.1.6 Petmary-eente4ement internal pressure shall not-e*eee&&,75 psig. h,k APPLICABILITY: CONDITIONS 1, 2 and 3. " "' 'Rev5 ACTION. ('L1i" M,1 < V D rg J q ((With the pr_imany-cor.t44ement limit, restore the internal pressureinternaltopressure in excess within the of the Ispecified limit within hour or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN Y within the following 24 hours. V(3 l SURVEILLANCE REQUIREMENTS sR P .o1 te sve \t 4.6.1.6 The pNeery wid.ginci'ent internal pressure shall be determined to be lessthanorequalt'((.7 psig at least once per 12 hours. [,
/
- Ch L,1 i rev B j
C' h^ S Egept when pe'rforming the test hguired by Spech ation 4M.I.b or, thef Q , Q V dipecialJU: l A ized by Abendment No. 2,_hrJwhen either Jnert gr'de-ipertJqgApt @'(pufging'),prfrrjarf contsinment'as r,egst redg3.6M.
,/ j jag d
.1. 9 0.
HATCH - UNil Z v 3/4 6-9 Amendment No. 2, 29 Wr)
l
\
DISCUSSION OF CHANGES [. \ ITS: SECTION 3.6.1.4 - DRYWELL PRESSURE ADMINISTRATIVE A.1 These allowances have been deleted. The first allowance is not needed since the specific Surveillance has been deleted (refer to ITS Section 3.6.1.8 for a Discussion of Changes for the deletion of the SR). The special Startup Test allowance has been deleted since the test is completed. Therefore, these deletions are considered administrative. IECHNICAL CHANGE - MORE RESTRICTIVE M.1 The allowance to exceed the drywell pressure limit during inerting and de-inerting has been deleted. This is an additional restriction on plant operation. TECHNICAL CHANGE - LESS RESTRICTIVE L.1 The drywell pressure limit is being increased to 1.75 psig. The current Technical Specifications limit for drywell pressure is 0.75 psig, which is the initial containment pressure value assumed in the safety analysis. An evaluation, GE-NE-A00-05873-02, dated April 1994, was performed to permit A increasing the limit to 1.75 psig. This evaluation, which was reviewed () and confirmed by Georgia Power Company, addresses the following issues that are affected by the pressure increase:
- Short-term DBA-LOCA containment pressure and temperature
- Long-term DBA-LOCA containment pressure and temperature
- Net positive suction head (NPSH) for pumps taking suction from the suppression pool
- LOCA containment hydrodynamic loads
- Safety / relief valve (S/RV) loads
- Appendix J containment leakage requirements Environmental Qualification (EQ)
- Anticipated transient without scram (ATWS)
O HATCH UNIT 2 1 REVISION Q u
DISCUSSION OF CHANGES ITS: SECTION 3.6.1.4 - DRYWELL PRESSURE TECHNICAL CHANGE - LESS RESTRICTIVE L.1 (continued) A. Short-Term DBA-LOCA Containment Pre:sure and Temperature The analyses used as the basis for the FSAR short-term DBA-LOCA containment pressure and temperature used an initial drywell and wetwell pressure of 0.75 psig. Therefore, an increase in the initial (maximum operating) drywell and wetwell pressures of 1.75 psig is expected to produce an increase in the peak drywell pressure relative to the value obtained with an initial pressure of 0.75 psig. The short-term drywell peak pressure is controlled by the break flow rate, the vent flow resistance, and the vent backpressure at the time of the peak drywell pressure. The vent backpressure at the time of the peak drywell pressure is established by the sum of the wetwell pressure at the time of the peak drywell pressure and the hydrostatic head at the vent exit. The wetwell pressure is established by the amount of non-condensible gas transferred from the drywell to the wetwell during the blowdown, O by the suppression pool temperature at the time of the peak drywell pressure, and by the initial wetwell pressure. If it is conservatively assumed that all noncondensible gas in the drywell has been transferred to the wetwell by the time of the peak drywell pressure, since the drywell volume is of similar size to the wetwell airspace volume. The increase in the wetwell pressure (and thus, the increase in the vent backpressure) at the time of peak drywell pressure will also be , approximately equal to the sum of the increases in the initial 1 drywell and wetwell pressures. Consequently, the increase in l the peak drywell pressure will also be approximately equal to i the sum of the increases in the initial drywell and wetwell i pressures. Therefore, for a 1.0 psi increase in the initial 1 drywell and wetwell pressures, the estimated increase in the peak short-term drywell pressure is 2.0 psi. This estimate was confirmed by reviewing the results of a short-term analysis conducted for a similar plant with a Mark I Containment. The results of this analysis show that an increase in both the initial drywell and wetwell pressures from 0.75 to 2.0 psig results in < 1 psi increase in the peak drywell pressure. The current peak calculated drywell pressure is O HATCH UNIT 2
/JA REVISION
i DISCUSSION OF CHANGES ITS: SECTION 3.6.1.4 - DRYWELL PRESSURE TECHNICAL CHANGE - LESS RESTRICTIVE L.1 (continued) 46.7 pdg, and with a 2 psi increase, is well below the design pressure of 56 psig. (A 7 psi margin still remains.) The peak drywell temperature during the DBA-LOCA is established by the saturation temperature at the peak drywell pressure. The short-term drywell temperature response (prior to initiation of drywell spray) will be controlled by the superheated steam temperature corresponding to the drywell pressure. For a 2 psi increase, the effect would u an approximate 2 F increase in the peak temperature. The current peak calculated drywell temperature is 290 F, and with a 2 F increase, is well below the design structural temperature of 340 F. (A 48 F margin still remains.) An increase in the initial pressure to 1.75 psig will have a negligible effect on the short-term DBA-LOCA wetwell temperature, since the wetwell temperature is controlled by the suppression pool temperature. B. Long-Term DBA-LOCA Containment Pressure and Temperature FSAR Figure 6.2-25 shows that approximately 0.7 psig was used as the initial value for the drywell and wetwell pressure. The long-term DBA-LOCA drywell and wetwell pressures are approximately equal due to operation of the suppression chamber-to-drywell vacuum breakers. The vacuum breakers are designed to fully open following a DBA-LOCA when drywell pressure drops below the vacuum breaker setpoint of 0.5 psid. This occurs when the blowdown phase of the LOCA is terminated and cold ECCS water overflows from the vessel into the drywell, cooling and depressurizing the drywell. In the long-term, this effect on the drywell and wetwell pressures due to an increase in the initial drywell and wetwell pressures will be less than the effect on the short-term pressures. This is due to the redistribution of noncondensible gases between the wetwell and. the drywell after the suppression chamber-to-drywell vacuum breakers open. Therefore, long-term pressures are expected to increase by the change in the initial drywell pressure (or wetwell pressure, given that the two are equal), or approximately 1 psi, not accounting for temperature changes in the drywell and wetwell airspaces during the event. If the increase due to long-term drywell and wetwell airspaces . heatup is considered, the additional incremental effect on the O HATCH UNIT 2 [ l.b REVISION /( A _
- . . - - - . .. -- .- .. - = . - - - . . _ _ . ~. DISCUSSION OF CHANGES l ITS: SECTION 3.6.1.4 - DRYWELL PRESSURE TECHNICAL CHANGE - LESS RESTRICTIVE L.1 (continued) long-term pressures will be small (about 0.2 psi). A conservative assumption is that the long-term pressure increase is the same as the short-term pressure increase. Since the current long-term peak DBA-LOCA pressure is 14 psig, a 2 psi increase is still well below the design pressure of 56 psig. (A margin of 40 psi still remains.) In addition, the long-term DBA-LOCA peak drywell and wetwell temperatures (after initiation of drywell sprays), which are controlled by the saturation temperature corresponding to the drywell conditions, will be negligibly affected by the initial pressure increase. (Conservatively, the temperature increase will be 2 F, similar to the short-term DBA-LOCA temperature increase.) C. Net Positive Suction Head (NPSH) for Pumps Taking Suction from the Suppression Pool The parameters which affect available NPSH are suppression pool A temperature, suppression pool water level, and wetwell airspace V pressure. The effect of the increase in the initial pressure is an increase in the wetwell airspace during pump operation. This increase in the wetwell airspace pressure will result in an increase in the available NPSH. Therefore, there is no adverse impact on available NPSH of increasing the initial drywell pressure limit. D. LOCA Containment Hydrodynamic Loads The defined LOCA hydrodynamic loads are loads due to pool swell, vent thrust, condensation oscillation, and chugging. The dominant input parameters for the pool swell tests include the drywell pressurization rate to the time of vent clearing, vent flow resistance, vent submergence, and initial drywell-to-wetwell pressure difference. The drywell pressurization rate is a function of the vessel break flow, and the drywell and vent volumes. These are negligibly affected by the initial drywell pressure. The pool swell tests are based on a zero drywell-to-wetwell pressure difference. Based on sensitivity studies, higher values of initial drywell-to-wetwell pressure differences will result in lower pool swell loads. This is due to the reduction in' the vent water leg and, therefore, in the pressure required to clear the vents. The other remaining dominant parameters (vent resistance and vent submergence) are not O expected during normal operation, the pool swell design loads HATCH UNIT 2 REVISION [d_C
DISCUSSION OF CHANGES ITS: SECTION 3.6.1.4 - DRYWELL PRESSURE TECHNICAL CHANGE - LESS RESTRICTIVE L.1 (continued) will not be adversely impacted by the increase in drywell pressure. Vent thrust loads result from imbalances and flow momentum changes in the vent system. The dominant parameters which affect the vent thrust loads include vessel break flow, vent geometry (and flow resistance), and the resulting drywell-to-wetwell pressure differences and flow rates throughout the vent system. The 1 psi increase in the drywell and wetwell pressures will have a negligible effect on break flow and vent flow resistance. Since the drywell-to-wetwell pressure differences and flow rates are controlled by the vent resistance and vessel break flow, the expected effect on these parameters will also be negligible. The condensation oscillation and chugging loads are affected by steam mass and energy flux through the vents (which is a function of break flow rate and vent configuration), air content O of the vent flow, and suppression pool water temperature. 1 psi increase in initial drywell pressure will slightly The increase the air content in the vent flow. This will tend to have a mitigating effect on the steam condensation loads due to condensation oscillation and chugging. The other parameters will be negligibly affected. Therefore, the initial pressure increase will not have an adverse impact on the condensation oscillation and chugging design loads. E. Safety / Relief Valve (S/RV) Loads The S/RV loads defined include loads on the S/RV discharge line (SRVDL) piping and the hydrodynamic loads on the torus. Loads on the SRVDL piping during S/RV actuation will be mainly controlled by the S/RV opening setpoint, S/RV flow rate, SRVDL geometry, and the water leg in the SRVDL. An increase in the wetwell pressure relative to the pipe pressure prior to S/RV actuation could delay water clearing and increase pressure loads in the pipe. However, the SRVDLs are equipped with vacuum breakers in the drywell portion of the piping. An increase in the wetwell operating pressure will produce a similar increase in the SRVDL pressure prior to S/RV actuation. Therefore, the effect of an increase in the wetwell pressure on the S/RV piping O HATCH UNIT 2 f.i.D A REVISION [
i I l l , , DISCUSSION OF CHANGES t ITS: SECTION 3.6.1.4 - DRYWELL PRESSURE I TECHNICAL CHANGE - LESS RESTRICTIVE L.1 (continued) loads will be negligible since the drywell-to-wetwell pressure difference is normally greater than or equal to zero. The effect of wetwell and SRVDL pipe pressures before S/RV actuation on peak torus pressures was examined during the 1/4 scale S/RV tests. The pipe initial pressure will not be lower than the wetwell initial pressure due to the vacuum breakers on the SRVDL. The tests show that peak torus pressures are relatively insensitive to initial wetwell and pipe pressures. F. Appendix J Containment Leakage Requirements The Surveillance and leakage rate requirements are based on and used in conformance in testing withto10beCFR is proposed 48.750, Appendix psig, J. Thethe which includes value of P,2 psi increase in peak pressure due to the initial drywell pressure increase from 0.75 psig to 1.75 psig. Therefore, this increase p in initial drywell pressure will not adversely impact the Q 1eakage rate requirements since they are also being modified to account for the change. In addition, the current Technical Specifications value for P, is 57.5 psig, which is greater than the actual peak pressure, even accounting for the 2 psi increase. G. Environmental Qualification (EQ) Based on a review of the current EQ pressure envelope, the EQ pressure envelope bounds the current peak calculated pressures. Generally, margins in the pressure envelope are greater than 2 psi. However, there are regions on the envelope where the margins are less than 2 psi, such as where the pressure envelope was set equal to the small steam line break drywell pressure (near 1800 seconds) and near 20 seconds and 10' seconds where the EQ envelope equals the calculated DBA-LOCA containment pressure. Based on engineering judgment, this small increase will not adversely impact EQ of components inside containment, since the time that the EQ envelope is exceeded is very small. The effect of a 2 psi increase on short-term and long-term drywell temperature responses is estimated to be approximately 2 F, as described earlier. A review of the current EQ temperature curves shows that there are generally large margins to the temperature EQ envelope. The only exception is near 150 HATCH UNIT 2 6' j E REVISION A',
'A
DISCUSSION OF CHANGES ITS: SECTION 3.6.1.4 - DRYWELL PRESSURE TECHNICAL CHANGE - LESS RESTRICTIVE L.1 (continued) G. Environmental Qualification (EQ) (continued) seconds where the EQ envelope equals the drywell temperature calculated for the 0.5 ft' steam line break. In this instance, it has been determined by engineering judgement, that this small conservative increase will not adversely impact the EQ of components inside containment, since the time that the EQ envelope is exceeded is small. H. Anticipated Transient Without Scram (ATWS) GE document NED0-24222 provides the results of generic studies of ATWS events, including the calculated pool temperature and containment pressure response. The results of this study for a plant similar in design to Hatch Unit 2 for ATWS with a two pump standby liquid control system (similar to the Hatch Unit 2 design) and without alternate rod insertion (which Hatch Unit 2 employs) show that the peak containment pressure is 11 psig. Os With a 2 psi increase, the peak pressure is still well below the design pressure of 56 psig. (A 43 psi margin still remains.) Based on the results of the evaluations described above, an increase in the drywell initial pressure limit from 0.75 psig to 1.75 psig will not result in exceeding any design margins. In addition, while there are small time periods of the EQ envelope where an increase in the initial drywell pressure limit may result in exceeding the EQ pressure and temperature envelopes by small amounts, it has been determined that this will not adversely impact EQ requirements. O HATCH UNIT 2 /JF REVISIONf(1 2-
DISCUSSION OF CHANGES CTS: SECTION 3/4.6.1.4 - MSIV LEAKAGE CONTROL SYSTEM TECHNICAL CHANGE - LESS RESTRICTIVE
" Specific" L.1 This Specification is being deleted. The NRC issued this change as Amendment 132 to the Unit 2 TS by letter dated March 17, 1994.
O O HATCH UNIT 2 1 REVISION,A[
%iS.:hM 5.2.)
ELECTRICAL POWER SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued)
& ktc;>3)3 AW 6. Verifying the pressure in both diesel air start receiver,s) '
L_. to be > 225 psig.
- g. . e' b fd23Sig. At least once per 184 days by verifying the diesel starts from ambien e to 2764-2825 kW* for diesel generator 2A, 2360-2425 kW for diesel u^ t T*generatorIB,and 2742-285 kW* for die 2el oenerator 2C ffU ;
FW; #f <l1Z0 seconda achieves and maintains a steady-state voltage of 4160 + ' 20 volts and a steady-state frequency of 60 1.2 HZ, and operates for i M4 M, 1 60 minutes thereafter." k ' b& c. At least once per 92 days by verifying that a sample of diesel7 A M l g ,4) fuel from the fuel storage tank, obtained in accordance with l 4 w uv 3 ASTM-D270-65, is within the acceptable limits specified in )
# g3 - ,D Table 1 of ASTM D975-74 when checked for viscosity, water and j sediment.
5 i
- d. At least once per 18 months (durino shutdow by:
. Subjecting th) lesel to an inspectio n accordance th !
't procedures prepa in conjunction with manufacture 's
( commendations for s class of standby s vice.*** , . 1 l
%,h
- Momentary variations outside this band shall not invalidate the test. {
*46 **A single 6-month (184-day) test for the IB diesel will satisfy the :
4aM3 BA S requirements for Unit 1 Specification 4.9.A.2.a.2 and Unit 2 ! Specification 4.8.1.1.2.b. The 6-month test will be performed using l the starting circuitry and emergency bus for one unit. The next ! 6-month test will be performed using the starting circuitry and emergency bus from the other unit. >
. 2.
*** F o he IB diesel generato single diesel inspection h ry 18 months '
will tisfy the requirements t 1 Specification 4.9 2.a.3 and Unit 2 ification 4.8.1.1.2.d.1 . O HATCH-UNIT 2 3/4 8-3a Amendment No. 83, 119
%) i i
i i Sfeci k k O M \ ELECTRICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
)R 3.6.l.I6 2. Verifying that the automatic load sequence timer is OPERABLE with the interval between each load block within 10% of its design interval.
