ML20196J929
| ML20196J929 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 06/27/1988 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20196J901 | List: |
| References | |
| NUDOCS 8807060481 | |
| Download: ML20196J929 (17) | |
Text
{{#Wiki_filter:-. ATTACHMENT A-1 Revise.the Beaver Valley Unit No. 1 Technical Specifications as 'follows. Remove Pages Insert Pages 3/4 3-12; 3/4 3-12 3/4 3-13 3/4 3-13 3/4 3-19a 3/4 3-19a 3/4 3-24a 3/4 3-24a 3/4 3-27a 3/4 3-27a 3/4 3-31a 3/4 3-31a 3/4 4-2d 3/4 4-2d t i l 8807060481 880627 ~ PDR ADOCK 05000334 p PNV
'-d TABLE 4.3-1, g JNTINUED) W to REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS mx Channel Modes in Which Channel channel Functional Surveillance Functional Unit Check Calibration Test Required E 15. Steam Feedwater Flow Mis-S R M 1, 2. match and Low Steam Gen-8 erator Water Level C $ 16. Undervoltage-Reactor N.A. R M 1. 8 Coolant Pumps v 17. Underfrequency-Reactor N.A R M 1 Coolant Pumps y 18. Turbine Trip xom a. Auto Stop Oil Pressure N.A. N.A. S/U(1) 1, 2 Os b. Turbine Stop Valve N.A N.A. S/U(1) 1, 2 @^ Closure ~o 0 19. Safety Injection Input from N.A N.A. R 1, 2 [- @N ESF H 20. Reactor Coolant Pump N.A N.A. R N.A Breaker Position Trip 21. Reactor Trip Breaker N.A. N.A. M(5, 11) 1, 2, 3 *, and S/U(1) 4*, 5* 22. Automatic Trip Logic N.A. N.A. M(5) 1, 2, 3 *, 4*, 5* 23. Reactor Trip System Interlocks A. P-6 N.A. N.A. M(9) 1, 2 B. P-8 N. A.. N.A. M(9) 1 C. P-9 N.A. N.A. M(9) 1 D. P-10 N.A. N.A. M(9) 1 E. P-13 N.A. R M(9) 1 24. Reactor Trip Bypass N.A. N.A. M(12), R(13), 1, 2, 3
- Breakers S/U(1) 4*,
5*
~ TABLE 4.3-1 (CONTINUED) ' NOTATION With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal. If not performed in previous 7 days. (1) (2) Heat balance only, above 15% of RATED THERMAL POWER. (3) - Compare incore to excore axial imbalance above 15% of RATED THERMAL POWER. Recalibrate if absolute difference > 3 percent. (4) - (Noe Used) l (5) Each train tested every other month. (6) Neutron detectors may be excluded from CHANNEL CALIBRATION. (7) - Below P-10. (8) - Below P-6. Required only when below Interlock Trip Setpoint. (9) The CHANNEL FUNCTIONAL TEST shall independently verify the (10) OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify the _ OPERABILITY of the Bypass Breaker trip circuit (s). The CHANNEL ~UNCTIONAL TEST shall independently verify the (11) OPERABILITY of the undarvoltage and shunt trip attachments of the Reactor Trip Bre%;ers. (12) Local manual shunt trip prior to placing breaker in service. (13) Automatic undervoltage trip. BEAVER VALLEY - UNIT 1 3/4 3-13 PROPOSED WORDING
m TABLE 3m3-3 (Continued) m> ENGINEERED SAFETY FEATURE ACTU2. TION SYSTEM INSTRUMENTATION E MINIMUM. Q TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES , ' ACTION c d 7. AUXILIARY FEEDWATER e a. Steam Gen. Water Level-Low-Low (Loop Stop Valves Open) i. Start Turbine m Driven Pump 3 stm. gen. 2/stm. gen. 2/stm. gen. 1, 2, 3 14 any stm. gen. mw 01 ii. Start Motor @g Driven Pumps 3/stm. gen. 2/stm. gen. 2/stm. gen. 1, 2, 3 14 i any 2 stm. gen. any 2 stm. gen. _8l$ wa o H b. Undervoltage-RCP (Start Turbine Driven Pump (3)-1/ bus 2 2 1 14 c. S.I. (Start Motor-Driven Pumps) See 1 above (all S.I. initiating functions and requirements) d. Emergency Bus Under-voltage (Start Motor 1/ bus 1 1 1, 2, 3 18 l' Driven Pumps) e. Trip of Main Feedwater Pumps - (Start Motor Driven Pumps) 1/ pump 1 1 1, 2, 3 18 l
