Text

Forwards Revised FSAR Pages Reflecting Changes Due to Field Testing Results & in Support of Tech Spec Review,Per 841004 Meeting
	ML17298B471
Person / Time
Site:	Palo Verde
Issue date:	11/05/1984
From:	Van Brunt E; ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	Knighton G; Office of Nuclear Reactor Regulation
References
	ANPP-31053-EEVB, NUDOCS 8411070013
	Download: ML17298B471 (202)
	v • d • e

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REGULATORY. FORMATION DISTRIBUTION S . EM (RIOS)

ACCESSION NBR;8411070013, DOC DATE: 84/11/05 NOTARIZED: YES DOCKET Unit ii Arizona Publi

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FACIL'.STN"50 528 Palo Verde Nuclear Stations 05000528 STN>>SO-529 Palo Verde Nuc'lear Stations 'Unit 2g Arizona Publi 05000529 STN-50 530 Palo Verde Nuclear Stations Unit 3i Arizona 'Publi 0S000530 AUTH'AME" 'UTHOR AFFILIATION VAN BRUNTiE',ED Arizona "Public Service Co.

RECIP ~ NAME~ RECIPIENT AFFILIATION KNIGHTONiG ~ Licensing Branch 3 I

SUBJECT:

" Forwards revised FSAR pages> reflecting, changes due'o field testing results 8 in suppor t of Tech Spec r eviewiper 841004 meeting, DISTRIBUTION CODE: 80010 COPIES RECEIVED:LTR g'NCL TITLE'. Licensing Submittal: PSAR/FSAR Amdts 8, Related Correspondence

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NOTES:Standardized plant ~ 05000528 Standardized plant ~ 05000529 Standardized plant ~ 05000530 RECIP IENTi COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE'/NAME LTTR ENCL NRR/Ol./ADL NRR L83~ LA 1

1 0

0 NRR L83 BC L'ICITRAi E'1 1 1

0 1

INTERNAL'DM/LFMB 1 0 ELO/HOS3 1 0 1 1 IE/OEPER/EPB 36 1 1 IE'ILE'E'/OEPER/jRB IE/DQA SIP/QA821 3S 1 1 1 1 NRR 1 NRR/OE/AEAB 1 0 ROEgM,L'RR/DE/CEB NRR/DE/EQB il 13 1

2 1

2 NRR/DE/EHEB NRR/DE/GB 28 1

2 1

2 NRR/DE/MEB 18 1 1 NRR/DE/MTEB 17 1 1 NRR/DE/SAB 24 1 1 NRR/DE/SGEB 25 1 1 NRR/OHFS/HFEB40 1 1 NRR/DHFS/LQB 32 1 1 NRR/OHFS/PSRB 1 1 NRR/OL/SSPB 1 0 NRR/DSI/AEB 26 1 1 NRR/OS I/ASB 1 1 NRR/DS I'/CPB" 10 1 1 NRR/DSI/CSB 09 1 1 NRR/OS I/ICSB 1 1 NRR/DS I'/METB 12 1 1 I'/PSB 19 16'RR/DS 1 1 N 8 22 1 1 NRR/DSI/RSB'3 1 ~ 1 EG FILE 04 1 1 RGN5 31 "3 R IB 1 0 EXTERNAL; ACRS 41 6 6 BNL(AMOTS ONLY) 1 1.

DMB/OSS (AMDTS) 1 1 FEMA~REP DIV 39 1 l.