- 3. Verifying the diesel generator capability to reject its
;g 7. 9. l 7 laroest sinala ehntdown (emergency) load while maintaining Voita ~
1
@ ge rthi at 4160 i 4Rs volts.
2K~Keiff [for tie "Ser I Heat Removal generator 7A 7tE ce Water (RHRSW) iesel generator 1 3 pu , at rated flow; f is would be eithe he IC or 2C RHR ump at rated flow; r diesel generato C this would be ther the 2B or 2D SW oump at trated flow 7During these load rejection tests, fine dieselm cgenerator snali not exceed the nominal speea plus 75% of thei difference between nominal speed and the overspeed trip
, setpoint, or 15% above nominal speed, which ever is lower., a 46 O
,Jek 1 h bk ) b-l,)
' For the 18 diesel generator a single partial load rejection test every 18 months will satisfy the requirements of Unit 1 Specification L 4.9. A.2.a.4 and tinit 2 Specification 4.8.1.1.2.d.3.
f O HATCH - UNIT 2 3/4 8-3b Amendment No. 83, 119 6411
_ _ . ~ . ._ _ _ _ _ . _ _ . . _ _ _ _ __ DISCUSSION OF CHANGES ITS: SECTION 3.8.1 - AC SOURCES-0PERATING ADMINISTRATIVE A.1 The details relating to the required day tank level have been moved to a Surveillance Requirement (proposed SR 3.8.1.3). No technical changes are being made; therefore, this change is considered administrative in nature. A.2 AC sources are considered a support system to the Distribution System - (proposed LCO 3.8.7). In the event AC Sources are inoperable such that a distribution subsystem were inoperable, the proposed LC0 3.0.6 would allow ; taking only the AC Sources ACTIONS; taking exception to taking the AC ! Distribution System ACTIONS. Since the AC Sources ACTIONS are not sufficiently conservative in this event, specific direction to take appropriate ACTIONS for the Distribution System is added (proposed Note to ACTION E). This format and construction implements the existing treatment of this condition within the framework of Improved Technical Specification methods. A.3 The format of the proposed Technical Specifications would allow multiple Conditions to be simultaneously entered. Three or more sources could be inoperable, ACTIONS being taken in accordance with the Specification, and - LC0 3.0.3 entry conditions not met. To preserve the existing intent for LC0 3.0.3 entry, ACTION H is proposed. A.4 Proposed Notes 1, 3, 5, 6, and 8 to SR 3.8.1.2, and Note 4 to SR 3.8.1.5 O have been added. Note 1 to SR 3.8.1.2 allows SR 3.8.1.5 to satisfy SR 3.8.1.2, since it is more restrictive than SR 3.8.1.2. Note 3 to SR l 3.8.1.2 allows the engine to be warmed up and gradually started. These methods are currently employed, and have been specifically added for clarity. Note 5 to SR 3.8.1.2 allows gradual loading. Note 6 to SR 3.8.1.2 allows for voltage transients prior to establishing steady state operation. Note 8 to SR 3.8.1.2 and Note 4 to SR 3.8.1.5 allow a SR to be performed on only one DG at a time. All of these are currently being performed, and have been specifically added for clarity. All of these changes are considered administrative in nature. A.5 The existing limitation on 18-month Surveillances to perform them "during shutdown" is more specifically presented in the proposed Surveillances. Each proposed SR contains a specific Note limiting the performance in certain MODES. While these limitations vary from SR to SR, each is - l- consistent with the BWR Standard Technical Specifications, NUREG 1433 l presentation (or bracketed option allowed based on plant specific justification) which defines the intent of "during shutdown" for each SR, and with. the guidance of Generic Letter 91-04. Additionally, the Note l clearly presents the allowance of the current practice of taking credit for unplanned events, provided the necessary data is obtained. l O i HATCH UNIT 2 1 REVISION [
DISCUSSION OF CHANGES ITS: SECTION 3.8.1 - AC SOURCES-0PERATING I U ADMINISTRATIVE A.5 (continued) In addition, since the swing DG is common to both units, SRs that allow 1 one performance to satisfy both units' requirements are allowed to be l performed while one unit is not shutdown, provided the SR is being l performed from the other unit. Since this is only a change in presentation of current practice, this change is considered administrative. l A.6 These two possible values for the overspeed trip point are fixed by the design of the DG unit. The appropriate value (i.e., the most limiting, which is 65.5 Hz) is presented in the proposed Technical Specifications. This presentation eliminates the basis for the accepted value from the Technical Specifications, moving it to the Bases. Since there is no difference in the requirement, this is an editorial presentation preference only. A.7 Proposed Note 1 to SRs 3.8.1.9, 3.8.1.10, and 3.8.1.17 and Note 2 to SR 3.8.1.13 have been added. This allows an engine prelube prior to DG start. The current Specifications do not prohibit this allowance and the p addition is provided for administrative in nature. clarity. As such, it is considered v A.8 The technical content of this requirement is being moved to Chapter 5.0 of the proposed Technical Specifications in accordance with the format of the BWR Standard Technical Specifications, NUREG 1433. Any technical changes to this requirement will be addressed with the content of proposed Specification 5.6.2. A.9 The requirement to perform this Surveillance after the 24 hour run has been deleted. As indicated by the *** footnote, it is acceptable to perform the test after a 2 2 hour run at 2 2565 kW. Therefore, since it is already allowed to be performed in this manner, this change is considered administrative. A.10 The technical content of current Specifications 4.8.1.1.2.a.2, 4.8.1.1.2.a.3, 4.8.1.1.2.a.6, 4.8.1.1.2.c, and 4.8.1.1.2.d.13 is being moved to LC0 3.8.3. The technical content of Specifications 4.8.1.1.3.a.4, 4.8.1.1.3.c, and 4.8.1.1.3.d is being moved to LC0 3.8.4 and LC0 3.8.5. The technical content of Specifications 4.8.1.1.3.a.1,2,3, and 4.8.1.1.3.b is being moved to LCO 3.8.6. This is in accordance with the format of the BWR Standard Technical Specifications, NUREG 1433. Any I technical changes to these requirements are addressed with the content of the proposed LCOs.
^N (O
HATCH UNIT 2 2 REVISION B
5 m iR. A + 3.s.4 ELECTRICAL POWER SYSTEMS i SURVEILLANCE REQUIREMENTS (Continued) l l r.a i k LLo 194 g,3 _ 2. The pilot cell specific gravity, corrected to 77 F, is ) m 2 1.205, L3. The pilot cell voltage is 2 2.0 volts, and gg3,g_q.s 4. The overall battery voltage is 2 120 o l t s.. ME e its ~ ) g b. At least once per 92 days by verifying tnat:
- 1. The voltage of each connected cell is 2 2.0 volts under geg float charge and has not decreased more than 0.17 volts !
p 7,9 b from the value observed during the original acceptance test, , l
- 2. The specific gravity, corrected to 77 F, of each con-nected cell is 2 1.205 and has not decreasea more than .
0.02 from the value observed during the previous test, ' and '
- 3. The electrolyte level of each connected cell is between g
.the minimum and maximum level _ indication marks.
- c. At least once per 18 months by verifying that:
f 6(Pf.8.4 3 1. The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration,
$232'ii 2. The cell-to-cell and terminal connections are <
Q@;'t free of corrosion and coated with anti-corrosion material, and Sc38.% 3. The battery charger will supply at st 400 amperes at a minimum of 129 volts for at leas poup
- d. At least once per 18 months,@ing shutaowQverifying ')
that either: A.& g N4,q 1. >The battery capacity is adequate to supply and maintai .l in OPERABLE status 61TTFThractuai eo rgencv loa for (2 hou when the battery is subjected to a battery service test, or A. M- )* Rogosed MeL fo SR S,bA,'/ 2. TheDrtery capacity is acequate to supply a dummy load !
; of the applicable profile given in Figure 3.8.2.3-1 while 1
- maintaining the battery terminal voltage 2105 volts. .. ;
i i HATCH - UNIT 2 3/4 8-14 Amendment No. 74 2Af 1 F
-- .. . _ ~ - . - - .
I Ef eri%cu% 3.68_ ELECTRICAL POWER SYSTEMS O SURVEILLANCE REQUIREMENTS (Continued) , i 5
\ l 35 .
profowl h t the completion of either o e above tests, the z.ad ad 6 tery charger shall be demonstr ed capable of recharg-
@ ing battery at a rate of at leas 150 amperes while f**y *,o supplyi normal D.C. loads. The batte shall be charged to least 95% capacity in s 24 ho_ c At least once per 60 months'durino shutdow]&
g e. i>by verifying that t , ,gg.g the battery capacity is at least 80% of the manufacturers ,
- (. rating when subjected to a performance discharge test /This 5 fp formance scharge test \snall De performe gsubsequfnt to Jt atisfacto y completionNf the required ba\ter gservice b (test --
}
(6c modiRed Gip;> ^ NN^CE e n
.sg 3 8 A.2.
e A 3, r chsGw3cAes+. S fL 3 6 4 f > ko SR 3 9b N L.5 ! ( i O HATCH - UNIT 2 2/4 8-15 Amendment No. 74
.ko f i
p u A,. A-4L 4e - J -A .m-. uG s- -w -- a- - : J, .-a .,k.4 m -,.n a r+_ 1 SPniLheag ELECTRICAL POWER SYSTEMS O SURVEILLANCE REQUIREMENTS (Cor.tinued) l 4.8.1.1.3 Each diesel generator battery and battery charger shall be demonstrated OPERABLE: j
,g,q.) a. At least once per 7 days by verifying that:
F1 . The electrolyte level of each pilot cell is between the I hebismsrsd\ minimum and maximum level indication marks, ' Ck rs &~ ' Iwi 8% i" 2. The pilot cell specific gravity, corrected to 77 F, is M5@- 2 1.205,
- 3. The pilot cell voltage is 2 2.0 volts, and _
56W 4. The overall battery voltage is zhltgy ggb !
- b. At least once per 92 days by verifying that:
- 1. The voltage of each connected cell is 2 2.0 volts under float charge and has not decreased more than 0.17 volts from the value observed during the original acceptance test, 3 2. The specific gravity, corrected to 77 F, of each connected cell is 2 1.205 and has not decreased more than 0.02 ;
j from the value observed during the previous test, and
- 3. The electrolyte level of each connected cell is between C the minimum and maximum level indication marIs. l t
- c. At least once per 18 months by verifying that:
g ,g,q.3 1. The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration, ; 5g,g,4q 2. The cell-to-celled terminal connections are ffeTQ g free of corrosion and coated with anti-corrosion _ material, and Mwg]e. ' gg, g,4, g 3. The battery charger will supply at least 1 amperes it ,6 a minimum of 129 volts for at least hou
- d. At least once per 60 month 36urTno shutdoRortry eritying that l g O 'b' N the battery capacity is at least 80% of the man facturers rating when subjected to a performance discharge test _fThis pe ormance 1 h ;
gpo>ed 'd1 barge test snal _ De periormea subseg nt to the satisf tory l 3 m g). comp tion of the req ired battery service est. _- t Y ?t)f -
& SfL g,% Proposch SR-so.84.s HATCH - UNIT 2 PM /4 8-6 y 3.g.9.g;a
.i %.3 5~tW
DISCUSSION OF CHANGES ITS: SECTION 3.8.4 - DC SOURCES-0PERATING /~'T U IECHNICAL CHANGE - LESS RESTRICTIVE (continued) L.4 This requirement has been deleted since there is no reason to perform two discharge tests, one right after the other. One test should be sufficient to demonstrate battery 0PERABILITY, The service test is a test which ensures the battery will perform as required in the accident analysis. The performance discharge test is a design test of the battery. Since the service test demonstrates proper OPERABILITY, there is no reason to require a second subsequent test. In addition, a Note has been added to proposed SR 3.8.4.7 to allow the modified performance discharge test to be l performed in lieu of the service test of SR 3.8.4.7. As stated in the BWR Standard Technical Specifications Bases, NUREG 1433, (proposed by NUREG change package NRC-15) this substitution is acceptable, because SR 3.8.4.8 represents a more severe test of battery capacity than SR 3.8.4.7. L.5 An allowance to perform a modified performance discharge test in lieu of a performance discharge test has been added to this Surveillance. The modified performance discharge test is a simulated duty cycle consisting of just two rates: 1 one minute rate published for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test. Since the ampere-hours removed by a rated 1 minute discharge represent a very small portion of the battery capacity, the test rate can be changed to that for the performance test without y compromising the results of the performance discharge test. O HATCH UNIT 2 4 REVISION A
DISCUSSION OF CHANGES ITS: SECTION 3.8.6 - BATTERY CELL PARAMETERS TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" LA.1 These requirements basically measure degradation of the given cell.
Degradation does not necessarily mean that the entire battery is inoperable. " Degradation" is proposed to suffice for the Technical Specifications requirement, while allowing the details of the definition of " degradation" to be relocated to the plant procedures. Degradation will now affect the Frequency of a battery performance discharge test i (proposed SR 3.8.4.8). The original acceptance test values were specified to be within 2.20 to 2.25 volts per cell (VPC). The 0.17 volt decrease specified in the current specification corresponds to 2.03 to 2.08 volts per cell. Therefore, it could be construed that the structure of the new ! specification covers the value and frequency of the old specification (0.17 from 2.25, and 92 days for the Category B limits), as well as the relocation of the specific cell deterioration from degradation value. t
" Specific" L.1 A 31 day Completion Time for restoring battery cell parameters has been O provided (Required Action A.3). This Completion Time is considered acceptable since sufficient battery capacity exists to perform the intended function and to allow time to fully restore battery cell -
parameters to normal limits. This change is consistent with IEEE Battery Working Group (BWG) recommendations in a letter from B. M. Radimer (IEEE BWG) to S. K. Aggarwal (NRC) dated August 2,1988. To support this new time, two additional requirements have been added. Required Action A.1-has been provided to verify pilot cell electrolyte level and float voltage 1 are within allowable values (Category C limits) within I hour when Category A or B parameters are not within limits. This change provides a quick indication of the status of the remainder of the battery cells. Required Action A.2 has been provided to verify battery cell parameters for all the cells are within Category C limits within 24 hours when Category A or B parameters are not within limits. These Category C values are the limits at which the battery would be considered immediately ; inoperable. This change provides assurance the battery is still capable i of performing its intended function. If Category C limits are not met, or ! the Category A and B limits are not restored within 31 days, proposed ' ACTION B requires the affected battery to be declared inoperable (and the appropriate ACTIONS of proposed LCOs 3.8.4 or 3.8.5 taken) In addition, a Note has been added to the ACTIONS to provide more explicit ; instructions for proper application of the Actions for Technical , Specifications compliance. In conjunction with the proposed Specification l 1.3, " Completion Times," the Note (" Separate Condition entry is allowed l O for each ....") and "one or more" provides direction consistent with the intent of the proposed Action. I HATCH UNIT 2 2 REVISION ,A
.-r ,, -
DISCUSSION OF CHANGES ITS: SECTION 3.8.6 - BATTERY CELL PARAMETERS (O _/ TECHNICAL CHANGE - LESS RESTRICTIVE
" Specific" (continued)
L.2 This allowance is acceptable based on guidance from Appendix A to IEEE-450. The level excursion allowed is temporary due to gas generation during the equalizing charge and would be expected to return to normal. O V O HATCH UNIT 2 /dik REVISION [
m UNIT 2 NO SIGNIFICANT IIAZARDS DETERMINATION O O l O
NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.3.1.1 - RPS INSTRUMENTATION L.10 CHANGE The No Significant Hazards Determination evaluation is provided in Georgia Power Company letter from J.T. Beckham, Jr. to the NRC, dated September 20, 1993. Subsequently, the NRC issued this change as Amendment 133 to the Unit 2 TS by letter dated March 15, 1994. O O HATCH UNIT 2 11 REVISIONA[ c I
NO SIGNIFICANT HAZARDS DETERMINATION
;9 ITS: SECTION 3.3 8.2 - RPS ELECTRIC POWER MONITORING l l
L.4 CHANGE j The No Significant Hazards Determination evaluation is provided in Georgia Power Company letter from J.T. Beckham, Jr. to the NRC, dated October 19, 1993. Subsequently, the NRC issued this change as Amendment 130 to the Unit 2 TS by letter dated November 24, 1993. O l O O HATCH UNIT 2 4 REVISION AN
'M i
t NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.1.1 - PRIMARY CONTAINMENT L.1 CHANGE (continued)
- 3. Does this change involve a significant reduction in a margin of safety?