Wg TABLE 3.3-4 (Continued) M x-ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS b M< FUNCTIONAL UNIT TRIP SETPOINT -ALLOWABLE VALUES C b ~ 7. AUXILIARY FEEDWATER a. Steam Generator Water Level-low-low > 12% of narrow range > 11% of narrow range instrument span each instrument span each g steam generator steam generator ?ow g]; b. Undervoltage - RCP 2; 2750 volts RCP bus 2; 2725 volts RCP bus a voltage voltage xE oN$$ c. S.I. See 1 above (all SI Setpoints) E O d. Emergency Bus Undervoltage 1 3350 volts 1 3325 volts l e. Trip of Main Feedwater Not Applicable Not Applicable l Pumps
TABLE'3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 11. Steam Generator Water Level-Low-Low a.
Motor-driven Auxiliary 60.0 Feedwater Pumps ** b. Turbine-driven Auxiliary 60.0 Feedwater Pumps ***
- 12. Undervoltage RCP a.
Turbine-driven' Auxiliary 60.0 Feedwater Pumps
- 13. Emergency Bus Undervoltage a.
Motor-driven Auxiliary 60.0 Feedwater Pumps
- 14. Trip of Main Feedwater Pumps a.
Motor-driven Auxiliary 60.0 Feedwater Pumps NOTE: Response time for Motor-driven Auxiliary 60.0 Feedwater Pumps on all S.I. signal starts
- on 2/3 any Steam Generator
- on 2/3 in 2/3 Steam Generators BEAVER VALLEY - UNIT 1 3/4 3-27A PROPOSED WORDING
er TABLE 4.3-2 (CONTINUED) g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION M:o Channel Modes in Which E Channel. Channel Functional Surveillance Q Functional Unit Check Calibration Test Required 7. AUXILIARY FEEDWATER C Z H a. Steam Generator Water 6 Level-Low-Low S R M 1, 2, 3 e b. Undervoltage - RCP f' R M 1, 2 m c. S.I. See 1 above (all SI surveillance requirements) cc o ta (n N $^ d. Emergency Bus N/A R R 1, 2, 3 I xy Undervoltage o te
- o w I
CD c. Trip of Main O Feedwater Pumps N/A N/A R 1, 2, 3 l-l [ L
REkCTOR COOLANT SYSTEM SURVEILLANCE' REQUIREMENTS 4.4.1.3.1 The required residual heat removal loop (s) shall be determined OPERABLE per Specification 4.0.5. 4.4.1.3.2 The required reactor coolant pump (s), if not in operation, shall be determined to be OPERABLE once per 7 days be verifying correct breaker alignments and indicated power availability. 4.4.1.3.3 The required steam generator (s) shall be determined -OPERABLE by verifying secondary side level equivalent to 12% narrow range at least once per 12 hours. 4.4.1.3.4 At least one coolant loop shall be verified te be in operation and circula_ing reactor coolant at least once per 12 hours. l l BEAVER VALLEY - UNIT 1 3/4 4-2d PROPOSED WORDING
.. -. ~. ~ ATTACHMENT A-2 Revise the Beaver Valley Unit No. 2 Technicdl Specifications as follows: Remove Pages Insert Pages 3/4 3-20 3/4 3-20 3/4 3-27 3/4 3-27 3/4 3 3/4 3-31 3/4 3-37 3/4 3-37 3/4 4-4 3/4 4-4 l l 1 1