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Arizona Public Service Company Director of Nuclear Reactor Regulation November 5, 1984 Attention: Mr. George Knighton, Chief ANPP-31053 EEVB/WFQ Licensing Branch No. 3 Di vi si on of Li censing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:
Palo Verde Nuclear Generating Station (PVNGS)
Units 1, 2 and 3 Docket Nos. STN-50-528/529/530 Revised PVNGS FSAR Pages File: 84-056-026; G.l.01.10
Dear Mr. Knighton:
Attached for your review are revised PYNGS FSAR pages reflecting changes due to field testing results and in support of the PVNGS technical specification review. This material was previously discussed with the NRC in a general manner at our meeting with members of the NRC staff on October 4, 1984.
Please contact me if you have any questions.
Very truly yours E. E. Van Brunt, Jr.
APS Vice President Nuclear Production ANPP Project Director EEVBJr/WFQ/no Attachment cc: E. A. Licitra w/a A. C. Gehr w/a R. P. Zimmerman w/a 841 OOX3 8~<<o5 PDR Aoocc~ Ogooog28 pDR A
4 I r
Noverober 5, 1984 ANPP-31053 STATE OF ARIZONA )
) san COUNTY OF MARICOPA)
I, Edwin E. Van Brunt, Jr., represent that I am Vice President, Nuclear Production of Arizona Public Service Company, that the foregoing document has been signed by me on behalf of Arizona Public Service Company with full authority to do so, that I have read such document and know its contents, and that to the best of my knowledge and belief, the statements made therein are true.
~XXX Edwin E. Van Brunt, Jr.
Sworn to before me this ~64 day of , 1984.
Notary Public My Commission Expires:
Xy Commission Explrea April 6, 1987
2 (r
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P t.
PVNGS FSAR STANDARD DES IGNS conditions to confirm system design in case of RCS rupture, and to prevent flow induced movement of unrestrained reactor vessel internals. By draining the RCS to the mid point of the cold legs prior to the test and by providing a pressure vent path through the removed pressurizer safety valves, and addi-tionally, in the case of the LPSI test, aligning the LPSI pump not under test to take a suction from the RCS and discharge to the Refueling Water Tank (RWT), sufficient room exists in the RCS to obtain the required test data. By ensuring a test back pressure less than or equal to the head of water at a level equal to the reactor vessel flange, as indicated by the refuel-ing level indicator, the intent of the original test prerequi-site is maintained. If insufficient test data can be obtained because of some unforeseen reason with the reactor vessel head and internals installed, portions of the test will be repeated with the reactor vessel head and internals removed, as needed to meet the acceptance criteria of the CESSAR test descriptions.
MA lhl 4AIP ,5,/.9.4,"I, 5;I.V,z. 9.n., S. 3. l.Z/
1.9.2.4.10 <Auxiliary Feedwater System (CESSAR Section 5.1.4.G.6$
FSAR SeCtiOn 5. 1.5.G.6+5'.l.5;Ci 7~5 .< 5 Z. < +)< > I,'i Q CESSAR Section 5.1.4.G.6 requires delivery of auxiliary feed-water to the steam generator(s) within 10 seconds of initiatio of an AFAS signal when normal offsite or normal onsite power is available, and assuming a single active failure. The PVNGS auxil'iary feedwater system is discussed in section 10.4.9. If a single failure occurs in the safety-related motor driven auxiliary feedwater pump, the interface requirements must be met by the steam turbine driven pump. Quick start tests on turbine driven auxiliary feedwater pump show a tim'e for the'team the pump to reach rated flow and head is approximately eleven seconds. Combustion Engineering has determined that auxiliary feedwater flow initiation 20 seconds after AFAS meets CESSAR Chapter 15 accident analysis criteria. Therefore, the existing design is acceptable.
HI )~ L<<~'3 Amendment 12 1.9-14 February 1984
The PVNGS design takes exception to four CESSAR interface requirements with respect to the main and auxiliary feedwater systems. These exceptions are:
l. An increase in the feedwater isolation valve closure time (in both the downcomer and economizer lines) from 4.6 to 9.6 seconds (CESSAR Section 5.1.4.I. .a),
2. A reduction in the auxiliary feedwater flowrate from 875 to 750 gpm per pump (CESSAR Section 5.1.4.G.7),
3. An increase in the auxiliary feedwater pump start times when normal a-c is available from 10 seconds to 22 and 29 seconds for the motor and turbine driven pumps, respectively (CESSAR Sec-eCt tion 5.1.4.G.6), and h
4. An increase in the time delay from 15 to 23 seconds in which interrupted auxiliary feedwater flow must be fully reestablished to the steam generators (CESSAR Section 8.3.1.3.d).
I Table 1. -4 provides a matrix which describes how these exceptions impact the Chapter 15 safety and Chapter 6 containment analyses. This table demonstrates that the consequences of all of these analyses remain acceptable.
Table 1.9-4 Impact of Hain and Auxiliary Feedwater Deviations from the CESSAR 'Interface Requirements (Sheet 1 of 3)
AFW FWIV Increased AFW AFW Increase in FSAR Closure Flow Flow Re-establish Section Time after Reduction Delay with Delay When. Overall (A) Event HSIS 875 to 750 gpm Normal a-c a-c Is Lost Impact 15.1.1 Decrease in feedwater tempEthtuCF See event 15.1.4 below 15.1.2 Increase in feedwater flow See event 15.1.4 below 15.1.3 IncreasPmain steam flow See event 15.1.4 below 15.1.4 Inadvertent opening ADV No impact No impact (G) None (B) 15.1.5 Steam line break Event re-analyzed see section 15.1.5 Acceptable 15.2.1 Ioss of external 'load See event 15.2:3 below 15.2.2 Turbine trip See event 15.2.3 below 15.2.3 Loss of condenser vacuum Signal not No impact (G) None (D) encountered 15.2.4 HSIV closure See event 15.2.3 above 15.2.5 Steam pressure regulator failure Not applicable 15.2.6. Loss of a-c power See event 15.3.1 below 15.2.7 Loss of normal feed flow See event 15.2.3 above 15.2.8 Feedwater line break Event reanalyzed, see section 15.2.8 Acceptable 15.3.1 Loss of reactor coolant flow Signal not System not actuated None encountered 15.3.2 Flow controller malfunction Not applicable 15.3.3 Single reactor coolant pump Signal not No impact Loss of a-c assumed - no impact None seizure encountered 15.3.4 RCP shaft break See event 15.3.3 above 15.4.1 Low power CEA withdrawal Signal not System not actuated None encountered 15.4.2 Full power CEA withdrawal Signal not System not actuated None encountered 15.4.3 CEA assembly drop Signal not System not actuated None encountered