The proposed change does not involve a significant reduction in the margin of safety because leakage testing and structural limits will continue to be met based on the peak containment pressure resulting from a design basis accident. The peak containment internal pressure of 48.7 psig l continues to be within the containment internal maximum allowable pressure > of 62 psig. There is no requirement for the test pressure to be higher than the peak accident pressure. The proposed change to P, will not change 3 the accident analyses and resultant radiological consequences for a postulated LOCA. The radiological consequences continue to be within the requirements of 10 CFR 100. The use of the revised P w leakage rate is measured and calculated appropriately.ill ensure that the O O - l HATCH UNIT 2 2 REVISIONf(h
NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.1.3 - PRIMARY CONTAINMENT ISOLATION VALVES L 8 CHANGE The allowed MSIV leakage is being revised from 11.5 to 100 scfh per valve and a combined maximum pathway leakage of 250 scfh for all four main steam lines is being added. The No Significant Hazards Determination for this change is provided in GPC letter dated January 6, 1994, and February 3, 1994, from J. T. Beckham, Jr. to the NRC. Subsequently, the NRC issued this change as Amendment 132 to the Unit 2 TS by letter dated March 17, 1994. U O HATCH UNIT 2 8 REVISION /I.'I_
. -. . - = - . _.
N0 SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.1.4 - DRYWELL PRESSURE L.1 CHANGE In accordance with the criteria set forth in 10 CFR 50.92, Georgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated in changing the initial drywell pressure because the primary contain'ent is , designed to accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from a LOCA. This meets the requirement of 10 CFR 50, Appendix A, Criterion 50 for the containment to retain its integrity during a design basis accident. Satisfactory leak rate testing at the value of the peak calculated containment pressure following a LOCA provides the assurance that any release of radioactive materials will be restricted to the provisions of 10 CFR 100 as provided in the safety analyses. The probability or consequences of an accident are not significantly increased, because there is no change to the containment design basis or the ability of the containment to perform the required function of O- preventing the release of radioactivity to the environment.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? '
The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated since the design features of the primary containment, as required by Criterion 16 of 10 CFR 50, Appendix A, are not altered. Testing at the calculated peak design basis LOCA pressure demonstrates that the primary containment and associated systems provide an acceptable barrier against the uncontrolled release of radioactivity to the environment. No new failure mode is introduced by changing the initial containment pressure, since the - assurance of integrity at the calculated accident pressure is maintained by testing at the appropriate value.
- 3. Does this change invoNe a significant reduction in a margin of safety? '
The proposed amendment does not involve a significant reduction in the margin of safety, because leakage testing and structural limits will continue to be met based on the peak containment pressure resulting from a design basis accident. The peak contaiment internal pressure of 51.6 psig continues to be within the containment internal maxinum allowable O
NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.1.4 - DRYWELL PRESSURE L.1 CHANGE (continued) pressure of 62 psig. The proposed change to the initial containment pressure does not change the resultant radiological consequences for a postulated LOCA. The radiological consequences continue to be within the requirements of 10 CFR 100. O O HATCH UNIT 2 2 fb REVISION,A' ^ A
NO SIGNIFICANT HAZARDS DETERMINATION CTS: SECTION 3/4.6.1.4 - MSIV LEAKAGE CONTROL SYSTEM L.1 CHANGE This specification is being deleted. The No Significant Hazards Determination for this change is provided in GPC letter dated January 6,1994, and February 3, 1994. The NRC issued this change as Amendment 132 to the Unit 2 TS by letter dated March 17, 1994. O l l0 HATCH UNIT 2 1 REVISION [gf).
NO SIGNIFICANT HA?ARDS DETERMINATION ITS: SECTION 3.8.4 - DC SOURCES - OPERATING L.4 CHANGE In accordance with the criteria set forth in 10 CFR 50.92, Georgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The DC electrical power sources are ussd to support mitigation of the consequences of an accident; however, they are not considered the initiator of any previously analyzed accident. As such, the performance of a modified performance discharge test in lieu of a service discharge l test will not increase the probability of any accident previously evaluated. The proposed SR continues to provide adequate assurance of OPERABLE batteries since a modified performance discharga test represents l a more severe test of battery capacity than a service discharge test. Therefore, the proposed change does not involve an increase in the consequences of any accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
O (,/ The proposed change does not introduce a new rmde of plant operation and does not involve physical modification to the punt. Therefore, it does not create the possibility of a new or different knd of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction 'n a margin of safety?
This change does not involve a significant reduction in a margin of safety since the proposed substitution of the modified performance discharge l test, in lieu of a service discharge test, continues to provide adequate : indication that the battery is capable of performing its design function. ' I I
\
HATCH UNIT 2 4 REVISION
1 1 NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.8.4 - DC SOURCES - OPERATING L,5 CHANGE > In accordance with the criteria set forth in 10 CFR 50.92, Georgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? ,
The DC electrical power sources are used to support mitigation of the t consequences of an accident; however, they are not considered the , initiator of any previously analyzed accident. As such, the performance of a modified performance discharge test in lieu of a service discharge test will not increase the probability of any accident previously evaluated. The proposed SR continues to provide adequate assurance of OPERABLE batteries, since the modified performance discharge test represents a more severe test of battery capacity than does a service discharge test. Therefore, the proposed change does not involve _ an increase in the consequences of any accident previously evaluated. l
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not introduce a new mode of plant operation and , does not involve physical modification to the plant. Therefore, the ' possibility of a new or different kind of accident from any accident previously evaluated is not created.
- 3. Does this change involve a significant reduction in a margin of safety?
This change does not involve a significant reduction in a margin of J safety, since the proposed substitution of the modified performance discharge test for the service discharge test continues to provide adequate indication that the battery is capable of performing its design function. i i 1 i O , HATCH UNIT 2 5' 4A REVISION A i
Q NUREG 1433 COMPARISON DOCUMENT - SPECIFICATIONS O l 4 O < I
.. . . . ~ - - _. ~.. . _ . . . - - . .. . .-
Drywell Pressure 3.6.1.4 3.6 CONTAINMENT SYSTEMS 3.6.1.4 Drywell Pressure LCO 3.6.1.4 Drywell pressure shall s S psig . g MODES 1, 2, and 3. APPLICABILITY:
~
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell pressure not- A.1 Restore drywell I hour within limit. pressure to within i limit. B. Required Action and B.1 Be in MODE 3. 12 hours . i associated Completion Time not met. AND ; B.2 , Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.4.1 Verify drywell pressure is within limit. 12 hours i O BWR/4 STS 3.6-19 Rev. O, 09/28/92 1
)
l
Drywell Air Temperature 3.6.1.5 3.6 CONTAINMENT SYSTEMS 3.6.1.5 Drywell Air Temperature L( ' LC0 3.6.1.5 Drywell average air temperature shall be s :{1351*F. APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell average air A.1 Restore drywell 8 hours temperature not within average air limit. temperature to within limit.
- 3. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.5.1 Verify drywell average air temperature is 24 hours within limit.
O BWR/4 STS 3.6-20 Rev. O, 09/28/92
AC Sources-Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.3.1 Detennine OPERABLE 4fhours DG(s) are not inoperable due to comon cause failure. E B.3.2 Perform SR 3.8.1.2/ h [24[ hours forOPERABLEDG(s) B Restore [ red G 7 o &A 9 to OPERABLE status. w- u.:4 z.
%
- L.
7 do1s g
$' +" I.daysfrom discovery of 5" #3 fai1 h A$ 10 LCO 3 A ). a, b , o t) g..
%Q,f EQ g%(f. Twofrequired}%ffsiteG 7 Dec-@ lare required 12 hours from D reuits inoperable. feature (s)j inoperable Dw en the redundant discovery Conditior [(yb TP em or muc [ required feature (s) are inoperable. concurrent ith inoperability of redundant k no 8 5' required O(> ( (pau cu a;I'bi t feature (s) AND gg Restore ne 24 hours
- p. Mrequired}4ffsite
[( ' circuit to OPERABLE status. (continued) O BWR/4 STS 3.8-3 Rev. O, 09/28/92
1 l AC Sources-Operating 3.8.1 1 __. ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME j m I6 h
/ OnefTequired}kffsite ----------
1 h
-NOTE-------------
circuit inoperable. Enter app'icable Conditions P2- and Requi ed Actions of l AND LCO 3.8. " Distribution 09/ Systems-Operating," when One Mrequired}kG Condition is entered with inoperable. no AC powe source to one [.11"1".2'.. . C ............. di V M , ' Restoreftequiredf Et.f b u , ( E /, offsite circuit to )12 hours OP3 /, OPERABLE status. OR
'/ g
,. s Restore (requiredl DG 12 hours
. E to OPERABLE status.
~
\
h c 0' fP g/ T wof{w-%t quiredhDGs 6E} V E.1 Restoreitne
+4uiredROGsto 2 hours /
Ofo\ inoperable.\ '^ OPERABLE s tatus.CF.2 h
- o, ,4 (continued)
^
ks L u.4 2
. i s c.
P'2 bl ?_ t, c O BWR/4 STS 3.8-4 Rev. O, 09/28/92
INSERT Notes 3.8.1.2
- 4. For the swing G, a Jingle test will satisfy this Surveillance both units, using the starting __.- u. i i cuitry of nit and synch ~r6iiTied W4160 V bus y or one perio test, and the starting *"
# cuitry of F nd s~ynchironited to 4160 V tius
~
~
u" ' durigthe(nex
.,e u ,s it test.
w t
- 5. DG 1 ings may incl gradual loading as recommended by the manufacturer.
Slav-\ (rY
- 6. t=rf transients above the upper voltage limit
='-- *- lued: r; do not invalidate this test.
- 7. Momentary transients outside the load range do not invalidate this test. !
- 8. This Surveillance shall be conducted on only one DG at a time.
i l O V HATCH UNIT 1(O M 2.
l l l i O INSERT Notes 3.8.1.5 i g ( 2. DG loadings may include gradual loading as recommended by the manufacturer. - -
/
- 3. hne n t e ry-t rarrsie n tt-a bo v e -t he-u ppe r-volt a g e-Ti mi t 1 pr-ier-to-loading-do-not--invalidate this-dest-
[f.3 Momentary load transients outside the load range
. do not invalidate this test.
@ f. This Surveillance shall be conducted on only one DG at a time. ,,.,, fl l
J f.6 For the swing G, a single test will satisfy this lB l Surveillance for both units, using the starting cuitry of @ and_ s_Ynchronized to 4160 V bus
"'^'
h("bgu.n 7 during th next periodic test.
*i M for one per cuitry of nit test and the starting and synchronized to 4160 V bus D ", ' '
O Q @4t ca.- L.s
~
i i . t i .O . HATCH UNIT 1 ( ttaiIr2.
AC Sources-Operating 3.8.1 O b SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY 7_ _ SR 3.8. ------..----------NOTES-------------------- g-^- 1. 7his4urveillance shall not DE 1 o
,g f gerformetM .
y NM < 2. Credit may-b taken for ned {3,01.{j p.e d hat satisfy this SR. Verif 6+= Q18 months each DG-operat4r.g rejects {et e ;=-- a loadg 978}dandill -
*M" a. Following load rejection, the h A grf0 del S yu O (~ frequencyiss}65.5 Hz;GrftO g C Ok -
- b. WithinN3]Mcondsfollowingload yi
, 9 dS]g {&r6ge1' rejection, the voltage is e 3740Q
&nd - 6 p
[ {Cfd - Gm{df vi, _ ands}4580
, u ithid6] seconds re;iectio , Jre u foll ia is & ; f[4 6
h t V@8.8]-Hf'in - Hz. SR 3.8.1 ;l{f - --..------------.. NOTES---.--------------.-
- 2. Wis Surveillance shall not '
~ ~ - perf d in MODE 1 or .
Inser+/ 'i-
- 2. Credit m ta or unplanned g[l.6
, o.
ev that satisfy t ___........ ......................... __. /
,v Verif each DG operating at a power factor Q18 months}o oes not trip and voltage is s
f/k. _O. 5 intained sk4800}(V'during and following n I* a load re 4-[2000]k". jection of a [1710] kgand-( (continued) BWR/4 STS 3.8-9 Rev. O, 09/28/92
AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8. --.-.-.--.--------. NOTES-------------------
- 1. All DG starts may be preceded by an engine prelube period.
- 2. This Surveillance shall not b performed in MODE 1, 2, or 3 4-9 g J hI edit may be taken for unplanned events that satisfy this SR.
Verify on an actual or simulated loss of f[18 months offsite power signal:
- a. De-energization of emergency buses;
- b. Load shedding from emergency buses; and DG auto-starts from standby condition c.
and: g
- 1. energizes pennanently connected loadsins/12} Seconds, [
- 2. energizes auto-connected shutdown y s loadsthrouh)[automaticload PI sequencey , ,
maintains steady state voltage / I 3. aj{37401-V and 5 ^V, 41 3
}(
- 4. maintains steady sta e requency /
ag58.B}%z and sp61.2}'-Hz, and !!
- 5. supplies permanently connected and [
auto-connectedshutdownloadsforj ef5}^1ninutes. l} (continued) O 3.8-10 Rev. O, 09/28/92 BWR/4 STS
t AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
~
SURVEILLANCE FREQUENCY is SR 3.8. . g ------------------NOTES-------------------
- 1. Momentary transients outside the load ,
and power factor ranges do not invalidate this test. , 3.~ _- . NA
- f. Q perfemed is Surveillance in, MODE'1 or sha
- 2. W IIo't be wt th' ~
(3. Credit may be taken etun events that satisfy this% planned
, ).......... ...______.....................__
Verify each DG operating at a power factor f[18 months}^-
==
j{0d}' operates for a 24 hours 1(
- a. Fo }' r loaded a 100fkWand
- b. For the remaining hours of the test loaded a {2850f-kW and 5 50} kW. l[
.O. ED 2q1 P9 it SR
. 8. .JF
---.----------- ---NOTES-------------------
- 1. This Surveillance shall be performed
- within 5 minutes of shutting down the DG after the DG has operated Il my[23' tours loaded a 401 kW.tntF--
Opg W5 ,2000]-h g W ; h CMomentary transients outside of load P/ range do not invalidate this test.
- 2. All DG starts may be preceded by an engine prelube period.
F{18 months]F Verify sf _ each DG starts and achieves, in12T' seconds, voltage '
- L lA58A;#andfrequencyaf58.8Ni i P l *'
s 61.2 Hz" Gnd o ffer sfcad y 5 +of,
- cuJ.%, '
or- rea t heel m e .c+m , e ve if oc e .3 37 4o y ,.s g 4 ::.1 ?, \/ cr c/ _ ./ (continued) m %, u s~u
~# + c ' \sa y 6on.h sr.t, % g
. ' "3 :. .
O h1 (
- r. .
g q 4c e
%yc.6.1 s.
4 .,.~ , -
. . . N.. , c d4.v-uj 0 L2 Wp 3' O, 09/28/92 3.8-13 Rev.
8WR/4 STS l
AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY g d '\ SR ------------------NOTES------------------- 3.8. Q G 1. This Surveillance shall not be - f9 perfonned in MODE 1, 2, or 3Q' g7 [ be taken for unplanned events that satisfy this SR. Verify each DG: f[18 months} *
- a. Synchronizes with offsite power source while loaded with emergency loads upon a simulated restoration of offsite power;
- b. Transfers loads to offsite power source; and
- c. Returns to ready-to-load operation.
e\ v SR 3.8.1 17,1 - - - - - - - - . - - - - - - - - - N O T E S '- - - - - - - - - - - - - - - - - - -
' JP This Surveillance shall not n MODE 1, 2, or 3 /
pe arr gg 2I(freditmaybetakenforunplanned events that satisfy this SR. Verify with a DG operating in test mode and >[18 months} M , connected to its bus, an actual or s v simulated ECCS initiation signal overrides p the test mode by:
- a. i Returning operation;and}" {DG to ready-to-load l Pl b. Automatically energizing the
- l l
f- - emergency load from offsite power. (continued) O 3.8-14 Rev. O, 09/28/92 BWR/4 STS 1 l
1 DC Sources-Operating 3.8.4 : ( SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is 7 days e on float charge. SR 3.8.4.2 Verify no visible corrosion at teminals 92 days and connectors. b p3 0_g a7) u ~ .4L Verify connection resistance dis
-s K 5E ohm, for inter- ii connectTUns',s
,1. -4 ohm, for inter rack connect 1 s,- k 5 .sE-4 ohm for int -tier connect s, an s [1.5E-4' ohm] fo terminal e nnections].f
~
SR 3.8.4.3 hD Veri y cells, cell plates, and bette.,z
&l P) t12]'mont s
.' racks show no visual indication of physical damage or abnormal deterioration. SR 3.8.4.4 hp 3] SM Remove visible corrosionk and Verify cell 16
.12Pmonths Or/
to cell and terminal connections are c4estf-and tight, : # coated with' anti-corrosion material. SR 3.8.4.5 (>P3 QQ ifyN}onnectionresistancedis' '" ^
'JE' Amon fI s
~ II.5 4 ohm ~ for int -cell connectio A
s ' 1. -4 ohm' for in r-rack connectio s, s 'l E-4 ohm' for i ter-tier connect' ns, k and s [1.5E-4' ohm] (co nections]y or terminal k (continued) O 3.8-25 Rev. O, 09/28/92 BWR/4 STS
DC Sources-Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY d
-NOTES---- j----------- '
SR 3.8.4.6 [C-------p-------lanceshal
- 1. ThfsSurveil not be rformed in MODE 1, , or 3. 8 27.
- 2. Credit may be taken for unplanned events that satisf this SR. j
.,......................................). .