n TABLE 3.3-3 (Continued) E ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 9 MINIPRJM M TOTAL NO. CHANNELS CHANNELS APPLICABLE g FbNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 7. AUXILIARY FEEDWATER (Continued) E Z d. Safety Injection (Start See 1 above (all SI initiating functions and requirements) Motor-Driven Pumps) m e. fterbine-Oriven-Pur.9 (2)-1/Trair. 1 i 1, 2, 3 18 -Dkaharge Pressure Lc - -WRh-Stes: '! he Open- -{St&rt "oter-Drive.; Pg;) %h d Y. Trip of Main Feedwater 1/ pump 2 2 1,2,3 18 ^ gw Pumps (Start Motor v1 Driven Pumps) 8. ENGINEERED SAFETY FEATL'9E INTERLOCKS E a. Reactor Trip, 2 1 2 1,2,3 45 S P-4 i c., s b. Pressurizer Pressure, 3 2 2 1,2,3 38 P-11 c. Low-Low Tavg, P-12 3 2 2 1, 2,- 3 38 ~ e m-
n. n TABLE 3.3-4 (Continued) E ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 9 TOTAL SENSOR TRIP N FUNCTIONAL UNIT ALLOWANC5(TA) DRIFT (S) SETPOINT ALLOWABLE VALUE Z E 7 7. AUXILIARY FEEDWATER (Continued) E b. Steam Generator Water U Level--Low-Low m 1. Start Turbine 15.5 14.18 1.67 > 15.5% of > 14.7% of Driven Pump harrow range harrow range instrument instrument i span span k o 2. Start Motor 15.5 14.18 1.67 > 15.5% of > 14.7% of ig Driven Pumps Earrow range Harrow range instrument instrument pa span span E" g[A> c. Undervoltage - RCP (Start 27.7 1.39 0.0 > 75% of nominal > 73% of nominal Turbine Driven Pump) Eus voltage Eus voltage s d. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints (Start Motor-Driven Pumps) and Allowable Values. -e--Turbine-Deiver. Pu;;;p 5.0 2.0 0 Discharg Disch:rge - --Discharge Pressurc !. w pressure pressure -with-Steam-Valve-Open-1 150 psig with 1 115 psig wie--- -(Start-Motor-Deiven-Pumps-)- -srteam-inlet steam-inlet -valves Open valves Oper eX. Trip of Main Feedwater M.A. N.A. N. A. N.A. N.A. Pumps (Start Motor-Driven Pumps)
TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 9. Loss of Power a. 4.16kv Emergency Bus Undervoltage 1 1 0.1 sec. (Lc,ss of Voltige) (Trip Feeder) b. 4.16kv and 480v Emergency Bus Under-90 1 5 sec. voltage (Degraded Voltage)
- 10. (Intentionally blank)
(
- 11. Steam Generator Water Level-Low-Low a.
Motor-driven Auxiliary -< 60.0 Feedwater Pump ** b. Turbine-driven Auxiliary -< 60.0 Feedwater Pump ***
- 12. Undervoltage RCP a.
Turbine-driven Auxiliary 5 60.0 Feedwater Pump (
- 13. Trip of Main Feedwater Pumps a.
Motor-driven Auxiliary 5 60.0 Feedwater Pumps -14r-Turbi r.: Ot4ver, hxiliary Feedwater- --Pump-Oischarge Peessure Lew a. "otor-driven Auxiliary i 60 + --Feedwatee-Pumps-- /VN. Control Room High Radiation a. Control Room Ventilation Isolation-1 180(6) 1 l l
- on 2.'3 in i/3 Steam Geaerstors
- on 2/3 any Steam Generator BEAVER 'e tEY - UNIT 2 3/4 3-31 l
- on 2.'3 in i/3 Steam Geaerstors
A/20f0StD WCAD//6-
e m ~^ ~ TABLE 4.3-2 (Continued) ,.~ ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 9 SURVEILLANCE REQUIREMENTS h CHANNEL MODES IN WHICH l; CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE 7 FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED E7 AUXILIARY FEEDWATER (continued) G d. Safety Injection (Start Motor-See 1 above (all SI surveillance requirements) ~ Driven Pumps) c. Tu:ti.7; driven-Fu:::p M.A. R R 1, 2, 3 -Bischari;c Pressurt Lm Wth- -Stea;;; '.'alve Optr. (Start "eter M --Oriver. Pu;;;ps) s kw e)(. Trip of Main N.A. N.A. R 1,2,3 rni Feedwater Pumps U (Start Motor-Driven Pumps) w E# 8. ENGINEERED SAFETY FEATURE INTERLOCKS
- b a.