I Table 1.9-4 Impact of Main and Auxiliary Feedwater Deviations (Continued) from the CESSAR Interface Requirements (Sheet Q of $ )
AFW FWIV Increased AFW AFW Increase in FSAR Closure Flow Flow Re-establish Section Time after Reduction Delay with Delay When Overall (A) Event MSIS 875 to 750 gpm Normal a-c a-c Is Lost Impact 15.4.4 Startup inactive RCP Signal not System not actuated None encountered 15.4.5 Flow controller malfunction Not applicable 15.4.6 Inadvertent deboration Signal not System not actuated None encountered 15.4.7 Inadvertent fuel loading Signal not System not actuated None encountered 15.4.8 CEA ejection Signal not System not actuated None encountered 15.5.1 Inadvertent ECCS operation Signal not, System not actuated None encountered 15.5.2 CVCS malfunction Signal not System not actuated None encountered 15.6.1 Inadvertent opening PSV See section 15.6.5 below 15.6.2 Letdown line break Signal not System not actuated None encountered 15.6.3 Steam generator tube rupture Signal not No impact(G) . Loss of a-c assumed (H) None(G) encountered No impact 15.6.4 Outside ctmt main steam fail Not applicable (BWR) 15.6.5 Loss of coolant accident No impact(I) Loss of a-c assumed- None(I).
No impact SGTR with fully stuck ADV Event reanalyzed (J) Acceptable 6.2.1 Ctmt analysis, peak pressure Acceptable(K) No impact(L) Acceptable (K) 6.2.1.8 Ctmt analysis, peak tempsRArdRf Event reanalyzed in section 6.2.1.8 Acceptable A This column identifies the PVNGS FSAR Sections that reference the CESSAR safety analyses.
For the inadvertent opening of an ADV transient, an increase in the feedwater isolation valve closure time from 4.6 to 9.6 seconds does not alter the minimum DNBR or the maximum RCS pressure of the event. Therefore with respect to the Standard Review Plan (SRP) criteria, there is no impact on the consequences of the event.
The AFAS t v.qgc.E (. f-~ gg4EEv p oF3 interface requirement for auxiliary feedwater delivery when a-c power is not available is 45 seconds following the generation of an signal, and, when a-c power is available, 22 seconds for the motor driven pump. Therefore, if flow reestablishment following the loss of a-c occurs in less than the difference between these two times, i.e., 23 seconds, then it is assured that the total delay is less than the 45 seconds assumed in the Chapter 15 safety analyses.
For the loss of condenser vacuum transient, the maximum RCS pressure occurs before the delivery of auxiliary feedwater. Therefore with respect to SRP criteria, there will be no impact on the consequences of the event.
The maximum RCS pressure of the feedwater line break transient is unaffected by the reduction in auxiliary feedwater flow because the pressure occurs before auxiliary feedwater delivery. However, this transient was reanalyzed in section 15.2.8 in order to demonstrate that 750 gpm is adequate for long term RCS heat removal. \
For the locked rotor and sheared shaft transients, the minimum DNBR occurs before auxiliary feedwater delivery, and is not changed.
In addition, the integrated atmospheric steam releases and therefore the radiological consequences of the event remain unchanged by the reduction in auxiliary feedwater flow. Therefore, with respect to the SRP criteria, there is no impact on the consequences of the event.
The steam generator tube rupture (SGTR) transient presented in CESSAR Section 15.6.3 is not impacted by the reduction in auxiliary feedwater because the integrated atmospheric steam releases, and therefore the radiological consequences of the event remain unchanged. However t the SGTR transient with a fully stuck open ADV is sensitive to this reduction in auxiliary feedwater flow.
This transient was reanalyzed using the PVNGS specific auxiliary feedwater system in response to NRC question(5EE /Vor% PC 4')
For the steam generator tube rupture with a-c available, the auxiliary feedwater system is not ac~ated. Therefore these changes do not impact the transients results and the tube rupture discussion is limited to the case with the loss of a-c power.
auxiliary feedwater flowrate of 750 not alter the reported results of the emergency core cooling system (ECCS)
The minimum performance analysis referenced in Section 15.6 '.
gpm does Therefore conformance to the 10CFR50.46 ECCS acceptance criteria is preserved.
The large break LOCA and long term cooling evaluations in the CESSAR-F Sections 6.3.3.2 and 6.3.3.4're unaffected by the reduced minimum aux. feed flowrate. The energy removal capability of large breaks overwhelms steam generator heat transfer and aux. feed flowrate considerations.~ost-LOCA RCS heat removal in the long term cooling mode (after 1 hr.) requires a much lower auxiliary feedwater flowrate to remove the diminished decay heat values which occur at this later time.
The small break LOCA evaluation in CESSAR Section 6.3.3.3 is sensitive to steam generator heat removal and hence auxiliary feedwater flowrate. However, an evaluation of the auxiliary feedwater flowrate og 750 gpm shows that this flowrate is sufficient to preserve he RCS depressurization core liquid inventory and hence, the peak c188ding temperature results presented in CESSAR.
This reanalysis was provided in response to NRC questions on the steam generator tube rupture transient via letter ANPP-30572 from E.E. Van Brunt to G. W. Knighton dated September 19, 1984.
pressure of a steam line. break transient will be increased slightly (pepsi) as a result of an increase in containment Th e pea k con the feedwater isolation valve closure time from 4.6 to 9.6 seconds. The peak containment pressure of a steam line break will s i be bounded by the loss of coolant accident, which is unaffected by the change.
w'till These changes to the interface requirements tend to reduce the quantity of auxiliary feed flow to the generators. As a result, the containment response to the transient will improve as the mass and energy releases will be reduced.
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FVNGS FSAR STANDARD DESIGNS The automatic closure of the isolation valves by the interlock signals is to isolate the SCS from the RCS on rising RCS pressure. The interlocks do not provide SCS overpressuriza-tion protection. The SCS relief valves which are aligned to the RCS during SCS operation provide this protection. The operator will have adequate indication to initiate manual closure of the isolation valves should automatic closure by the interlock signals be precluded by a common mode failure of both transmitters.