Verify each required battery charger di8 months f supplies af400 amps for station service p) subsystems, and e 100 amps for DG f subsystemsy atpa 4129PV for n ,{ M our f - I SR 3.8.4.7 -------------------NOTES-------------------
- 1. SR 3.8.4.8 may be perfomed in lieu L{fY g SR 3.8.4.7.bnce per -50 ine.dhi.e P4 2 )
h
,g-1 .;. p . / 2. This Surveillance shall not b c
2-a,r. derformeLina4nnE 4.m w... w E.,o 1. 2sr#p-~ h, JhPed'iCmay"bhYak~eHT6r"3. kl,m>a)p (( 3.6A16
~
On~pTidife~d I events that satisfy this SR.
........................................... e,
( La Verify battery capacity is adequate to 418 months f supply, and maintain in OPERABLE status, the required emergency loads for the design duty cycle when subjected to a battery service test. (continued) i O 3.8-26 Rev. O, 09/28/92 BWR/4 STS l
l l O INSERT SR 3.8.4.7A The modified performance discharge test in INSERT SR 3.8.4.78 the service test in n. O l-
% um v m=z '
DC Sources-Operating l 3.8.4 ! SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY v SR 3.8.4.8 ---- --- --- ------- - - NOT E;5'
# This Surveillance shall not arformeddoM00EMA3fx
- be_ [Z 4, % s- , 3 w we.,. %,. u c, y f may en or unp inn ~ed 64 b events that satisfy this SR.
Verify battery capacity is 2:/[80[o he 60 months manufacturer's rating when subjected to a performance discharge test AND y
~ >
finsc<T '. w F'---NOTp---\ l ' '" \ Only a licable\ 0 (3.6.4.6 when attery j 2 g.eN S j
- I'$"l radation or
'- J as reached :
[85]% of - d0 h $I.$$.$_. .$._ l mon s f 3 e 5 n ifA 't 1 - L_ I O l 3.8-27 Rev. O, 09/28/92 BWR/4 STS
DC Sources-Shutdown 3.8.5
~
3.8 ELECTRICAL POWER SYSTEMS 3.B.5 DC Sources-Shutdown e A LCO 3.8.5 i)Celec ical power subdystems shall be O ERABLE to su ort the DC electrical powpf distribution sub ystem(s) re ire by L 3.8.10. " Distribution Systems- utdown." C L ,(^D dd APPLICABILITY: b MODES 4 and 5, During movement of irradiated fuel assemblies in the Jsecondary}tcontainment. N*i t, . . t ACTIONS .- CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare affected Imediately DC electrical power required feature (s) subsystems inoperable. inoperable. " E A.2.1 Suspend CORE Imediately ALTERATIONS. AND A.2.2 Suspend movement of Imediately irradiated fuel s assemblies in the r% f{secondaryK-() / ,
- containment.
u.3 1 AND o.t (Continued) O 3.8 28 Rev. O, 09/28/92 BWR/4 STS
O INSERT SR 3.8.4.8 or a modified performance discharge test r)\ O l O :
;;s m ez a uur a 1
t Inverters-Shutdo 3.S{8 i
/ ACTIONS / ~
l \{0NDITION REQUIRED ACTION COMPLEIf0NTIME I I \ / A. (continued)N A.2.4 Initiate action to Immediately {
\ restore [ required] / 1
\ inverters to OPERABLE status.
/
N j
/
\ /
SURVEILLANCE REQUIREMENTS \ / FREQUENCY SURVEILLANCE \ j ,
'\ / l SR 3.8.8.1 Verify correct inverter voltage 7 days .
[ frequency,] and alignments /tos [ required] ' AC vital buses. , =O \ , .
/ N s
1 / \ ' I e i
! I 1 \ I i \
i / \ l I
! \,
I \ i
. \.
1 'k i
/
,i
,/
\ i ,
/ '/
/ - . .
3.8-37 Rev. O, 09/28/92 BWR/4 STS 1
Distribution Systems-Operati
- 3. .
3.8 CTRICAL POWER SYSTEMS @ 3.8 Distribution Systems-Operating 1 -7 ~' LCO . .F f[DivisioK 1] and [Div)fion 2] AC, DC,f{and AC vital bps] gv w plectpital power di,stribution subsyAems shall be OMRABLE)
~ / j
. t.
APPLICABILITY: MODES 1, 2, and 3. D 1 t A/B ta ?cr Y' ACTI a 6 N COMPLETION TIME j& \ CONDITION REQUIRED ACTION fc h ,. - h
$j O&
,Q/ One' TAC electR &l A) Restore AC electrical 8 hours j
' power aistribution b power distribution subsystem, inoperable. subsystems to AND \
OPq % OPERABLE status. g 't .. , m[, G P- L5) 16 hours from g ' discovery of B.OneACvitalbus B.1 Restore AC vital bus J. hour inoperable. ~~---_ distributiV,
"' xsubsystems to AND fj Of7 y -OPERABLLs,tatus. t
- 16 hours from 1 NRC 2 ~ -discovery of 6 fat 1ure.to '
meet LCO bN gA 5} Onedstation service Restore,0C electrical power distrioution 2 hours (~IT61ectrical power ; distribution subsystem subsystems to AND ( l inoperable. OPERABLE status.
~w , , 2 5 y 16 hours from ;
/b. A \i discovery of
4 ' '
su e.-
; -- . , UA L . Ce m faiIty e m.1neet
\. ',., L:4t im LCO e m. m
\/ kL- Q p o ((q g (Continued) b^ r c b incorpco re )
BWR/4 STS [ -N.Cer 3 r0JI e / reU 3.8-38 Rev. O, 09/28/92 e t c 4, li lo !! e f (4,. 4, i
O INSERT A/B 3.8.7 J***
,,i t u.s 2.
A. orborerequired I" u.n nit FAC or DC A.1 Re 7 days ei rical power %stora_tequired3U d'DC subsystems inoperable. subsystem (s) to OPERABLE status. ye B. One i g B.1 Restore DG DC 12 hours l electrical power bu JG DC electrical power distribution distribution AND subsystems inoperable. subsystem to OPERABLE status. 16 hours from discovery of failure to meet LC0 3.8.7.a l O l O HATCHUNIT1hm'rz.
NUREG 1433 COMPARISON DOCUMENT - BASES O O
Control Rod Block Instrumentation B 3.3.2.1 A U B 3.3 INSTRUMENTATION B 3.3.2.1 Control Rod Block Instrumentation BASES BACKGROUND Control rods provide the primary means for control of reactivity changes. Control rod block instrumentation includes channel sensors, logic circuitry, switches, and relays that are designed to ensure that specified fuel design limits are not exceeded for postulated transients and accidents. During high power operation, the rod block monitor (RBM) pruvides protection for control rod withdrawal error events. During low power operations, control rod blocks from the rod worth minimizer (RWM) enforce specific control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA). During shutdown conditions, control rod blocks from the Reactor Mode Switch-Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent criticalities. The purpose of the RBM is to limit control rod withdrawal if ,q localized neutron flux exceeds a predetemined setpoint Q during control rod manipulations. It is assumed to function to block further control rod withdrawal to preclude a MCPR Safety Limit (SL) violation. The RBM supplies a trip signal to the Reactor Manual Control System (RMCS) to appropriately inhibit control rod withdrawal during power operation above the low power range setpoint. The RBM has two channels, either of which can initiate a control rod block when the channel output exceeds the control rod block setpoint. One RBM channel inputs into one RMCS rod block circuit and the other RBM channel inputs into the second RMCS rod block circuit. The RBM channel signal is generated by averaging a set of local power range monitor (LPRM) signals at various core heights surrounding the control rod being withorawn. A signal from one average power range monitor (APRM) ch anel assigned to each Reactor Protection System (RPS) trip sy.; tem s ap ies a reference signal for the RBM channel in the same tnp system. This reference signal is used to detemine which RBM range setpoint (low, intermediate, or high) is enabled. If the APRM is indicating less than the low power range setpoint, the RBM is automatically bypassed. The RBM is also automatically bypassed if a peripheral control rod g is selected -
- p. % )
(Ref.1). Dn+ M] A N~ (continued) BWR/4 STS B 3.3-43 Rev. O, 09/2B/92
Control Rod Block Instrumentation B 3.3.2.1 BASES BACKGROUND The purpose of the RWM is to control rod patterns during (continued) artu , such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods inserted to 10% RTP. The sequences i effectively limit the potential amount and rate of reactivity increase during a CRDA. Prescribed control rod Nbgn sequences are stored in the RWM, which will initiate control rod withdrawal and insert blocks when the actual sequence pM\ p deviates beyond allowances from the stored sequence. The RWM detemines the actual sequence based position indication for each control rod. The RWM also uses feedwater flow and steam flow signals to detemine when the reactor wer i above the preset power level at which the R8N is WM automatically bypassed (Ref. 2). The RWM is a single channel system that provides input into both RMCS rod bloc 3 circuits. With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This Function prevents inadvertent criticality as the result of a control rod withdrawal during MODE 3 or 4, or during MODE 5 when the reactor mode switch is required to be i:1 the shutdown position. The reactor mode switch has two channels, each inputting into a separate RMCS rod block circuit. A rod block in either RMCS circuit will provide a control rod block to all control rods. APPLICABLE 1. Rod Block Monitor SAFETY ANALYSES, LCO, and The RBM is designed to prevent violation of the MCPR APPLICABILITY SL and the cladding 1% plastic strain fuel design limit that may result from a single control rod withdrawal error (RWE) event. The analytical methods and assumptions used in evaluating the RWE event are summarized in Reference 3. A statistical analysis of RWE events was perfomed to detemine the RBM response for both channels for each event. From these responses, the fuel thermal perfomance as a function of RBM Allowable Value was detemined. The Allowable Values are chosen as a function of power level. Based on the specified Allowable Values, operating limits are established. (continued) BWR/4 STS B 3.3-44 Rev. O, 09/28/92
INSERT AA for Backaround Section. B 3.3.2.1
. A rod block signal is also generated if an RBM Downscale trip or an Inoperable trip occurs, since this could indicate a problem with the RBM channel. The Downscale trip will occur if the RBM channel signal decreases below the Downscale trip setpoint after the RBM signal has been normalized. The Inoperable trip will occur during the nulling (normalization) sequence, if:
the RBM channel fails to null, too few LPRM inputs are available, a module is not plugged in, or the function switch is moved to any position other than
" Operate." The Bypass Time Delay ensures that the normalized signal is passed to the trip logic within the appropriate time. The delay is between the time the signal to the reference is nulled and the signal is passed to the trip logic.
A O I O neum4 tom n
1 l 1 Control Rod Block Instrumentation l B 3.3.2.1 ph BASES APPLICABLE 1. Rod Block Monitor (continued) SAFETY ANALYSES, LCO, and The RBM Functien satisfies riterion 3 of the NRC Policy APPLICABILITY Statement g Two channels of the RBM are required to be OPERABLE, with
;p a [ their setpoints within the appropriate Allowable Valuesfer "--
d l-the associat+d-pan cengr, to ensure that no single instrument failure can preclude a rod block from this C h Function. The actusi setpoints are calibrated consistent
/ with applicable setpoint methodology.p g Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable.
Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor
,m power), and when the measured output value of the process
( ) parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the and process, analytic somelimits, of the corrected instrumentfor calibration,ThetripLg,,' errors. ' setpoints are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because ) instrumentation uncertainties, process effects, calibratio tolerances, instrument drift, and severe environment ,__ (for channels that must function in harsh environments as fo, L defined by 10 CFR 50.49) are accounted for. U The RBM is assumed to mitigate the consequences of an RWE event when operating a 29% RTP. Below this power level, the consequences of an RWE event will not exceed the MCPR SL and, therefore, the RBM is not required to be OPERABLE : (Ref.3). When operating'< 90% RTP, analyses (Ref. 3) have shown that with an initial MCPR E 1.70, no RWE event will I result in exceeding the MCPR SL. Also, the analyses demonstrate that when operating at a 90% RTP with MCPR e 1.40, no RWE event will result in exceeding the MCPR 3 (V (continued) O, 09/28/92 I BWR/4 STS B 3.3-45 Rev. i
Control Rod Block Enstrumentation B 3.3.2.1 BASES APPLICABLE 1. Rod Block Monitor (continued) SAFETY ANALYSES, LCO, and SL (Ref. 3). Therefore, under these conditions, the RBM is APPLICABILITY also not required to be OPERABLE.
- 2. Rod Worth Minimizer p.7b The RWM enforces the banked position withdrawal sequence (BPWS) to ensure that the initial conditions of the CRDA analysis are not violated. The analytical methods and l A- dMoR. M + . assumptions used in evaluating the CRDA are sumarized in Ochr a d e. (,,4,3 References 4, d, e, and 7.4 The BPWS requires that control i ( W er,< % rods be moved in groups, with all control rods assigned to a
. , . . ( . . . ', q h specific group required to be within specified banked positions.
Requirements that the control rod sequence is in
'gM, h /'U compliance with the BPWS are specified in LCO 3.1.6, " Rod V * '[ Q Pattern Control."
The RWH Functio isfies Criterion 3 of the NRC Policy r Statemen . y q ) Since the RWM is a N -d system designed to act as a backup to operator control of the rod sequences, only one channel of the RWM is available and required to be OPERABLE (Ref. 7). Special circumstances provided for in the Required Action of LCO 3.1.3, " Control Rod UPERABILITY," and LCO 3.1.6 may necessitate bypassing the RWM to allow continued operation with inoperable control rods, or to allow correction of a control rod pattern not in compliance with the BPWS. The RWM may be bypassed as required by these conditions, but then it must be considered inoperable and the Required Actions of this LCO followed. Compliance with the BPWS, and therefore OPERABILITY of the RWM, is required in MODE" 1 and 2 when THERMAL POWER is
< 10% RTP. When THERMAL POWER is > 10% RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal /gm fuel damage limit during a CRDA (Refs. 5 and 7). In MODES 3 and 4, all control rods are required to be inserted into the core; therefore, a CRDA cannot occur. In MODE 5, since only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.
(continued) BWR/4 STS B 3.3-46 Rev. O, 09/28/92
Primary Containment B 3.6.1.1 [ B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment BASES BACKGROUND The function of the primary containment is to isolate and contain fission products released from the Reactor Primary System following a Design Basis Accident (DBA) and to confine the postulated release of radioactive material. The i primary containment consists of a steel lined, reinforced l concrete vessel, which surrounds the Reactor Primary System I and provides an essentially leak tight barrier against an uncontrolled release of radioactive material to the environment. The isolation devices for the penetrations in the primary containment boundary are a part of the containment leak , tight barrier. To maintain this leak tight barrier: I f
- a. All penetrations required to be closed during accident conditions are either: ,
1
/^x 1. capable of being closed by an OPERABLE automatic l V y Containmentv Isolation system, or i
_ 2. closed by manual valves, blind flanges, or
'- de-activated automatic valves secured in their
'IM "
closed positions, except as provided in D , LCO 3.6.1.3, " Primary Containment Isolation Valves (PCIVs)";
~~}
j b. The primary containment air lock is OPERABLE, except J , J' as provided in LCO 3.6.1.2, " Primary Containment Air
,g c .e i 6 3 3 Lock"; Q ~~~ ~ ~
H' 3 ' ' c. 3 YpHisorfred-sealinhmechanism-associat~ed%Tth penetration,ls CPERABE excii@t'as-provi11ed in a Z /.. gprart _
). ~~~_
This Specification ensures that the perfonnance of the primary containment, in the event of a DBA, _ meets the assumptions used in the safety analyses of References 1 j and 2. SR 3.6.1.1.1 leakage rate requirements are in ' confonnance with 10 CFR 50, Appendix J (Ref. 3), as modified by approved exemptions. L (continued) BWR/4 STS B 3.6-1 Rev. O, 09/28/92
Primary Containment B 3.6.1.1 BASES (continued) APPLICABLE The safety design basis for the primary containment is that SAFETY ANALYSES it must withstand the pressures and temperatures of the l limiting DBA without exceeding the design leakage rate. l The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage. Analytical methods and assumptions involving the primary , containment are presented in References 1 and 2. The safety I analyses assume a nonsechanistic fission product release i following a DBA, which forms the basis for determination of offsite doses. The fission product release is, in turn, based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assume in the safety analyses is not exceeded. g The maximum allowable 1 ag3 rate for the primary containment (L,) ism 1.2 y weight of the containment air
~ per. 24 hours at the maximum peak containment pressure (P,) z
,_)/ )
- MI Onl ofM57.5Mg er [0.84M by weight cf th: cc::t-ain;;at air ,-- {
b :r 24 hours st-the-reduced--pressttre-of " ({284]-psit (1 ' k U 2 o M D >(Ref. 1). g Primary containment satisfies Criterion 3 of the NRC Policy Statemen { , LCO Primary containment OPERABILITY is maintained by limiting leakage topithia the acceptance criteri ef 10 CT" 50, - N Appcadix J (",;f. 3)c' Compliance with this LCO will ensure a FJ'N nA primary containment configuration, including equipment c
~
s hatches, that is structurally sound and that will limit
- ieakage to those leakage rates assumed in the safety (G P.1) v analyses.