Reactor Trip, P-4 N.A. N.A. R 1,2,3 t9 b. Pressurizer Pressure, P-11 H.A. R M 1,2,3 Lor Low T,yg, P-12 N.A. R M 1,2,3 c. e a
REACTOR COOLANT SYSTEM ( SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required residual heat removal loop (s) shall be determined OPERABLE per Specification 4.0.5,-and by verifying-that-each-retidu& heat-7::cval pump develops a dif ferent4al presstice-of->-126-ptid-on-reeiretdet-ion- -44s w 4.4.1.3.2 The required reactor coolant pump (s), if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker align-ments and indicated power availability. 4.4.1.3.3 The required steam generator (s) shall be determined OPERABLE by ( verifying secondary side level greater than or equal to 15.5 percent narrow range at least once per 12 hours. 4.4.1.3.4 At least one coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours. ( BEAVER VALLEY - UNIT 2 3/4 4-4 ftedfGS Eb lx10A'b LU6-
ATTACHMENT B Safety Analysis Beaver Valley Power Station Proposed Technical Specification Change Unit 1 Change No. 151 Unit 2 Change No. 7 Description of amendment request: The proposed amendment would (1) revise the BV-1 surveillance frequency for item 19 on Table 4.3-1 from monthly to refueling, -(2) delete the BV-1 and BV-2 Turbine Driven Pump Discharge Pressure Low auxiliary feedwater initiating signal from the ESF instrumentation tables, and (3) revise the BV-1 and BV-2 Residual Heat Removal (RHR) pump surveillance requirements. The BV-1 surveillance frequency for Table 4.3-1 item 19, Safety Injection Input from ESF, has been changed from monthly to refueling for clarification of the required instrumentation testing frequency. The manual ESF input is only required to be tested every 18 months as specified by note (4) and the automatic SI input to the reactor trip logic is tested monthly on a staggered test basis in accordance with item 22. Therefore, item 19 only applies to the manual ESF input which is tested every 18 months. As a result of this change note (4) is no longer required and is being deleted. These changes are consistent with the BV-2 technical specifications and do not affect the FSAR or any regulatory basis. The Turbine Driven Pump Discharge Pressure Low auxiliary feedwater initiating signal has been deleted from the BV-1 and BV-2 ESF Instrumentation tables since this is only a backup cignal and not a primary initiatin, signal. The accidenc analyses do not take credit for this
- signal, therefore, this signal is not required to satisfy any criteri' related to auxiliary feedwater initiation.