Amendment 13 1.9-16 August 1984
1.9.2.4. 12 Containment Sump Isolation Valves Actuation Time Acceptance Criteria (CESSAR Section 6.2.4/FSAR Section 6.2.4)
CESSAR Table 6.2.4-1 indicates an opening time of 20 seconds for the following valves which are actuated by a recirculation actuation signal (RAS): SIA-UV673, SIA-UV674, SIB-UV675, and SIB-UV676. The recorded actuation times for these valves during valve stroke testing exceeded the acceptance. criteria of 20 seconds (as indicated in CESSAR Table 6.2.4-1). An increased actuation time of 25 seconds as indicated on FSAR Table 6.2.4-2 was reviewed against the safety analysis requirements for these valves and found to be acceptable. These valves are required to open upon receipt of an RAS. The RAS is generated by a low refueling water tank (RWT) level signal. These valves are required to open before the RWT level reaches the level of the safety injection (SI) pump suction line to ensure an uninterrupted source of water for the SI pumps. Sufficient margin is available in the low RWT setpoint to allow for a five 4$~second increase in the valve open time. The increased actuation time limit will not-affect these valves operability or durabilty. This deviation from these valves'riginal specification is acceptable. Any mechanical failure of a valve or its operator which would not cause the valves'ctuation time to exceed 25 seconds would not jeopardize the function of the valve.
PVNGS FSAR REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS generator(s) for residual heat removal and be capable of the following:
a. Maintaining the NSSS at hot standby with or without normal offsite and normal onsite power available.
b. Facilitating NSSS cooldown at a maximum administratively controlled rate of 75F/h from hot,standby to shutdown cooling initia-tion with or without normal offsite or onsite power available. (The shutdown P~AK cooling system becomes available for plant gA6 cooldown when the RCS temperature and pressure are reduced to approximately 350F E'.
and 400 psia.)
Refer to section 10.4 for a description of the system design.
will deliver The AFS flow to the steam generator(s) automat'call upon receipt of an AFAS as follows:
a. < P'ithin g8- secon s when normal offsite or normal <Pnsite power is available. The deviation from the CESSAR requirement of 10 seconds is acceptable to Combustion Engineering as, discussed in sec-
b. <
Within 45 seconds when normal onsite and normal offsite power are not available
7. Each of the safety-related auxiliary feedwater pumps is capable of delivering gal/min to the intact steam generator downcomer nozzle.
8. The auxiliary feedwater temperature will be no less than 40F and no greater than 180F.
0 Amendment 12 5.1-36 February 1984
PVNGS FSAR REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 3 ~ Capability for monitoring each MSIV, MFIV, ADV, and blowdown line isolation valve position is provided locally and in the control room.
4 ~ The accuracy of the feedwater temperature measure-ment devices is +1F for any calorimetric measurement.
I. Operational/Controls
1. A power-operated MSIV capable of establishing shutoff under conditions of design pressure, design temperature, and flow conditions resulting from a break upstream or downstream is provided in each main steam line outside of containment.
Refer to section 10.3.2.
2. Each MSIV position is monitored and controlled locally and in the control room.
3. An MSIS closes the MSIV bypass valves and MSIVs.
4 ~ The full "open to close" stroke time of each MSIV and MSIV bypass valve is +seconds or less upon receipt of an MSIS.
5. The ADVs fail closed and are capable of being remote manually positioned to control the plant cooldown rate,.
6. The ADVs are provided with remote manual operators such that the valves can be operated from the control room and remote shutdown panel in the event of a loss of normal power supply.
7. In the combined event of either a steam line break or steam generator tube rupture and the loss of power operation of the atmospheric dump valves, personnel access to the manual operators of the intact valves on the other steam generator is possible.
5.1-38
PVNGS FSAR-REACTOR COOLANT SYSTEM AND
~, CONNECTED- SYSTEMS
8. A MSIS actuatiori signal will close the MSIVs, MSIV bypass valves, MFIVs and the steam generator blowdown valves.
9. Redundant feedwater system isolation valving is provided in both the economizer feedlines and the downcomer feedlines such that the following P/<~
criteria are met when the effects of single PA>~
failure criteria are imposed:
a. Complete, termination of forward feedwater flow is assumed within'wd seconds after receipt of an Msis. 'rfh.c4 ~~~c~~~
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b. Abrupt complete termination o reverse feed-water flow with the existance of a reverse flow condition. Check valves are considered to be an acceptable means of achieving the above.
10. The economizer and downcomer feedwater line isolation valves (MFIVs) in each main feedwater line are remote-operated and capable of maintain-ing a leak rate of less than 1000 cc/h under the main feedwater line pressure, temperature, and flow resulting from the transient conditions I
associated with a pipe break on either side of the valves.
11. The safety-related AFS can be controlled from either the control room or remote shutdown station as described in section 7.3.1.
12. The AFS is controllable such that post-accident operation will not result in overfilling the intact steam generator(s).
13. The auxiliary feedwater pumps of the AFS are designed for operation when steam generator 5.1-39
PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-4 MASS AND ENERGY RELEASE FOR CONTAINMENT PEAK PRESSURE/TEMPERATURE ANALYSIS (Sheet 1 of 4)
A. Most severe hot leg break Break type: Double-ended hot leg slot break Pipe ID, 42 in.
Break area 19.24 ft2 Energy Mass Flow Reactor Release Rate Vessel Accident Time Rate Million Enthalpy Pressure Phase (s) (ibm/s) Btu/s (Btu/ibm) psia Blowdown Refer to CESSAR Table 6.2.1-10
a. 70 psia containment backpressure case