Individual leakage rates specified for the primary containment air lock are addressed in LCO 3.6.1.2. (continued) BWR/4 STS B 3.6-2 Rev. O, 09/28/92
i Primary Containment Air Lock B 3.6.1.2 O O BASES i BACKGROUND containment leakage rate to within limits in the event of a I (continued) DBA. Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in the unit safety analysis. ART 6T1'.TTTalage;' rate ' Tequiremdhts conforw(th 10'CFR-50 4 Appihdix J (Ret.has modified,by' approved exe D ions. j - Q.3 APPLICABLE The DBA that postulates the maximum release of radioactive SAFETY ANALYSES material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment is {,4) designed with a maximum allowable leakage rate (L,) of 1.2% by weight of the containment air per 24 hours at the e calculated maximum peak containment pressure (P ) of
- ulo y L5% SE fr psig (Ref. J) . This allowable leakage rate forms the l basis for the ac'ceptance criteria imposed on the SRs
* @ O'd[ g*7 associated with the a' lock.
Primary containment - lock OPERABILITY is also required to (^ minimize the amount of fission product gases that may escape primary containment through the air lock and contaminate and pressurize the secondary containment. The primary containment air lock satisfies Criterion 3 of theNRCPolicyStatemen(.Od* b f,36 i LC0 As part of primary containment, the air lock's safety function is related to control of containment leakage rates following a DBA. Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event. The primary containment air lock is required to be OPERABLE. i For the air lock to be considered OPERABLE, the air lock i interlock mechanism must be OPERABLE, the air lock must be i in compliance with the Type B air lock leakage test, and I both air lock doors must be OPERABLE. The interlock allows only one air lock door to be opened at a time. This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be (continued) BWR/4 STS B 3.6-7 Rev. O, 09/28/92 1
=
Primary Containment Air Lock I B 3.6.1.2 BASES LCO OPERABLE. Closure of a single door in each air lock is (continued) sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for nonnal entry and exit from primary containment. APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are - reduced due to the pressure and temperature limitations of these MODES. Therefore, the primary containment air lock is not required to be OPERABLE in MODES 4 and 5 to prevent leakage of radioactive material from primary containiaent. ACTIONS The ACTIONS are modified by Note 1, which allows entry and exit to perform repairs of the affected air lock component. If the outer door is inoperable, then it may be easily accessed to repair. If the inner door is the one that is inoperable, however, then a short time exists when the containment boundary is not intact (during access through g the outer door). . The, abi44ty to open the OPERABLE door, V fpr~ even if it means the primary containment boundary is temporarily not intact, is acceptable due to the low Q' probability of an event that could pressurize the primary f containment during the short time in which the OPERABLE door is expected to be open. The OPERABLE door must be immediately closed after each entry and exit. The ACTIONS are modified by a second Note, ich ensures
/
/ . k,c% appropriate remedial measures are taken; n-necessary Pursuant to LCO 3.0.6, actions are not required, even ~,f 1
'I C #P' i h primary containment is exceeding its leakage limit.
\oYY 4,n Therefore, the Note is added to require ACTIONS for g et cf ; ,dr.'#',J LCO 3.6.1.1, " Primary Containment," to be taken in this event.
nf v
\pDY (.p3 g el p ic A.I. A.2. and A.3 c f ' W, ,
\ With one primary containment air lock door inoperable, the OPERABLE door must be verified closed (Required Action A.1) in the air lock. This ensures that a leak tight primary (continued)
BWR/4 STS B 3.6-8 Rev. O, 09/28/92 i _ _ _ _ _ _ _ _ _ _ -
l Drywell Pressure B 3.6.1.4 ' B 3.6 CONTAINMENT SYSTEMS B 3.6.1.4 Drywell Pressure BASES BACKGROUND. The drywell pressure is limited during normal operations to preserve the initial conditions assumed in the accident analysis for a Design Basis Accident (DBA) or loss of coolant accident (LOCA). APPLICABLE Primary containment performance is evaluated for the entire SAFETY ANALYSES spectrum of break sizes for postulated LOCAs (Ref.1). Among the inputs to the DBA is the initial primary
. Analyses assume an
['1._j containment initial drywell internal pressurepressure of48.75 (Ref.psig 1) P This limitation ensures that the safety analysis remains valid by maintaining the expected initial conditions and ensures that kl J the peak LOCA drywell internal pressure does not exceed the maximumallowableoff62( The maximum calculated drywell pressure occurs during the O reactor blowdown phase of the DBA, which assumes an instantaneous recirculation line break. The calculated peak r57. Q psig l drywell (Ref.1). pressure fory1 this limiting event is/I t g 2. g Drywell pressure satisfies Criterion 2 of the NRC Policy y Statemen . "% (rec 2) ?h ~ LCO , n the event of a DBA, with an initial drywell pressure (). f). 3-s .75 psigWthe resultant peak drywell accident pressure w 1 be maintained below the drywell design pressure, g APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining drywell pressure within limits is not required in MODE 4 or 5. ! (continued) BWR/4 STS B 3.6-33 Rev. O,09/28/92 l r
Drywell Pressure B 3.6.1.4 EASES (continued) ACTIONS A.1 With drywell pressure not within the limit of the LCO, drywell pressure must be restored within 1 hour. The Required Action is necessary to return operation to within the bounds of the primary containment analysis. The I hour Completion Time is consistent with the ACTIONS of LCO 3.6.1.1, " Primary Containment," which requires that primary containment be restored to OPERABLE status within 1 hour. B.1 and B.2 , If drywell pressure cannot be restored to within limit l within the required Completion Time, the plant must be l brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full l power conditions in an orderly manner and without challenging plant systems. O SURVEILLANCE SR 3.6.1.4,1 REQUIREMENTS Verifying that drywell pressure is within limit ensures that l unit operation remains within the limit assumed in the primary containment analysis. The 12 hour Frequency of this SR was developed, based on operating experience related to trending of drywell pressure variations during the applicable MODES. Furthermore, the 12 hour Frequency is considered adequate in view of other indications available I in the control room, including alams, to alert the operator l to an abnormal drywell pressure condition. i FSAR, Section >[6.2 8 6 2 oh d *4' REFERENCES [f )() ,1. w as a ly
/ W.
O BWR/4 STS B 3.6-34 Rev. O, 09/28/92
- . - . ~ . . . - - - - _ - . . - - . .- - - - .. - - _ _ - . - _ - . - . MUW 2 Ok Primary Containment Hydrogen Recombiners B 3.6.3.1 BASES ACTIONS M (continued) to prevent exceeding this limit, and the low probability of failure of the OPERABLE primary containment hydrogen recombiner. Required Action A.1 has been modified by a Note indicating that the provisions of LCO 3.0.4 are not applicable. As a result, a MODE change is allowed when one recombiner is inoperable. This allowance is provided because of the low probability of the occurrence of a LOCA that would generate hydrogen in amounts capable of exceeding the flamability limit, the low probability of the failure of the OPERABLE subsystem, and the amount of time available after a postulated LOCA for operator action to prevent exceeding the flamability limit. B.1 and B.2 ( C 5fie ~ 's te: This Condj.tlon.de-only~aTroweilor units ( with exna e ytfr6geLe,,trel-system acceptabbt_to, the < _y staff. ; With two primary containment hydrogen recombiners inoperable, the ability to perfom the hydrogen control hy e function via alternate capabilities must be verified by n1 administrative means within I hour. The alternate hydrogen T control capabilities are provided by the iPrimary Containment)Jne ntag System .. .... .. .,... .. .... gth'4trof
.je r Co..tei ;.at " r phere Oihti= f.y:t::, . The 1 hour Completion Time allows a reasonable period of time to verify
@e r'l"> $ that a loss of hydrogen control function does not exist.
.ir ;r's Mete. Tii ivt h iiiv is te be used-if-a non ,
p p;hnicaFSpecificat11m Tec liiternate-hydrogen-tontrot-foncWn, 1,- 4s used, to dustify this-tondition. In addition, the alternate hydrogen control system capability must be S bA.Ms _ -- ' verified-every 12 hours thereafter to ensure its continued
' availability.t {Both} the (initial} verification fand all bl '
' $ subsequent verifications} may be performed as an 00 " administrative check by examining logs or other information '
to determine the availability of the alternate hydrogen control system. It does not mean to perform the t i Surveillances needed to demonstrate OPERABILITY of the I alternate hydrogen control system. If the ability to perform the hydrogen control function is maintained, (continued) BWR/4 STS B 3.6-81 Rev. O, 09/28/92 utJ iT 2 33)
g)'lcJh Primary Containment Hydrsgen Re omb fY BASES g ACTIONS B.1 and B.2 (continued) continued operation is permitted with two hydrogen recombiners inoperable for up to 7 days. Seven days is a reasonable time to allow two hydrogen recombiners to be inoperable because the hydrogen control function is maintained and because of the low probability of the occurrence of a LOCA that would generate hydrogen in amounts capable of exceeding the flammability limit. C.1 If any Required Action and associated Completion Time cannot be met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.3.1.1 , REQUIREMENTS l Performance of a system functional test for each primary containment hydrogen recombiner ensures that the recombiners are OPERABLE and can attain and sustain the temperature necessary for hydrogen recombination. In particular, this i SR verifies that the minimum heater sheath temperature increases to a (1200}*F in s 1.Si hours and that it is r maintained > {1150}?F and < ((1300k'F for a {4}; hours [l ' ; < thereafter to check the ability of the recombiner to function properly (and to make sure that significant heater elements are not burned out). Operating experience has shown that these components usually pass the Surveillance when performed at the {18} month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. SR 3.6.3.1.2 This SR ensures there are no physical problems that could affect recombiner operation. Since the recombiners are (continued) g BWR/4 STS B 3.6-82 Rev. O, 09/28/92 utJiT2odj
t CAD System M OWN I d B 3.6.3.E BASES 9 l ACTIONS B.1 and B.2 (continued) \ D the alternate hydrogen control system capability must be veriff&overy 12 hours thereaf ter to ensure its continued k
'f[ l w w@ cc%/~ availability.h 18othl the -{ initial) verification fand all
[] subsequent verifications}: may be perfomed as an administrative check by examining logs or other information to determine the availability of the alternate hydrogen control system. It does not mean to perform the Surveillances needed to demonstrate OPERABILITY of the alternate hydrogen control system. If the ability to perform the hydrogen control function is maintained, continued operation is permitted with two CAD subsystems inoperable for up to 7 days. Seven days is a reasonable time to allow two CAD subsystems to be inoperable because the hydrogen control function is maintained and because of the low probability of the occurrence of a LOCA that would generate hydrogen in amounts capable of exceeding the flamability limit. With two CAD sub:ystems inoperable, one CAD subsystem must be restored to OPERABLE status within 7 days. The 7 day Completion Time is based on the low probability of the (s) occurrence of a LOCA that would generate hydrogen in the amounts capable of exceeding the flamability limit, the amount of time available after the event for operator action to prevent exceeding this limit, and the availability of other hydrogen mitigating systems. L.1 If any Required Action cannot be met within the associated - Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. (continued) BWR/4 STS B 3.6-95 Rev. O, 09/28/92 2n O \ :ra l _ ____ _ _ _ _ - _ _ _N
I /N 'h 3 GT9 d gy MT BASES (continued) [ e SURVEILLANCE REQUIREMENTS SR 3.6. . . gg h g~ GM - that there is e ] galNf liquid nitrogen Verifying supply in % CEyncm will ensure at least (7} days of post-LOCA CAD operation. This minimum volume of liquid 9 6 Sub nitrogen allows sufficient time after an accident to replenish the nitrogen supply for long tem inerting. This Q is verified every 31 days to ensure that defsystem is capable of perfoming its intended function when required. The 31 day Frequency is based on operating experience, which has shown 31 days to be an acceptable period to verify the liquid nitrogen supply and on the availability of other hydrogen mitigating systems.
-~
hf SR 3.6.3.W.2 Verifying the correct alignment for manual, power operated, and automatic valves in each of the CAD subsystem flow paths provides assurance that the proper flow paths exist for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable because the CAD System is manually initiated. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. The 31 day Frequency is appropriate because the valves are operated under procedural control, improper valve position would only affect a single subsystem, the probability of an event requiring initiation of the system is low, and the system is a manually initiated system. REFERENCES 1. Regulatory Guide 1.7, Revision b1}. C n-khN
- 2. FSAR, Section (M*
(Q-V^ s b" BWR/4 STS B 3.6-96 Rev. O, 09/28/92 Jta i T l o rd
RHRSW System B 3.7.1 B 3.7 PLANT SYSTEMS k ') t B 3.7.1 Residual Heat Removal Service Water (RHRSW) System 1 BASES , 1 1 BACKGROUND The RHRSW System is designed to provide cooling water for the Residual Heat Removal (RHR) System heat exchangers, required for a safe reactor shutdown following a Design Basis Accident (DBA) or transient. The RHRSW System is operated whenever the RHR heat exchangers are required to operate in the shutdown cooling mode or in the suppression pool cooling or spray mode of the RHR System. The RHRSW System consists of two independent and redundant subsystems. Each subsystem is made up of a header, two
>[4000} gpm pumps, a suction source, valves, piping, heat Opt exchanger, and associated instrumentation. Either of the two subsystems is capable of providing the required cooling capacity with,ee pumpsoperating to maintain safe shutdown fg' conditions. The two subsystems are separated from each other by normally closed motor operated cross tie valves, so that failure of one subsystem will not affect the OPERABILITY of the other subsystem. The RHRSW System is designed with sufficient redundancy so that no single active pd -3 component failure can prevent it from achieving its design
}.jb'j function.. The RHRSW System is described in the FSAR, Section[{9.2.735) Reference 1.
(AnM1N - .. . dAmr 2.) Cooling water is pumped by the RHRSW pumps from the , p{Altamaha Riverf through the tube side of the RHR heat d g' exchangers, and discharges to the (circulating water fiume[ - A minimum flow line from the pump discharge to the intake structure prevents the pump from overheating when pumping against a closed discharge valve. gy (g The system is initiated manually from the control \ room. If operating during a loss of coolant accident (LOCA)l, the system is automatically tripped to allow the diesel generators to automatically power only that equipment necessary. terefleed the me The system can be manually _c+2,-teri anu +4mn in-4-'" "'- '^N ~ - started any time the LdCA sign $i'is 7nanu51iy overridden or ( clears. A - - - - j The.Wu be Nut l3.sbW ag b awm
& LP d M isceu;nf, S
i l (continued) BWR/4 STS B 3.7-1 Rev. O, 09/28/92 i
RHRSM System B 3.7.1 BASES (continued) APPLICABLE The RHRSW System removes heat from the suppression pool to O SAFETY ANALYSES limit the suppression pool temperature and primary . containment pressure following a LOCA. This ensures that fN6 the primary containment can perform its function of limiting the release of radioactive materials to the environment ,. *Qwr 2)., following a LOCA. The ability of the RHRSW Sy, stem,to..M. 4ner s>. *. . support long term cooling of the reactor cr y f ,y,y, .,. containmentisdiscussedintheFSAR,(hapter and!_15 ~~**..a (Refs. 2 and 3, respectively). These/ analyse ici: y assume that the RHRSW System will provide acequate cooling Am support to the equipment required for safe shutdown. These analyses include the evaluation of the long tem primary h"8IO containment response after a design basis LOCA. The safety analyses for long term cooling were performed for
.m various combinations of RHR System failures. The worst case
[.W ' single failure that would affect the perfomance of the RHRSW System is any failure that would disable one subsystem 3, ,(t4n n 2'y . ., of the SW System. As discussed in the FSAR,
,' 'Se'ct U 6.2.1.4.3A (Ref. 4)) for these analyses, manual 3 N.31T initi'ation of the UFtKAbLL KHRSW subsystem and the j associated RHR System is assumed to occur 10}f minutes after
,...h6 8T 1> %;fQ a DBA. The RHRSW flow assumed in the ana yses is 4000)gpm
/ , m,3., g per pump with two pumps operating in one loop. I this t . w.gw ~
case, the max < mum sunnrettion chamber water temperature and pressure arey206.4k'F and 436.59Npsig, respectively, well s-AN % / balow the design temperature oT p.8 0k'F and maximum gl qv6.4 L. f allowablepressureoff62hpsig. _g_ g
~
The RHRSW System satisfies Criterion 3 of the NRC Polic Statemen . .
&nti2) g d- @ . . 4 u n ir 15 v
LCO Two RHRSW subsystems are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming the worst case single active failure occurs coincident with the loss of offsite power. An RHRSW subsystem is considered OPERABLE when:
- a. Two pumps are OPERABLE; and (continued)
BWR/4 STS B 3.7-2 Rev. O, 09/28/92 l l
AC Sources-Operating B 3.8.1 BASES SURVEILLANCE .SR 3,8.1.2 an SR ,8r (continued) REQUIREMENTS (see Note 3)of-SR-3dv1-EDwhen a modified start procedure Ogg, as described above is used. If a mo4ified start % not h-g used, the 12 second start f requiremedf SR 3.8. .Fa' 5' b
% st c p eq
- c. , g t , .. , ,, Since SR 3.8.1 deFrequires a 12 second start, it i are h restrictive than SR 3.8.1.2, and it may be performed in lieu of SR 3.8.1.2. This procedure is the intent of Note 1,e h,%q{ ,5R 18,44 P26 ss, n .