This change does not aff .t the FSAR or any regulatory basis. The BV-1 ano BV-2 surveillance requirement for RHR pump testing, 4.4.1.3.1 has been revised to provide consistency with the other pump testing requirements. Currently, this surveillance requirement includes verifying thet each pump develops a differential pressure of 2 112 psi when tested on recirculation flow. We have found that testing the pumps on recirculation subjects the pumps to higher vibration levels than full flow testing and leads to unnecessary pump degradation. We have also determined that full flow testing in accordance with ASME XI as required in specification 4.0.5 provides a more accurate indication of pump operability, since this is the condition at which the pump was designed to be operated. The pump testing requirements can be satisfied when the RHR system is normally in service without any special valve lineups or system configuration changes which would unnecessarily complicate the testing procedures. Therefore, this change will simplify the testing requirements and does not affect the FSAR or any regulatory basis. l i l l l l l
(c; ATTACHMVNT C No Significant Hazards Evaluation Beaver Valley Power-Station Proposed Technical Specification Change -Unit 1 Change No. 151. Unit 2 Change No. 7 Basis _ for proposed-no significant hazards-consideration determination: _ The Commission has provided standards for determining whether- .a significant hazards consideration exists (10 CFR 50.92(c)). A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not (1)' involve a significant increase in the probability or consequence of an accident previously evaluated, (2) create the possibility of a new or. different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The. proposed changes do not involve a significant hezard consideration because: 1. The BV-1 surveillance frequency for Table 4.3-1 item 19 Scfety Injection Input from ESF has been changed from monthl: - to refueling for clarification of the. required instrumentction testing frequency. The manual ESF input is only required to be tested every 18-months as specified by note (4) and the automatic SI input to the reactor trip logic is tested monthly on a staggered test basis in accordance with item 22. Therefore, item 19 only-applies to the manual ESF input which is tested every 18 months and R will replace M(4) as the surveillance frequency for this item. This change is consistent 'vith the BV-2 Technical Specifications and is administrative in nature by providing clarification of the surveil.1.ance frequency. The Turbine Driven Pump Discharge Pressure Low signal which initiates a signal to start the motor driven auxiliary feedwater pumps was added to the BV Technical Specifications by Amendment 90 and was added to the BV-2 Technical Specifications to reflect the change to BV-1. This signal was added to the Technical Specifications to reflect all the auxiliary feedwater initiating signals identified in the FSAR. However,~ we have determined that this is only a backup signal and not a primary auxiliary feedwater start signal taken credit for in the accident analysis. Therefore, to reduce l._ confusion related to this signal and to maintain consistency with the FSAR assumptions, we have determined that this signal should l be deleted from the Technical Specifications. i Surveillance requirement 4.4.1.3.1 for BV-1 and BV-2 RHR pump testing has been revised to eliminate the requirement to test the pumps on recirculation. There is no ASM3 XI requirement to test the RHR pumps on recirculation and all the pump data required to determine pump performance can be obtained when operating the pumps during Modes 4, 5 and 6 at higher flows. Testing the purc.ps L
r ' ATTACHMENT C Page 2 in-the normal system configuration is consistent with the pump testing requirements of ASME XI and was requested by NRC reviewers in meetings held at Beaver Valley. FSAR Section 9.3 states that the active compenents of the RHRS are in intermittent use during normal plant operation and no additional periodic toits are required. Testing the pumps on recirculation normally subjects the pumps to higher vibration levels than operation in the normal configuration and leads to unnecessary pump degradation. Pump testing performed in accordance with ASME XI l provides a more accurate indication of pump operability since this is the system condition for which the pumps were designed and normally operated. Therefore, this change is administrative in nature because it allows pump testing to be performed at normal operating conditions to verify pump operability and does not affect the FSAR. Therefore, these changes are administrative in nature providing clarification. and consistency to improve the understanding of the requirements and do not affect the probability of occurrence or the consequences of a previously evaluated accident. 2. No change in plant operations or to equipment or components is required. These changes are administrative in nature and will not affect the safe operation of the plant. These changes are consistent with the FSAR accident analyses and will not create the possibility of a new or different kind of accident from those described in the FSAR. 3. Thesc changes are administrative in nature and are consistent with the accepted criteria for operating, testing and verification of system operability. These changes do not affect the Technical Specification Bases and will not reduce the margin of safety of the plant. The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples (51 FR 7751) of amendments that are considered not likely to involve significant hazards considerations. Example (i) relates to "A purely administrative change to technical specifications: for example, a change to achieve consistency throughout the technical specifications, correction of an
- error, or a
change in nomenclature". The proposed change relates to this example in that the proposed changes are requested to provide clarification for consistency between the various BV-1 and BV-2 specifications concerning the surveillance frequency for Safety injection Input from ESF on Table 4.3-1; deleting the backup auxiliary feedwater initiating
- signal, Turbine Driven Pump Discharge Pressure Low, from the ESF instrumentation tables; and removing the requirement to test the RHR pumps on recirculation.
Therefore, based on the above considerations, it is proposed to characterize the change as involving no significant hazards consideration. .}}