b. Recirculation begins c ~ Safety injection realigned; 50% to hot legs and 50% to cold legs. Excess hot leg safety injection water carries significant decay and sensible heat to containment sump.
d. RCS sensible heat addition to containment completed.
e. Includes direct floor spillage for disc rge leg break.
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was)(energy rc.l'cRJc J feb lAscld Fol c[l pA a*+
~4'4'c klo p~ t f. pea- W ~ w ~/gal Pn s'echo',(,B.
6.2.1-13
PVNGS F SAR CONTAINMENT SYSTEMS Table 6.2.1-4 MASS AND ENERGY RELEASE FOR CONTAINMENT PEAK PRESSURE/TEMPERATURE ANALYSIS (Sheet 4 of 4)
D. Most severe secondary system break (pressure)
Break type: Main steam line slot break Pipe ID: 28 in.
Break area: 4.00 ft2 Reactor power level: .0%,
Mass Release Energy Release Time Rate Enthalpy 'ate (s) (ibm/s) (Btu/ibm) (million Btu/s)
Refer to CESSAR Table 6.2.1-19 E. Most severe secondary system break (temperature)
Break type:. Main steam line slot break Pipe ID: 28 in.
Break area: 8.78 ft2 Reactor power level: 102%
Mass Release Energy Release Time Rate Enthalpy Rate (s) (ibm/s) (Btu/ibm) (million Btu/s)
Refer to CESSAR Table 6.2.1-1 6."2.1-16
0 PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-7 ENGTNEERED SAFETY FEATURE SYSTEMS OPERATING ASSUMPTIONS FOR CONTAINMENT PEAK PRESSURE ANALYSIS (Sheet 1 of 2)
Value Used for Full Peak Pressure System/Item Capacity Analyses Passive Safety Injection System A Number of safety injection tanks 4 4 Pressure setpoint, psig ; 600 600 Volume, ft /tank 3
1927 1927 Active Safety Injection Systems High-presyuqe safety injection~
Number of lines 2 2 Number of pumps 2 1/2(b)
Flowrate, gal/min/pump 1130 1130 Low-pressure safety injection<<)
Number of lines 2 2 Number of pumps 2 1/2(b)
Flowrate, gal/min/pump 5000 5000 Containment Spray System Number of lines i 2 1 Number of pumps 1 Number of headers 1 Flowrate, gal/min/pump Heat Exchangers Shutdown heat exchangers Type Shell Shell and and U-Tube U-Tube Number
a. From CESSAR Section 6.3
b. 1 = minimum ECCS; 2 = maximum ECCS
{'. NET Flail gugw/( ltvtEdT)os PH45E '4PM I X'5 g$ "$ 0 4fhh PoLLP4/l~g 0Lo5uRG oP- A~~<+4P< 8~1 45S gp{ VE'rv RAS.
6.2.1-19
0 PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-7 ENGINEERED SAFETY FEATURE SYSTEMS OPERATING ASSUMPTIONS FOR CONTAINMENT PEAK PRESSURE ANALYSIS (Sheet 2 of- 2)
Value Used for Full Peak Pressure System/Item Capacity Analyses Heat transfer area, ft 10,840 10,840 Overall heat transfer> 372 285 coefficient, Btu/h-ft - F Flowrates:
Recirculation side, gal/min 3890 3890 '
Exterior side, gal/min 14,400 14,400'ssential Source of cooling water Essential COL>NQ coolin'g
~are,R water At rated flow seconds ]M5E<1 (Offsite Power Available/
Loss of Offsite Power) g v.n f'C ,z II'V)..
Loss of coolant accident . 77/87 77/87 Main steam line break 80/90 80/90 accident both maximum and minimum ECCS performances were evalu-ated. For the containment heat removal systems, mini-mum system capacity is conservative for calculating containment peak pressures.
Passive heat sink data is provided in table 6.2.1-8.
Part A of the heat sink table is a detailed list of the geometry of each heat sink and part B describes the resulting simplified heat sink'odels used for computer input. Node spacing used for concrete, steel, and steel-lined concrete heat sinks is fine enough to ensure an accurate representation of the thermal gradient in each slab. In concrete, the node spacing varies from 0.003274 to 0.004233 feet depending on the 6.2.1-20
PVNGS FSAR CONTAINMENT SYSTEMS 6.2.1.1.3.3 Accident Chronolo . Accident chronologies for the most severe reactor coolant system breaks and MSLBs are provided in table 6.2.1-10. It is assumed that time equals zero at the start of each accident. K~SGR i A 6.2.1.1.3.4 Ener Balance. For the most severe reactor coolant system pipe breaks and the most severe secondary coolant system pipe break, a detailed energy balance was per-formed to show the distribution of energy prior to the accident, at the time of peak pressure, at the end of the blowdown phase, at the end of the core reflood phase (for LOCA), and 1 day (8.64 x 10 s) after the accident (for LOCA). This information is presented in table 6.2.1-11.
6.2.1.1.3.5 Functional Ca abilit of Containment Normal Ventilation S stems. Containment maximum and minimum design pressures are based on conservative assumptions of initial atmospheric pressures and temperatures within the containment.
The functional capability of the containment normal ventilation systems to maintain initial containment atmospheric conditions within the range of temperature and pressure defined for normal plant operation is discussed in section 9.4. Section 16.3/4.6 of the Technical Specifications gives the limiting conditions of containment temperature and pressure for normal plant opera-tion and describes the action that will be taken if these con-ditions are exceeded.
6.2.1.1.3.6 Protection A ainst Severe External Loadin . The design basis accident for containment external pressure design has been determined to be inadvertent actuation of the contain-ment spray system. Consideration was also given to misopera-tion of the containment normal purging system (i.e., operation 0
6.2.1-34
Insert A to Page 6.2.1-34 The accident chronology for the PVNGS-specific 102% power MSL slot break used for the in-containment equipment qualification peak temperature
-k8 analysis of section 6.2.1.8 is provided in table 6.2.1~
PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-10 ACCIDENT CHRONOLOGIES (Sheet 4 of 5)
D. Worst case main steam line break (pressure)
Break type: 0% power MSL slot break Loss of cooling train Time (s) Event 0.0 Break occurs 4.55 Reactor trip signal 4.55 Main steam isolation signal 4.55 Main feedwater isolation signal 5.45 Turbine admission valve closed 5.45 Reactor trip begins 5.45 Main steam isolation valves start to close 5.45 Main feedwater isolation valves start to close Containment .spray actuation signal (10 psig containment pressure)
10. 45 Main steam isolation valves closed 10.45 Main feedwater isolation valves closed 28 Containment spray pump at full speed pT FLILL, 80 Containment, spray<flow initiated inside containment building 80 Peak containment temperature of 393F occurs 194 Peak containment pressure of 42.8 psig occurs 210 Blowdown ends 6.2.1-38