The normal 31 day frequency for SR 3.8.1.2 (see L.<a Table 3.8.1-1, " Diesel Generator Test Schedule") is y p_f f.yU * ' (corisistent wTtFR@atory 6uidPIJ-(Ref,-1)";- The-184-d5y Frequency. fat _SR 3.8cir7-is-a-reduction-in-col 6-testh N 5 L[g]h -conshtent-wirth-Generic Frequencksprovid Lcti.er es-n (Hei. ?)t--Thes6 h equate assurance of DG OPERABILITY, g>fy
- l. lo t I while minimi ing de radation resulting from testing.
v'z - d'_
"~T3_3,g14 ;-3L i.'
~
9
- g[N 'M: Eurvei.Llanee verifies that the DGs are capable of ((ff) ,
/ f il,,,,. ,,.%)..M, F) equivalent-of-the-maximsm s
synchronizing and accepting 1 greater-than-or-equal-to-thE expected eccident-4eade. A 6A < m D u ' minimum run time of 60 minutes is required to stabilize g ,,,. , . 4. d., w engine temperatures, while minimizing the time that the DG A s.a is connected to the offsite source.
.{ 3$ m,,w2 k-
, h . . .J e , a Although no power factor requirements are established by !
c, a. w, ~ * ' this SR, the DG is normally operated at a '
. o,eu betweeno[0.8 lagging 3Snd[1.0]?The>{0.8bowerfacto.alue f is the / h
\
design rating of the machine, whileI{1.0 bis an operational : limitation,[to ensure cinuhti9 currents are =inisind}#--I J fl l' Jhe load-band-i+-provided +n avoid routine overinading-oR j the DG,---Rout 4ne everleeding may result in more frequeht / Of N l-teardown-inspections in ww dance-with- =dort, rece....mndation [ Iha-novinal 31 day Ficquency for-thit-Surve441ence-(s'ee-Tabla ~4 8.1-1-)-46-conshtent- wi th Regulatory-Guide -l .9.. . [ N (Ref J.)....- (mo f L. l fo % .B'1 ('3 C in %
)- (continued) 1 BWR/4 STS B 3.8-17 Rev. O,09/28/02 l
AC Sources-Operating hv Ck bWCdtk5She CMMtUn.' h S j kh n' q -{l f 5 r I o < 4 i 0} -l((t Eic(d6 db\IC NC-C * 'I I BASES ((A y \/MLQC_liW t+ c1D_JLrt_jpgjd(d;iic hi
. Ol SURVEILLANCE [$R 3,8.1.3 (con & ~ ~N REQUIREMENTS [Fk/. - i U ' Not .odifies this Surveillance to indicate that diesel engine runs for this Surveillance may include gradual ,
! loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimiz g {%;
j NoteModifies this Surveillance by stating that momenth (h , e UL
' g-6 F
transients because of changing bus loads do not invalidate ; this test. SimMarly, sementery pour fedu. ;.rentienTspi / s-(L n}gt-) I above--the'-Mmit 9 de-act irvalidate-the-test.J}'n a De conduct Te OP9.Not on on y one DG at a time in order to avoid common cause i failures that might result from offsite circuit or grid ! perturbations. l Note-4-stipulates-a-prarequisite-requhement for-perfomafice\ ; of this-.SR_ A successful DG start-must.-precede _this_ test 4c bf- credit satisfactory-perfomance. I SR 3.8.1. h This SR provides verification that th L level of fuel oil in the day tank fend-enghe-mounted tar 4 is at or above the M level at which fuel oil is automatically added. The level
~
h,',o,y.t "w ,,,m.2 is expressed as an equivalent volume in gallons, and is selected to ensure adequate fuel oil for a minimum of I hour I of DG operation at full load _plus 10%./
/ v. L < - +
( q ,) .,, _m a' 'o 7 .- . . )The N 31 day Frequency is adequate to ensure that a sufficient va , , .h g m-- g T supply of fuel oil is available, since low level alarms are ((- e \,-. s provided and fecH4tfoperators would be aware of any large
. , ;, f.. nL4' uses of fuel oil during this period.
( ,
% q SR 3.8.1 p Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water Removal of water from the environment fuel oil day {end in nginc ordermounted to survive.I~ Tanks once every 430:' days , - ,
eliminates the!necessary environment for bacterial survival. +., 3
@ (continued)
O BWR/4 STS B 3.8-18 Rev. O, 09/28/92 l _ _ _ _ _ _ _ _ _ _
INSERT SR 3.8.1.5 (continued) To minimize the wear on moving parts that do not get lubricated when the engine is not running, this SR has been modified by a Note (Note 1) to indicate that all DG starts for this Surveillance may be preceded by an engine prelube , period and followed by a warmup prior to loading. ! Note 2 modifies this Surveillance to indicate that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized. Note 3 modifies this Surveillance by stating that momentary voltage or load transients because of changing bus loads do / not invalidate this test. ; Note 4 indicates that this Surveillance is required to be conducted on only one DG at a time in order to avoid common sl cause failures that might result from offsite circuit or i grid perturbations. A fA t O, To minimize testing of the swing DG, Note 5 allows a single -l test (instead of two tests, one for each unit) to satisfy the requirements for both units, with the DG started using the starting circuitry of one unit and synchronized to the : ESF bus of that unit for c e periodic test and started using : the starting circuitry of the other unit and synchronized to the ESF bus of that unit during the next periodic test. This is allowed since the main purpose of the Surveillance, to ensure DG OPERABILITY, is still being verified on the proper frequency, and each unit's starting circuitry and breaker control circuitry, which is only being tested every ' second test (due to the staggering of the tests), historically have a very low failure rate. If the wing DG fails one of these Surveillances, the DG should be considered inoperable on both units, unless the cause of the failure can be directly related to only one unit. l L_ ._.
INSERT SR 3.8.1.6 (lb N MAOP i l This Surveillance tests the applicable logic associated with the Unit I swing bus. The comparable test specified in the Unit 2 Technical Specifications tests the applicable logic associated with the Unit 2 swing bus. Consequently, a test must be performed within the specified Frequency for each unit. The Note specifying the restriction for not performing the test while the unit is in MODE 1 or 2 does not have applicability to Unit 2. As the Surveillance represents separate tests, the Unit 1 Surveillance should not be performed with Unit 1 in MODE 1 or 2 and the Unit 2 test should not be performed with Unit 2 in MODE 1 or 2. i INSERT SR 3.8.1.7 (RL .d D f,U. f The largest single load for DGs IA and 1C is a core spray pump at rated flow (1275 bhp). For DG 18, the largest single load is a residual heat removal service water pump at rated flow (1225 bhp). This surveillance may be ' accomplished by either a) tripping the DG output breaker with the DG carrying greater than or equal to the largest i single load while paralleled to offsite power or while solely supplying the bus, or b) tripping the largest single load with the DG solely supplying the bus. Although Plant Hatch Unit 1 is not committed to IEEE-387-1984 (Ref.12), this SR is consistent with the IEEE-387-1984 requirement that states A U HATCH UNIT 1 4 UJy1' L B 3.8 I l l
INSERT SR 3.8.1.6 { ONd 1 Mf,CV,, _ ) This Surveillance tests the applicable logic associated with the Unit 2 swing bus. The comparable test specified in the Unit 1 Technical Specifications tests the applicable logic associated with the Unit I swing bus. Consequently, a test must be performed within the specified Frequency for each unit. The Note specifying the restriction for not performing the test while the unit is in MODE 1 or 2 does not have applicability to Unit 1. As the Surveillance represents separate tests, the Unit 2 Surveillance should not be performed with Unit 2 in MODE 1 or 2 and the Unit 1 test should not be performed with Unit 1 in MODE 1 or 2. INSERT SR 3.8.1 1 ( OAd 1 WQfJ f' The largest single load for each DG is a residual heat
\ removal service water pump at rated flow (1225 bhp). This
- Surveillance may be accomplished by either a.) tripping the DG output breaker with the DG carrying greater than or equal to the largest single load while paralleled to offsite power or while solely supplying the bus, or b.) tripping the ,
largest single load with the DG solely supplying the bus. Although Plant Hatch Unit 2 is not committed to IEEE-387-1984 (Ref. 11), this SR is consistent with the IEEE-387-1984 requirement that states Ch G l HATCH UNIT 1 & UNIT 2 B 3.8
]
- . - . .. - ~_
AC Sources-Operating B 3.8.1 BASES 1 SURVEILLANCE SR 3.8.1. (continued) REQUIREMENTS g cand 3 , this represents 65.5 Hz, equivalent to 75% of the difference between nominal speed and the overspeed trip setpoint. [TiI ' e, voltage, and frequency tolerances sp.ecMf1Rf,in this SR ar ed from Regulatory G W (Ref. 3) recommendations fo?'tesp ed ad sequence intervals.
%{ The [6] seconds specif 60% of the 10 second /
hyv load sequence intanET associated wit
' residual removal (RHR) pumps during an undervo
~
the nj t s concurrent with a LOCA.) T e vol ge and N sp " ange k, frequency-- the
'{de;ecifiert rir:t are p consistent
=d by thewith DG. the SR 3.8.1. corresponds to p,gua oc themaximumfyquency g.g,c me +3.B_1.9 6' state voltage Gadwhile c ag m=gexcursion, he;ern:y SR v;isee "
3.8.1.AM 6 4 -11 c z W D ito which the ty__... ust recover following load rejection.
', fed-oM3 'J0Q @The.A8 month)4requency ecommendation of Regulatory Guide is 1.108 consistent (Ref. 9 with t u ; <-
+,
' ' (> \ 6 & (?J.so) f' n order to ensure that the DG is tested under d '
h . conditions that are as close to design basis conditions D'. [ l possible, testinght:t bf performedeusing-a-power--feetttt _ ['/s.,n..a mi _ C (s._{D _9] _ Thh pcu:r fe:ter i: theren ta ha renresentative f tha actual desi inductiv: leMinn that the M
, ga... ,4 d.p
. M h would. experience. vu i,esis
/ 4,. , r I t <.r w . . .-
,. g P
. . *. .'.' w ";*This SR is modified by two Notes. The reason for Note 1 is i
i a. ,' . + that, during operation with the reactor critical,
. \c performance of this SR could cause perturbations to the electrical distribution systems that could challenge C continued steady" state operation and, as a ~ ult, plant
- Og safety it may be takensystems. . ott 2events for unplanned : kt:wleds::; theft is SR.
that satisfy 4eyiewer's Note: The above MODE restrictions may be delettT if ithn4ealemonstrated to the staff, basis, that perf5hnng the SR with-the'Jn.a.-plant' reactor in any specific of the restricted MODES can satisfynfallowing criteria, as i p applicable: y -
'N_s%.s
- a. , PerYo'rmance of the SR will not render any safety s.
system or component inoperable; d" (continued) BWR/4 STS B 3.8-21 Rev. O, 09/28/92
AL Sources-Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.9 (continued) - REQUIREMENTS ~ . .
- b. Performance 'of-the..SR will not cause_ptrturbations to any of the electrical distribut4tni' system.s that could result in a challenge tp steidy-state operation or to plant safety systems ~,'and s . ,~
- c. Performance /of the SR, or failure of the SR, wil'1 not ..
f 2. P20, ~cause', or result in, an A00 with attendant challenge
/ to plant safety systcas. _
[ %Tu v. b -~~* SR 3.8.1[1h
/b' n
This Surveillance demonstrates the DG capability to reject a full load without overspeed tripping or exceeding the predetemined voltage limits. The DG full load rejection may occur because of a system fault or inadvertent breaker tripping. This Surveillance ensures proper engine generator load response under the simulated test conditions. This test simulates the loss of the total connected lead that the u
~~
DG experiences following a full load rejection and verifies that the DG does not trip upon loss of the load. These g acceptance criteria provide DG damage protection. While the DG is not expected to experience this transient during an event, and continues to be available, this response ensures that the DG is not degraded for future application, including reconnection to the bus if the trip initiator can be corrected or isolated. In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as ng possible testing must be perfomed using a power factor sM, This power factor is chosen to be representative ( of the actual design basis inductive loading that the DG l would experience. Thep{18monthf~frequencyisconsistentwith / -@') . '- g, . t recommendation of Regulatory Guide 1.108 VRef. 9 and is (012 - intended to be co istentaith expected fue cycle lengths. (/JA 1 M d"I
/,1mm1/M/ #d*15"*'U
- This SR is modified by Notes.j The reason for Note s that during operation with the reactor critical, perfomance
/utfocb ty.M i of this SR could cause perturbations to the electrical
( /pdf de nn nN'lg j distribution sy:;tems that would challenge continued steady v.ttiu #. d state operation and, as a result, plant safety systems.
.ja (continued)
BWR/4 STS B 3.8-22 Rev. O, 09/28/92
Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 .
. ~a < s '"
O BASES SA ~%M /- . SURVEILLANCE SN (continued) y P!9 REQUIREMENTS failure of this SR, pro ' cc nt is ' ' JrimJ-pe rmance of the Surveil anc . - [k. sA REFERENCES' 1. FSAR, Section h5dl m u.:4 i A "egeleteryGuide1.15f6 3 ^t:SI t:195, 1" S , u.:s t u..t i FSAR,h i , r, i FSAR, Chapter E uai
- 6. ASTri 3Lendards. D4054-[ );Ds75- D41M-N; cii vor D1572-[ ]t D2677-f 1: and_D22 &- X,.i $#owd 7, ASME,-Beiler =d Dressure- VesseLCode, Sectian YIL O
e BWR/4 STS B 3.8-49 Rev. O,09/28/92
l DC Sources-Operating B 3.8.4 nn. D B 3.8 ELECTRICAL POWER SYSTEMS l B 3.8.4 DC Sour.es-Operating BASES BACKGROUND The DC electrical power system provides the AC emergency power system with control power. It also provides both motive and control power to selected safety related equipment. Also, these DC sebaysicm2 pivvide DC eiculisual d
--powei iv inverteT'5phfch7n7 urn puwei ihe-AC--vi-tabhows A Of'? As required by 10 CFR 50, Appendix A, GDC 17 (Ref. 1), the DC electrical power system is designed to have sufficient independence, redundancy, and testability to perform its safety functions, assuming a single failure. The DC electrical power system also conforms to the recomendations of Regulatory Guide 1.6 (Ref. 2) and IEEE-308 (Ref. 3).
The station service DC power sources provide both motive and PZL PZ2 - a control power to selected safety related 9 : p-m . n nii < -
-%x as circum-breaks centrol pc cr fer the nonsafety related y,
Q ixepp-<4-' 4100 V, and all C00 " ar.d hwer, AC disti ;betivn sysicm2. *- Each DC subsystem is energized by one 125/250 V station 7m service battery and three 125 V battery chargers (two OP2(,
@A[J] normally inservice chargers and one;rspare charger). Each battery is exclusively associated with a single 125/250 VDC bus. Each set of battery chargers exclusively associated with a 125/250 VDC subsystem cannot be interconnected.with any other 125/250 VDC subsystem. The normal and backup chargers are supplied from the same AC load groups for which the associated DC subsystem supplies the control power. The loads between the redundant 125/250 VDC subsystem are not transferable except for the Automatic Depressurization System, the logic circuits and valves of which are normally' fed from the Division 1 DC system. 6Wheir f cocebe dfsi' Ql
~-
[A , gjQt2G
,,1 The diesel generator (DG) DC power sources prov %W$u6ntr61
\ k co1egA, p ^, M, 'd s( and instrumentationj ower for their respective addition, DGt7A and 2C300 power sourceOirovidt$ circuit
~
In g
- rL" breaker control powW for the41oads on the 4160 yand-2G emergensy4uses< Each' DG DC subsystem is er 6-2F, hw
( i c.p. A ' i by one 125 V battery and oned25 Y batter charge &gized b^
' %,v - _~ A Movisions-exist for-connectin skargert g e- Terna F battery '- '
.n. s: e cwn h uNAa. Jc=ct ):/o.% cQ3 J .)
)% bi[(Y During nonnal opera T6ii~fh'e'IE loads are powered from the C-(battery chargers with the batteries floating on the system.
wn-- s su Lw{ u w - .. p e kv ~ (continued) i BWR/4 STS B 3.8-50 Rev. O, 09/28/92 l
I INSERT FOR BACKGR0Vf10 SECTI0f1 BASES 3.8.4 vat ye;+ 2 Ve et i Ve + 2. Llu+ t .a.;+ 2. VAC Lases 1 (or 2)E and 1 (or 2)F, and DG 1 (or 2)C power source provides circuit breaker control power for the respective Division II loads on 4160 VAC
- buses 1 (or 2)F and 1 (or 2)G.