0 Insert I 4.55 Containment pressure reaches reactor trip analysis setpoint of 6 psig 4 '5 Containment pressure reaches Main Steam Isolation Signal (MSIS) analysis setpoint of 6 psig 5.55 High containment pressure reactor trip signal and MSIS generated 5.70 Turbine admission valves closed 5.70 Reactor trip breakers open Containment pressure reaches Containment Spray Actuation Signal (CSAS) analysis setpoint of 10 psig CSAS generated 6.04 Rods start to drop 10.45 Main steam isolation valves closed 10.45 Main feedwater isolation valves closed (see Section 1.9.2.4.10)
PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-10 ACCIDENT CHRONOLOGIES (Sheet 5 of 5)
E. Worst case main steam line break (temperature)
Break type: 102% power MSL slot break Loss of cooling train Time (s) Event 0.0 Break occurs 3.8 Reactor trip signal Main steam isolation signal Main feedwater isolation signal 4.7 Turbine admission valve closed Reactor trip begins Main steam isolation valves start to close Main feedwater isolation valves start to close 5.0 Containment spray actuation signal (10 psig containment pressure) 9.7 Main steam isolation valves closed Main feedwater isolation valves closed 28 Containment spray pump at full speed AT FOLL.
80 Containment spray flow initiated inside containment building Peak containment temperature of 401F occurs 150 Peak containment pressure of 41.1 psig occurs 170 Blowdown ends 6.2.1-39
Insert A (to Table 6.2.1-10 sh. 5 of 5) 3.80 Containment pressure reaches reactor trip analysis setpoint of 6 psig 3.80 Containment pressure reaches Main Steam Isolation Signal (MSIS) analysis setpoint of 6 psig 4.80 High containment pressure reactor trip signal and MSIS generated 4.95 Turbine admission valves closed
~ 4 95 Reactor trip breakers open 5.0 Containment pressure reaches Containment Spray Actuation Signal (CSAS) analysis setpoint of 10 psig.
6.0 CSAS generated S
'.29 Rods start to drop 9.7 Main steam isolation valves closed 9.7 Main feedwater isolation valves closed (see Section 1.9.2.4.10)
PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-25 MISCELLANEOUS AND ADDITIONAL ENERGY RELEASES Time Energy (sec) (Btu/h) 500 6
500 20.8956 E6 86,400 20.8956 E6 86,400 lE6
a. All miscellaneous energy added after blowdown ends (500 sec), up to 1 day (86,400 sec).
1 6.2.1.4
~ ~ Mass and Ener Release Anal sis for Postulated Secondar S stem Pi e Ru tures Inside Containment Refer to CESSAR Section 6.2.1.4. The MSLBs analyzed for con-tainment pressure/temperature were the slot breaks listed in CESSAR Tables 6.2.1-11 through 6.2.1-20. The guillotine ruptures for the same size break and power levels were evaluated as being less severe than the comparable slot breaks.
6.2.1.5 Minimum Containment Pressure Anal sis for Performance Ca abilit Studies on Emer enc 'Core Coolin S stem 6.2.1.5.1 Introduction and Summary Appendix K to 10CFR50 provides the required and acceptable fea-(6) tures of emergency core cooling system (ECCS) evaluation models.
Included in this list is the requirement that the containment March 1982 6 '.1-96I Amendment 8 f6
Insert A to Page 6.2.1-96I In addition, PVNGS-specific mass/energy release data for the 102$ power MSL slot break with a 9.6 second feedwater isolation valve closure time used for the in-containment equipment qualification peak temperature analysis are provided in table 6.2.1-27.
PVNGS FSAR CONTAINMENT SYSTEMS be at the maximum of their uncertainty ranges, and their thermal properties (conductivity and heat capacity) are also maximized.
6.2.1.5.3.7 Heat Transfer to Passive Heat Sinks. Refer to CESSAR Section 6.2.1.5.3.7.
6.2.1.5.3.8 Containment Pur e S stem. The analysis presented in this section has been performed including the effects of the eight inch power access purge system. The power access purge system is assumed to be operating at the time of the postulated LOCA. The purge system isolation valves are fully closed 8'+ . seconds after a containment isolation actuation signal generated by high containment pressure (5.0 psig). It is assumed that only dry air is removed from the containment atmosphere through the purge system. This conservatively minimized the calculated containment pressure.
6.2.1.5.4 Results For the limiting large break LOCA, 1.0 x DEG/PD, the minimum containment pressure response for Palo Verde is not more limiting than the minimum containment'pressure response used in analyzing the performance of the ECCS. A comparison of the Palo Verde and CESSAR containment pressure response is shown in figure 6.2.1-22. The blowdown peak containment pressure for Palo Verde is lower than the CESSAR blowdown peak containment pressure. However, during the blowdown phase of the loss-of-coolant accident - approximately the first twenty-five seconds - the flow from the break is choked and therefore RCS hydraulic behavior is insensitive to the containment pressure response. During the reflood phase, the break flow is not choked and therefore the RCS hydraulic behavior is sensitive to the containment pressure, the Palo Verde pressure response exceeds CESSAR response.
April 1983 6.2.1-960 Amendment 11
0 PVNGS FSAR p~~R,7 ~A CONTAINMENT SYSTEMS loss of offsite power.
In-containment safety-related equipment required to operate post-LOCA is qualified to the LOCA design basis accident environ-ment as specified in table 3E-1 (sheet 1 of 7). This environ-ment bounds the calculated pressure-time and temperature-time response of figure 6.2.1-3 which shows the worst case LOCA transient as discussed in section 6.2,1.1.3,1.D.
6.2.1.9 References Shoenhoff, H. M. and Braddy, R. W., "Containment and Safety-Related Equipment Transient Temperature Analysis Following a Main Steam Line Break," Bechtel Power Corporation, May 1975.
2. BN-TOP-4, Rev. 1, Oct. 1977, Subcompartment Pressure and Temperature Transient Analysis, Bechtel Power Corporation, San Francisco, CA.
3. Idel'chik, I.E., Handbook of Hydraulic Resistance (AEC-TR-6630), 1966.
ASHRAE Handbook of Fundamentals, 1972.
5. Interim Staff Position of Environmental Qualification of Safety-Related Electrical Equipment (NUREG-0588),
December 1979.
6. Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Cooled Nuclear Power Reactors, Federal Re ister, Vol. 39, No. 3 - Friday, January 4, 1974.
7. "Calculative Methods for the C-E Large Break LOCA Evaluation Model", CENPD-132, August, 1974 (Proprietary).
"Updated Calculative Methods for the C-E Large Break LOCA Evaluation Model" CENPD-132, Supplement 1, February 1975 (Proprietary).
Amendment 8 6.2.1-98 March 1982
Insert A to Page 6.2.1-98 It differs from the analysis of the Main Steam Line break of Section 6.2.1.1.3.1 in that the mass/energy release data provided in table 6.2.1-27 are based on a blowdown model that includes the PVNGS-specific containment volume and main steam piping configuration and includes the PVNGS-specific feedwater isolation valve closure time of 9.6 seconds. Furthermore, the analysis also assumes the 6-inch steam supply line from the failed steam generator to the auxiliary feedwater pump turbine remains open with a stuck open check valve allowing crossflow of steam from the intact steam generator to the failed generator until operator action AT'team 30 minutes terminates the flow.
Finally, the equipment qualification analysis utilizes 8g, condensate re-evaporation as allowed by Appendix B of NUREG-0588 (5) . A chronology of events for this main steam line break analysis is provided in table 6.2.1-28.
The combined effects of the PVNGS-specific modeling of the 102/ MSL slot break compared with the previous equipment qualification peak containment temperature analysis based on the CESSAR mass/energy release data for the same break used in Section 6.2.1.1.3.1 is a 10 F reduction in peak contain-ment vapor temperature (359.6 vs 369.8 F) and a O.l psig increase in peak containment pressure (41.2 vs 41.1 psig).
Table 6.2.1-27 Mass and Energy Release for Containment Equipment Qualification Peak Temperature Analysis Break type: Main steam line slot break Pipe ID: 28 in.
Break area: 8.78 ft Reactor power level: 102'/
Isolation valve closure times: 4.6 sec MSIV 9.6 sec 8~ MFIV Stuck open CV on supply line to AF pump turbine Mass Release Energy Release Time Rate Enthalpy Rate (s) (ibm/s) (Btu/ibm) (Million Btu/s) 0.0 9833.75 1190.04 11.702518 0.2 7303.17 1190.69 8.695786 0.6 7039.74 1191.87 8.390426 1.0 6849.83 1192.71 8.169867 2.0 6524.69 1194.18 7.791653 3.0 6301.20 1195.21 7.531249 4.0 6127.74 1195.99 7.328734 5.0 gogo.9 1195.40 5.571987 6.0 4759.56 1195.02 5.687764 7.0 4843.94 1194.57 5.786444 8.0 4920.50 1194.13 5.875717 9.0 4988.01 1193.72 5.954272 9.5 5016.57 1193.54 5.987483 10.0, 2853.53 1192.92 3.404022 10.5 2864.79 1192.77 3.417024 11.0 2870.74 1192.67 3.423858 11.5 2871.49 1192.65 3.424678 12.0 2867.25 1192.68 3.419704 12.5 2858.25 1192.77 3.409235 13.0 2844.85 1192.91 3.393662 15.0 2767.35 1193.77 3.303575 20.0 2559.88 1196.08 3.061811 25.0 2445.21 1197.29 2.927615 30.0 2374.74 1198.00 2.844945 35 ' 2284.15 1198.90 2.738457 40 ' 2206.28 1199.64 2.646736 45.0 2142.79 1200.22 2.571811 50.0 2075.62 1200.80 2.492405 60.0 1948.28 1201.84 2.341530
Table 6.2.1-27 (Continued)
Mass Release Energy Release, Time Rate Enthalpy Rate (s) (ibm/s) (Btu/ibm) (Million Btu/s) 70.0 1828.22 1202.70 2.198796 80.0 1720.88 1203.34 2.070804 tfglO o 1203.87 1.938305 90.0 100.0 1492.90 1204.24 1.797816 110.0 1361.07 1204.41 1.639290 120.0 1249.51 1204.36 1.504854 .
130.0 1177.06 1204.22 1.417440 140.0 1044.42 1203.85 1.305478 150.0 970.36 1203.09 1.167428 155.0 903.62 1202.46 1.086565 160.0 830.21 1201.57 0.997553 165.0 751.09 1200.37 0.901587 170.0 676.99 1198.94 0.811673 180.0 512.9 1200.4 0.615700 185.0 380.6 1227.3 0.467100 190.0 261.4 1239.5 0.324000 194.5 131.1 1240.3 0.162600 1800.0 131.1 1240.3 0.162600 1800.1 0 0 ntegral: 539924 ibm 656.434 Million Btu (a) Operator action to terminate steam crossflow
Table 6.2.1-28 ACCIDENT CHRONOLOGY FOR CONTAIR1ENT EQUIPtKNT QUALIFICATION PEAK TEMPERATURE ANALYSIS Break type: 102'wer MSL slot brea stuck o en 6-inch c eck valv loss of one cooling train Time (s) Event 0.0 Break occurs 3.0 Containment pressure reaches Safety Injection Actuation Signal (SIAS) analysis setpoint (5 psig) 3' Containment pressure reaches reactor trip and Main(S)earn Isolation Signal (MSIS) analysis setpoint (6 psig) 4.0 SIAS generated 4.6 Reactor trip signal and MSIS generated
(~)
4.75 Turbine admission valves 'closed (b) 4.75 Reactor trip breakers open (b) 5.09 Rods start to drop (b) 7.0 Containment pressure reaches Containment Spray Actuation Signal (CSAS) analysis setpoint (10 psig) 8.0 CSAS generated 8.0 Containment Spray (CS) valves start to open 9.5 Main steam isolation valves closed (b) 14.50 Main feedwater isolation valves closed (b) 18 CS valves fully opened 20 CS pump loaded on essential bus 25 CS pump at full speed 80 CS headers filled and full spray flow established
Table 6.2.1"28 (Continued)
Tl.me (s) Event 80 Peak containment vapor temperature of 359.6 F occurs 178 Peak 'containment pressure of 41.2 psig occurs 194.5 Failed steam generator dryout occurs 5l-8AM 1800 Operator action terminates auxiliary feed pump turbine line crossflow through stuck open check valve 1800 Blowdown into containment ends. Containment vapor temperature at 241~F. Containment pressure at 27.4 psig.
a. Value used in pipe break blowdown analysis.
b. Events based on time to reach reactor trip and MSIS setpoint of 3.6 seconds from pipe break blowdown analysis.
TMAX 3698'F 1.1 PS MAX U 300 40 O1'0 K
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z sc O 1 SPRAY TRAIN z 200 ON@80sec 20 2
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10 1000 TIME FOLLOWING BREAK Isec)
Palo Verde Nuclear Generating Station FSAR EQUIPMENT QUALIFICATION PRESSURE AND TEMPERATURE RESPONSE MSLB (SLOT) AT 102% POWER Figure 6.2.1-20 December 1980 Amendment 3
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Table 6.2.4-2 CONTAINMENT -ISOLATION SYSTEM (Sheet l of 9) Valve Position Pri- Secon-ESF(bi mary( dary(ai Pene- Actua- Actua- hctua- Closure tration Valve Valve tion tion Shut- Post- tion Time(c) Pone~ Number System Numbers Operat'or Mode Mode Normal down hccident Failur Signal (Sec) Sourra 1,2, 3,4 Main steam SGE-UV170 SGE-UV171 q'( hccumuiai SGE-UV180 Hydraulic 0 MSIS 1 SGE-UV181 SGE-PSV691 SGE-PSV692 None SGE-PSV694 Safety Nh C NA SGE-PSV695 SGE-PSV575 SGE-PSV576 None SGE-PSV557 Safety Nh C C NA NA SGE-PSV558 SGE-PSV574 SGE-PSV577 None SGE-PSV556 Safety Nh C NA SGE-PSV559 SGE-PSV573 SGE-PSV578 None Nh SGE-PSV555 Safety Nh C C C NA SGE-PSV560 Notes a. Position indications for remotely actuated valves are shown in the control room.
b. The parameters sensed and the values which generate actuation signals are given in CESSAR Section 7.3.