U,;fI Un'& 2. Unl4 I Ue:+ L j O l l l l I j 1 HATCH UNIT 1 t (ApjIT S
DC Sources-Operating B 3.8.4 l [- l BASES (continued) t, h l ACTIONS ( (g$1 C n ,D (-{*;
- U Condition epresents onet ivision with a loss of ability "r "'t to completely respond to an event, and a potential loss of
$' gje, ability to remain energized during normal operation. It is therefore imperative that the operator's attention focus on stabilizing the unit, minimizing the potential for complete loss of DC power to the affected division. The 2 hour limit is consistent with the allowed time for an incperable DC Distribution System division.
If one of the required DC electrical power subsystems is inoperable (e.g., inoperable battery, inoperable battery charger (s), or inoperable battery charger and associated 7- -
. inoperable battery), the remaining DC electrical power subsystems have the capacity to support a safe shutdown and
(\.I_._jp~ to mitigate an accident condition. Since a subsequentworst case single failure could A owe u 6 _ result in the loss of minimum necessary DC electrical subsystems to mitigate a
; worst case accident, continued power operation should not O m* T exceed 2 hours.
The 2 hou <co)mpletion Time is based on L Regulatory Guide 1.93 (Re 7Er) and reflects a reasonable time to assess unit status a a function of the inoperable
'DC electrical power subsystem and, if the DC electrical power subsystem is not restored to OPERABLE status, to prepare to effect an orderly and safe unit shutdown.
D OR # 1 an 4 @ If the stat 4en-serdDC electrical power subsystem cannot be restored to OPERABLE status within the. required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging' plant systems. The Completion Time to bring the unit to MODE 4 is consis with the time required in Regulatory Guide 1.93 (Ref M. 7 f_ h :' W @ O W) w (continued) BWR/4 STS B 3.8-53 Rev. O, 09/28/92 ,
DC Sources-Operating B 3.8.4 c -. !
- BASES Y
ACTIONS C.1 m T (continued) \ If the DG DC elittricalpower subsystem cannot be restored } to OPERABLE status in the assofiated Completion Time, the j associated DG may be inca; fable of 'perfonning its intended
~
function and must be imediately declared 7noperable. Thisi declarationalsorequiresentryintoapplicableC6nditions) k and Required Actions for an inoperable DG, LCO 3.8.1, "AC/ xSources Operati.ng " ' SURVEILLANCE SR 3.8.4.1 REQUIREMENTS Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the charging system and the ability of the batteries to perfom their intended function. Float charge is the condition in which the charger is supplying the continuous charge .
^
required to overcome the internal losses of a battery (or battery cell) and maintain the battery (or a battery cell) &
- - (,
*'] -
, in a fully charged state. h Voltage requirements are W Qib'1) based on the nominal design voltage of the battery and are consistent with the initial voltages assumed in the battery sizing calcuqtions.VThe 7 day Frequency is co ' stent with manufactur rs comendations and IEEE-450 (Ref. J)]
O fG f7 SR 3.8.4.2 Visual inspection to detect corrosion of the battery cells and connections, or measurement of the resistance of each inter-cell, inter-rack, inter-tier, and terminal connection, provides an indication of physical damage or abnomal deterioration that could potentially degrade battery e perfomance. ( The connection C ' - [ -must-be ric ;r,are resistance limits %td!hhtd--f::r4his tharrf04-above-the-res4 SRG-stance-as-meesttredk
\. # during-instaMation-or-not abava the ceiling valne' .
w y ,, 4 established by the __mamilKturer A.
.% n J "d'
The Frequency for these inspections, which can detect conditions that can cause power losses due to resistance heating, is 92 days. This Frequency is considered [{ l O' (continued) lI BWR/4 STS B 3.8-54 Rev. O, 09/28/92 ! l
INSERT Action 3.8.4 A/B([bdd_NMIOth A Ad If one or more of the required Unit 2 DG DC electrical power subsystems is inoperable (e.g., inoperable battery, inoperable battery charger (s), or inoperable battery charger and associated inoperable battery), or if the swing DG DC electrical power subsystem is inoperable due to performance of SR 3.8.4.7 or SR 3.8.4.8, and a loss of function has not , occurred as described in Condition E, the remaining DC electrical power subsystems have the capacity to support a safe shutdown and to mitigate an accident condition. In the case of an inoperable Unit 2 DG DC electrical power subsystem, since a subsequent postulated worst case single , failure could result in the loss of certain safety functions (e.g., SGT System), continued power operation should not exceed 7 days. The 7 day Completion Time takes into account ' the capacity and capability of the remaining DC sources, and O is based on the shortest restoration time allowed for the systems affected by the inoperable DC source in the respective system Specification. In the case of an inoperable swing DG DC electrical power subsystem, since a subsequent postulated worst case single failure could result in the loss of minimum necessary DC electrical subsystems to mitigate a postulated worst case accident, continued power operation should also not exceed 7 days. The 7 day Completion Time is based upon the swing DG DC electrical power subsystem being inoperable due to performance of SR 3.8.4.7 or SR 3.8.4.8. Performance of these two SRs will result in inoperability of the DC , battery. Since this battery is common to both units, more time is provided to restore the battery, if the battery is inoperable for performance of required Surveillances, to preclude the need to perform a dual unit shutdown to perform these Surveillances. The swing DG DC electrical power subsystem also does not provide power to the same type of equipment as the other DG DC sources (e.g., breaker control power for 4160 V loads is not provided by the swing DG battery). The Completion Time also takes into account the capacity and capability of the remaining DC sources. O u HATCH UNITj 2 83.8 4
O l INSERT Action 3.8.4 A/B (continued) ([(dyl ggg,-} l 1 IL1 If a Unit 1 or swing DG DC electrical power subsystem is inoperable (for reasons other than Condition A), the remaining DC electrical power subsystems have the capacity to support a safe shutdown and to mitigate an accident condition. Since a subsequent postulated worst case single l failure could result in the loss of minimum necessary DC electrical subsystems to mitigate a postulated worst case accident, continued power operation should not exceed 12 hours. The 12 hour Completion Time provides a period of time to correct the problem commensurate with the importance of maintaining the DG DC electrical power subsystem OPERABLE. (The DG DC electrical power subsystem affects both the DG and the offsite circuit, as well as the breaker closure power for various 4160 V AC loads, but does not affect 125/250 V DC station service loads). O n v
- b. +
HATCH UNITf 2 B 3.8 .
i l 1 l O INSERT Action 3.8.4 A/B OfM d g(6(cy} ! i 1L1 If one or more of the required Unit 1 DG DC electrical power subsystems is inoperable (e.g., inoperable battery, inoperable battery charger (s), or inoperable battery charger and associated inoperable battery), or if the swing DG DC electrical power subsystem is inoperable due to performance of SR 3.8.4.7 or SR 3.8.4.8, and a loss of function has not occurred as described in Condition E, the remaining DC electrical power subsystems have the capacity to support a safe shutdown and to mitigate an accident condition. In the case of an inoperable Unit 1 DG DC electrical power subsystem, since a subsequent postulated worst case single failure could result in the loss of certain safety functions (e.g., SGT System), continued power operation should not exceed 7 days. The 7 day Completion Time takes into account the capacity and capability of the remaining DC sources, and is based on the shortest restoration time allowed for the systems affected by the inoperable DC source in the 0-r respective system Specification. In the case of an inoperable swing DG DC electrical power subsystem, since a subsequent postulated worst case single failure could result in the loss of minimum necessary DC electrical subsystems to mitigate a postulated worst case accident, continued power operation should also not exceed 7 days. The 7 day Completion Time is based upon the swing DG DC electrical power subsystem being inoperable due to performance of SR 3.8.4.7 or SR 3.8.4.8. Performance of these two SRs will result in inoperability of the DC battery. Since this battery is common to both units, more time is provided to restore the battery, if the battery is inoperable for performance of required Surveillances, to preclude the need to perform a dual unit shutdown to perform these Surveillances. The swing DG DC electrical power subsystem also does not provide power to the same type of equipment as the other DG DC sources (e.g., breaker control power for 4160 V loads is not provided by the swing DG battery). The Completion Time also takes into account the capacity and capability of the remaining DC sources. 1 Q ! HATCH UNIT 1 r ( b g B 3.8 l l
O INSERT Action 3.8.4 A/B (continued) ((( ri-FJ ye rgor] L1 If a Unit 2 or swing DG DC electrical power subsystem is inoperable (for reasons other than Condition A), the remaining DC electrical power subsystems have the capacity to support a safe shutdown and to mitigate an accident condition. Since a subsequent postulated worst case single failure could result in the loss of minimum necessary DC electrical subsystems to mitigate a postulated worst case accident, continued power operation should not exceed 12 hours. The 12 hour Completion Time provides a period of i time to correct the problem commensurate with the importance - of maintaining the DG DC electrical power subsystem OPERABLE. (The DG DC electrical power subsystem affects both the DG and the offsite circuit, as well as the breaker closure power for various 4160 V AC loads, but does not affect 125/250 V DC station service loads). (A_) HATCH UNIT 1 + Qua i & B 3.8 a
4 O INSERT SR 3.8.4.1 The voltage requirement for battery terminal voltage is based on the open circuit voltage of a lead-calcium cell of nominal 1.215 specific gravity. Without regard to other battery parameters, this voltage is indicative of a battery that is capable of performing its required safety function. A O l O HATCH UNIT 1 & UNIT 2 8 3.8
O
,V INSERT SR 3.8.4.2 h/r eslabliskad)
L- to maintain connection resistance as low as reasonably pp possible to minimize the overall voltage drop across the battery and the possibility of battery damage due to heating of connections. The resistance values for each battery connection are located in the Technical Requirements Manual k f (Reference 9).
/O V
G NJ l HATCH UNIT 1 d t"~ 2- B 3.8 l
O I INSERT SR 3.8.4.4/5 l l are established to maintain connection resistance as low as reasonably possible to miniraize the overall voltage drop . Ogg across the battery and the possibility of battery damage due d o heating of connections. .The resistance values for each -
, o battery connection are located in the Technical Requirements
( { Manual (Reference 9). ; The 18 month Frequency of the Surveillances is based on
~...
engineering judgment, taking into consideration the desired f plant conditions to perform the Surveillance. Operating (@ experience has shown that these components usually pass the , b SR when performed at the 18 month Frequency, Therefore, the Frequency was concluded to be acceptable from a reliability (standpoint. O J L h v . HATCH UNIT 1 4 UtJ M 1 B 3.8
DC Sources-Operating B 3.8.4 a~ i BASES l SURVEILLANCE SR 3.8.4.2 (continued) REQUIREMENTS acceptable based on operating experience related to detecting corrosion trends. SR 3.8.4.3 Visual inspection of the battery cells, cell plates, and battery racks provides an indication of physical damage or l fm abnormal deterioration that could potentially degrade ; { ~Tm c ),As'4 .
]-- Jattery perfomance. _
Q;( I % TheDunth4requen_cy for this SR is consistent-with-- IEEE-450 (Ref. 7), which>commendCdetailed visual inspectio,,peWclindition and rackintegrity arly _ basis. ----.._ SR 3.8.4.4 and SR 3.8.4.5 Visual inspection and resistance measurements of inter-cell,
- inter-rack, inter-tier, and teminal connections provides an indication of physical damage or abnomal deterioration that could indicate degraded battery condition. The anti-corrosion material is used to help ensure good electrical connections and to reduce teminal deterioration.
The visual inspection for corrosion is not intended to R @ EY'S NOTE require removal of and inspection under each terminal
> jot 5,ge ,
connection. gen, WOS-G; c5 The removal of visible corrosion is a preventive maintenance c SR. The presence of visible corrosion does not necessarily ' represent a failure of this SR, provided visible corrosion e is removed during perfomance of this Surveillance. 7,)qgc) The connection resistance limithfor tMs SP must bc ne more . than 20'. ebeve-thz resistence as mcasured during. i ns t a lh tiMr -o r-n o t- -above-the -eeH4 ng-valu e-e stablisheiby hu-
. y & ',
i e_ the manufacturer. The Iz mondr4cemency of these SRs is consistent wit h _ . -~
; ,po andt de-tciled visual
- IEEE-450 (Ref. 7), whic '
pl inspection ion and inYpeu.iwoft4to cell a connection resistance on a yearly basis %. ~ % O (continued) B 3.8-55 Rev. O, 09/28/92 BWR/4 STS
1 DC Sources-Operating B 3.8.4 : BASES SURVEILLANCE SR 3.8.4.6 g 4 REQUIREMENTS (continued) Battery charger capability re. rement are based on the design capacity of the chargers (Ref. . According to h Regulatory Guide 1.32 (Ref. lfr), thd2t battery charger supply C is required to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences. The minimum required amperes and duration ensures that these requirements can be satisfied. The frequency is acceptable, given the unit conditions required to perform the test and the other administrative controls existing to ensure adequate charger performance during thesef18 monthl^ intervals. In addition, this Frequency is intended to be consistent with expected fuel cycle lengths. g
- p-
.- ThisiRjrmodified -by two~ N5tes. The reason,for Sotiil is' q phat perf6 ming-the_ Surveillance wou}d.-remove a required DC electrical power subsystem from tervice, perturb the i py/, electrical distributiomsylitem,~and 4hallenge safety
~~
systems. Note A T added to this SR to acknowledge that credit mayMtaken for unplanned events that satisfy-the Suyvei g SR 3.8.4.7 A battery service test is a special test of the battery's capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length corresponds to the design duty cycle requirements as specifie$-4n Reference 4 h The Frequency of 418 monthsfis onsistent with e grecomendations of Regulatory uide 1.32 (Ref.4 ) and VRegulatory Guide 1.129 (Ref. , which state that the battery service test should be performed during refueling n ' operations or at some other outage, with intervals betweeof
/pt testsnottoexceed{18 y nths 7 (:klh 4
This SR is modified by t @& I l{7" Notes. Note 1 to SR 3.0.4./ allows the %ce per 5%weths perfomance of/P 3_83.S i 02(M 3.6 4.7 $ l (continued) ! l l l BWR/4 STS B 3.8-56 Rev. O, 09/28/92 1
A O INSERT SR 3.8.4.7A a modified performance discharge test in lieu of a service test. The modified performance discharge test is a simulated duty cycle consisting of just two rates: the 1 minute rate published for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test, both of which envelope the duty cycle of the service test. Since the ampere-hours removed by a rated 1 minute discharge represent a very small portion of the battery capacity, the test rate can be changed to that of the service test. A modified performance discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity. Initial conditions for the modified performance discharge test should be identical to those specified for a service discharge test. A O O HATCH UNIT 1 Y llOIT 2 1 l
DC Sources-Operating
, -o,4 ~ool coed 4,e, S MA 4'~~
,('} Morss4ed &% If2?- A s a peej h/ BASES k b'"d f d a' 4 c /er(cieng e ry d us cr6tg e -te s }
SURVEILLANCE SR 3.8.4.7 (continued) REQUIREMENTS ~ f - Lclicu-of SP 3.3.4.7. Thi-s-subsrt-i-tetton-4s sasceptable because-l$ . M ( ' SR LB.d .8 represent +-e-mere--severe tatt nf'.tutttery-capac4ty
" \Mihan 90 ' _8 A _ b The reason for Note 2 is that performing the Surveillance would remove a required DC electrical power h g subsystem from service, perturb the electrical distribution I system, and challenge safet a stems. Wete 3 isiadded-t 6 (M
W m this-SR-to-acknowledge-tha r dit may be taken fbr
],,s N unplanned events that satis the Surveillancef \'
h t C1 j
}4]. g SR 3.8.4.8 .A ikr, f f
(%,% b A battery fonr.ancevtest is a -test-of$onstant curren ' s f capacitytof7r+ettery, nemalij der in thecacimmd I y IM6 h s eendition., after-having-been in cerviceh to detect any change in the capacity determined by the acceptance test. ' d
\ 29 3 6_y- c;x - r The testvi-s-utended to d + amine-overall battery a y%* degradation-due-to-agea nd " sages O
V ~ [MEO
"*f " "
he acceptance crite i or this Surveilla Y nsisten with IEEE-450 (Ref. 7) and IEEE-485 (Ref. These P] i d~ . references recommend that the battery be replaced if its i
%# capacity is below 80% of the manufacturer's rating. -Ae-
[',~s A,m Am ,$ capacity of-SO'. 3 hews-thatthe battery rate of deterioration ' o , , e <a' iseincreasin g even-+f-there i: emple capac4ty-te ::t the x',c
) 6. J ~~T5ad requ+eementt ? r-[ Q n
' J O. gdCg J e Frequercy fbliisTe'st is (j)f/ the battery shows degradatioMor o"
j. O monthmor-every-M-monte -- has teacheOsk nf its
- s A5 .) . r"j r f expected liferi Degradation is Trfdicated, according to b~ h ,
C ' b 3 ' ,".,
/ than 3%erelative to itf capacity (on the previous ____lEEE T f perfoTance test orfwhen it isA10%'below the manufact'urer'sJ.7 (n b y' " ,_ C 7 rating. All these Frequencies are consistent with the
/ recommendations in IEEE-450 (Ref 17. ^
,+ l\$..Q
[cn 0 l This SR is modified by/.-tWo%otef The reason for/Nole f-is & 4 kn g;' that performing the Surveillance would remove a required DC electrical power subsystem from service, perturb the 9 electrical distribution system, and challenge safety Note 2 is added-to-thh4D to acknowledge + hat o. 667 q( aystems.A edit may be taken for unplanned even h
'Surveillancef
(~N/ n : 3 G/
. ^- D (continued)
BWR/4 "1u STS [.L vB 3.8-57
. . TM w] h Rev. O, 09/28/92
~
t 4. fsei ([<~ ~A~ T ~ - W L T t'~iz,i M }1J' - 5"~ ~bC Sources-Oper
, ,m . ~2 BASES (cotinued)
REFERENCES 1. 10 CFR 50, Appendix A, GDC 17.
- 2. Regulatory Guide 1.6.
/ LL 3. IEEE Standard 308
]~_ [Ap FSAkILChp @" ' ' '
FSAR, Chapteh [P3 h Regulatory Guide 1.9{ J IEEEStandard45bi9R) h h Regulatory Guide 1.32, February 1977. [j ~ Regulatory Guide 1.129, December 1974. (,v-% IEEE Standard 48 O 933. V l t_q-v-v v w s..g _.1
**~ J. j l
Q pg. airJseT hf 9 BWR/4 STS B 3.8-58 Rev. O, 09/28/92 l
... _ _ _ _ _ _ . _ _ . . _ . _ . - _ . . - ~ . __ . _ - .
i O : 2 INSERT }R 3.8.4.7 , The swing DG DC battery is exempted from this restriction, since it is required by both units' LC0 3.8.4 and cannot be ; performed in the manner required by the Note without > resulting in a dual unit shutdown. r O 1 I O HATCH UNIT 1 & UNIT 2 B 3.8 i
O INSERT SR 3.8.4.8A l l A battery modified performance discharge test is described in the Bases for SR 3.8.4.7. Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.8; however, only the modified performance discharge test may be used to satisfy - SR 3.8.4.8, while satisfying the requirements of SR 3.8.4.7 at the same time. 1 O O HATCH UNIT 1 4 d 01 I ^4_.
l n(./ INSERT SR 3.8.4.8 Note The swing OG DC battery is exempted from this restriction, since it is required by both units' LC0 3.8.4 and cannot be , performed in the manner required by the Note without resulting in a dual unit shutdown. O O HATCH UNIT 1 8 3.8 l
NUREG I433 COMPARISON DOCUMENT- JUSTIFICATION r] FOR DEVIATION O t O i j
JUSTIFICATION FOR DEVIATION FROM NUREG 1433 () ITS: SECTION 3.6 - CONTAINMENT SYSTEMS l PLANT SPECIFIC DIFFERENCES (continued) P.14 An SR has been added consistent with current Plant Hatch Technical Specifications. The excess flow isolation dampers are required to be OPERABLE to support the purge valve Technical Specification allowances. Thus, this SR demonstrates that the dampers are OPERABLE. P.15 This allowance has been deleted since the Plant Hatch design includes only one solenoid for each LLS valve. P.16 The Frequency has been changed to "In accordance with the Inservice Testing Program," consistent with other Frequencies for similar components covered by the IST Program. The current IST Frequency is 92 days, so no technical change is made. P.17 This Specification has been deleted for Unit 1, since Unit 1 does not have this system installed. For Unit 2, this Specification has been deleted as justified in the Georgia Power Company letters from J.T. Beckham to the NRC, dated January 6, 1994 and February 3, 1994. Subsequently, the () \_g e NRC issued this change as Amendment 132 to the Unit 2 TS by letter dated March 17, 1994. P.18 An ACTION has been added to allow both RHR suppression pool cooling subsystems to be inoperable for up to 8 hours prior to requiring a unit shutdown. Thia ACTION is allowed in the current Hatch Unit 2 TS (Hatch Unit 1 does not have a TS requirement on this system), and is consistent with the NUREG ACTION provided when two RHR suppression pool spray subsystems are inoperable (LCO 3.6.2.4, ACTION B) . The reasons for allowing 8 hours is similar to the reasons why 8 hours is allowed for suppression pool spray; the proposed 8 hour Completion Time provides some time to restore at least one subsystem, yet is short enough that operating an additional 8 hours is not risk significant. In addition, if one of the two subsystems is restored prior to the expiration of the 8 hours, a unit shutdown is averted. Thus, the potential of a unit scram occurring during the shutdown required by the NUREG ACTIONS, which then could result in the need for a subsystem when it is inoperable, has been decreased. The first condition of Condition B has been modified to reflect the addition by deleting the words "of Condition A", since this condition now applies both to current Condition A, as well as new Condition B. O HATCH UNITS 1 AND 2 3 REVISION (
JUSTIFICATION FOR DEVIATION FROM NUREG 1433 ITS: SECTION 3.6 - CONTAINMENT SYSTEMS ( ) { PLANT SPECIFIC DIFFERENCES (continued) P.55 The proper description of the system has been provided. P.56 Unit 1 uses a CAD System, not H2 Recombiners and the Drywell Cooling System. Thus the reference to the other systems has been changed. P.57 The safety analysis for Plant Hatch Unit 2 does not assume an initial oxygen concentration. However, the Specification does meet criterion 4 of the NRC Policy Statement, thus it is retained for this reason. P.58 As described in comment No. P.27, the Unit 1 secondary containment includes both the common refueling floor and the Unit i reactor building. Due to the design of the buildings, the reactor building can be separated from the common refueling floor such that Unit 1 secondary containment only includes the common refueling floor. This
" modified" containment is currently licensed to be used when Unit 1 is not in MODES 1, 2, and 3. Since the OPDRVS applicability has been added and it affects the reactor
/g building, " modified" containment is not allowed during (,,/ OPDRVs. This description is included in appropriate places. P 59 This phrase has been deleted since Plant Hatch safety analysis does not assume a wind angle. P.60 The filter tests are not all in accordance with RG 1.52, Rev.2. However they are in accordance with the VFTP. Thus, this is the reference mentioned in the Bases. P.61 The MSIV leakage limit for Plant Hatch Unit 2 was proposed to be changed per GPC letter dated January 6, 1994, and subsequently, the NRC issued this change as Amendment 132 to the Unit 2 TS by letter dated March 17, 1994. Amendment 132 includes the 250 scfh limit and the requirement to reduce leakage to 5 11.5 sofh if an MSIV exceeds the 100 scfh limit. Further technical discussion is provided in the GPC January 6, 1994, letter and the corresponding NRC SER dated March 17, 1994. P.62 The acceptance criteria time has been changed to s 100 seconds, consistent with the time in the SR (in the Technical Specifications) and the safety analysis. P.63 The Unit 1 safety analysis does not have the same level of detail as the Unit 2 analysis. Therefore, the Unit 1 Bases has been modified to reflect that the Unit 2 FSAR analysis O' is appropriate for Unit 1. See comment No. P.46. HATCH UNITS 1 AND 2 9 REVISION N
l l l gg JUSTIFICATION FOR DEVIATION FROM NUREG 1433 i ITS: SECTION 3.6 - CONTAINMENT SYSTEMS () l GENERIC APPROVED /PENDING CHANGES TO NUREG 1433 ) l GP.1 Changed to be consistent with NUREG change package BWR-14 , Items C.1, C.3, C.4, and C.6. I l GP.2 Changed to be consistent with NUREG change package BWR-15 j Items C.1, C.2, C.3, C.4, C.5, C.6, C.7, C.8, C.9, C.10, C.11, C.13, C.14, C.15, C.17, and C.19. GP.3 Changed to be consistent with NUREG change package BWR-16 Items C.3, C.4, C.7, C.8, C.9, C.15, C.20, C.22, C.23, C.24, C.25, C.26, and C.28. l GA.4 Change approved per package BWR-06 Items C.4 and C.5, S/20/93. GA.5 Change approved per package BWR-04 Items C.7 and C.8, 4/22/93. GA.6 Change approved per package NRC-02 Item C.15, 5/20/93. Ge.7 Change approved per package NRC-07 Item C.1, Rev. 1, i 12/16/93. /~)T L A U IIATCH UNITS 1 AND 2 10 REVISION p i
JUSTITICATION FOR DEVIATION FROM NUREG 1433 ITS: SECTION 3.8 - ELECTRICAL POWER SYSTEMS ( PLANT SPECIFIC DIFFERENCES [ P.10 The upper voltage limit is restricted for steady state i operation (refer to Discussion of Change "M.8" in the i markup of the current Technical Specifications). However it is not the intent that starting voltage excursions be- l considered a failure of the DG Surveillance. P.11 Due to the Plant Hatch design, both fuel oil transfer pumps ' are needed to ensure fuel oil transfer capability in the event of a design basis accident coincident with a single active failure. The Plant Hatch requirement for fuel oil requires that the five tanks (one per DG) have sufficient fuel oil to provide for four DGs for 7 days. Four tanks do not carry enough fuel oil to power the four remaining DGs (assuming the single failure is a DG). Thus, this is why capability to transfer fuel from one tank to a different DG : is required. Each tank has two associated fuel transfer pumps, and while the pumps are 100% capacity pumps, their i power supplies are such that the loss of a power supply with one pump already inoperable could result in the loss of ability to transfer the fuel oil from that tank. < Therefore, since this fuel oil is being credited, both fuel oil transfer pumps per required tank are needed. A 30 day completion time is proposed for when one pump is inoperable since the fuel oil can still be transferred assuming no additional single failures, and in some cases, can be transferred if the single failure does not affect the power supply to the remaining pump. In addition, since the NUREG assumed only one pump was required, it placed the requirement in DG LCO 3.8.1. Since both are now required i at Plant Hatch, it has been placed in DG fuel oil LCO 3.8.3 for consistency with other fuel oil requirements. The SR l has also been moved to this new location. P.12 The current licensing basis at Plant Hatch provides for each DG reaching only " synchronous speed" initially on each fast start, and applies the voltage and frequency range limitations only for " steady state" operation. P.13 The plant specific single largest load component / load value is proposed to be supplied with the following criteria:
" greater than or equal to the single largest post accident load." This presentation supplies the necessary Technical Specifications requirement while avoiding the confusion associated with a specific values and/or a specific component. When utilizing a value, constructing the test to ,
trip the component operating at full load may not quite , result in meeting the value, but should satisfy the test Oe j
JUSTIFICATION FOR DEVIATION FROM NUREG 1433 ITS: SECTION 3.8 - ELECTRICAL POWER SYSTEMS ( ) PLANT SPECIFIC DIFFERENCES (continued) P.13 (continued) requirement since the largest load was rejected satisfactorily. P.14 These limits imposed on return to steady state frequency following a single load rejection, are controlled by plant l procedures, and are not presented as specific TS requirements. The specific criteria referenced would not be appropriate for certain methods of performing this test, e.g., if performed while the DG waF loaded only with the single largest load. Furthermore, hin criteria is not currently included in the Plant " + t .icensing basis. The specified voltage is the nominal _ voltage range and is consistent with the CTS voltage range. P.15 At Plant Hatch this test is only performed with the DG separated from offsite power (as detailed in the proposed Bases). In this case, a specific power factor requirement r" is not appropriate since power factor will be determined ( ,}/ solely by the connected equipment. Therefore, the portions of Generic change BWR-17 Items C.2 and C.3 regarding power factor are omitted. P.16 On the initial DG start the minimum DG parameters (e.g., frequency) are the limits which must be reached within 12 seconds, and the ranges apply to steady state operation. This is in agreement with the current Plant Hatch Technical Specifications, and is clarified with these proposed changes. P.17 With offsite power available, all loads remain energized. The Plant Hatch design does not sequence loads onto offsite power. Verification that components remain energized from offsite power is already accomplished with various system functional tests required by individual system Specifications. P.18 Typographical / grammatical error or Writer's Guide convention corrected. P.19 Corrected for current Plant Hatch licensing basis. In addition, NUREG SR 3.8.3.6 is a preventive maintenance type of SR. This type of SR is generally allowed to be plant controlled and not controlled by Technical Specifications. + This is similar to the current DG inspection SR that has \- been removed from TS. HATCH UNITS 1 AND 2 4 REVISION A
l JUSTIFICATION FOR DEVIATION FROM NUREG 1433 ITS: SECTION 3.8 - ELECTRICAL POWER SYSTEMS P.20 Comment number not used. O O uATcu u m S 1 Aso 2
/ a REv1SIesgg
JUSTIFICATION FOR DEVIATION FROM NUREG 1433 (( ) ITS: SECTION 3.8 - ELECTRICAL POWER SYSTEMS PLANT SPECIFIC DIFFERENCES (continued) P.41 The bracketed values of resistance specified in the NUREG are vendor recommended values; that is, values at which some action should be taken, not necessarily when the OPERABILITY of the battery is in question. In addition, the safety analyses do not assume a specific battery resistance value, but typically assume the batteries will supply adequate power. Connection resistance is determined by the contact resistance between the connector and battery post, as well as the material, shape, and length of the connector bar/ cable. Contact resistance is affected by the irregularity of contact surfaces, the level of corrosion between contact surfaces, and the tightness of the connection. The type of connection is determined by the location of the connection (inter-cell, inter-tier, inter-rack, and terminal) and is characterized by connectors varying in shape and length. The allowable resistance range for each type of connection is different for a particular battery. Since batteries of different sizes may require connectors of different sizes and lengths, s_) connection resistance is often different from one plant to another and from one battery to another. A single OPERABILITY resistance value for each battery connection type is not practical. The key issue is the overall battery resistance. Between surveillances, the resistance of each connection varies independently from all the others. Some of these connection resistances may be higher or lower than others, and the battery may still be able to perform its function and should not be considered inoperable. The proposed Hatch ITS include the surveillance to survey the battery resistances to ensure they are within limits. These limits will be specified in the Technical Requirements Manual. This allows appropriate battery resistance values to be specified and the levels at which action will be taken if: 1.) the manufacturer recommended values are exceeded and 2.) when the OPERABILITY of a battery is questioned. Current procedures address resistance values and evaluate changes in resistance values. Please note that the CTS do not include a surveillance (s) equivalent to SR 3.8.4.5 and SR 3.8.4.2. As identified in HATCH UNITS 1 AND 2 7A REVISION A' '
1 1 JUSTIFICATION FOR DEVIATION FROM NUREG 1433 () ITS: SECTION 3.8 - ELECTRICAL POWER SYSTEMS PLANT SPECIFIC DIFFERENCES P.41 (continued) the Markup of the Current Technical Specifications Section 3.8.4 M.1 (Unit 1) and M.3 (Unit 2), these surveillances are additional requirements, even without specific resistance values. Based on the above discussion, we believe the Hatch ITS proposed specification is appropriate. In summary, the NUREG values specified tend to be manufactures' values, not OPERABILITY values. The configuration of the batteries will lead to several different values, not just three. Hatch CTS do not include these requirements, and we currently have procedures for performing battery inspections. P.42 The substitution of a modified performance discharge test for a service test may be helpful to gather additional data points for trending capacity as a battery nears its end of life, but before more frequent testing would normally be O required. For this reason, this substitution should be allowed, though not required. Since the modified performance discharge test envelopes the duty cycle of the service test, thus making it a harsher test on the battery, it may be substituted for the service test at any time. (This is stated in the draft revision of IEEE-450.) Also, to simplify procedures, the use of a modified performance test may be substituted for the service test throughout the life of the battery. Design configuration controls should verify the continued enveloping of the service test duty cycle by that of the modified performance discharge test. t U' HATCH UNITS 1 AND 2 7B REVISION B
JUSTIFICATION FOR DEVIATION FROM NUREG 1433 ITS: SECTION 3.8 - ELECTRICAL POWER SYSTEMS GENERIC APPROVED /PENDING CHANGES TO_FUREG 1433 GA.1 Change approved per package BWR-07 Item C.1, 9/8/93. GA.2 Change approved per package BWOG-05 Item C.1, S/20/93. GP.3 Changed to be consistent with NUREG change package BWR-17 Items C.2, C.3, C.4, C.5, C.6, C.7, C.9, C.10, C.11, C.12, l and C.13. Change approved per package WOG-13 Item C.4, C.5, C.8, C.9 1 GA.4 ) and C.12; 9/15/93, 7/28/93, and 4/22/93. GA.5 Change approved per package WOG-14 Item C.1, 5/20/93. Change approved per package CEOG-01 Items A, C.1, C.3; GA.6 6/29/93 and 7/28/93. GA.7 Change approved per package NRC-02 Item C.1, 7/28/93. GP.8 Changed to be consistent with NUREG change package BWR-18 Item C.64 and C.65. GA.9 Change approved per package WOG-26 Item C.4, 7/28/93. GA.10 Change approved per package BWR-08 Item C.1 and C.4, 10/7/93. GP.11 Changed to be consistent with NUREG change package BWR-18 Item C.2. GP.12 Comment number not used. GP.13 Changed to be consistent with NUREG change package BWR-22 Item C.1. GA.14 Changed to be consistent with NUREG change package NRC-15, Item C.1, 4/30/94. GP.15 Changed to be consistent with NUREG change package NRC-20. REVISION $( HATCH UNITS 1 AND 2 )}}