c. "Opens" means valve opens on actuation signal vice closes.

d. Valves are essential. Operator action reguired to open valve. I Symbols: N.h. - not applicable K - locked closed
- fail open EA Class IE bus h FO EB - Class IE bus B FC - fail closed EC - Class IE bus C FAI - fail-as-is ED - Class IE bus D MSIS - main steam isolation signal N - normal power source CSAS - containment spray actuation signal 0 - open CPIAS - containment purge isolation actuation signal C - closed AFAS - auxiliary feedwater actuation signal A - automatic SIAS - safety injection actuation signal R - remote operation RAS - recirculation actuation signal M - manual local operation, CIAS - containment isolation actuation signal DU - Data currently unavailable - bracket indicates any one signal actuates each valve
0 Table 6.2.4-2 CONTAINMENT ISOLATION SYSTEM (Sheet 2 of 9) Valve Position Pri- Secon- ) mary>> d dary>>a) ESF>> Pene- Actua- Actua- Actua- losure tration Valve Valve tion tion -Shut- Post- tion ime <<) PoMcr Number System Numbers Operator Node Mode Normal dovn Accident Failure Signal >>Sec) Source 1,2, Main SCE-PSV572 3' s team SGE-PSV579 Safety NA C None Nh NA SCE-PSV554 SGE-PSV561 1,2, Hain SGA-UV134 Motor C 0/C FAI AFAS Opens 3,4 steam SGA-UV138 SCA-HV184 SGB-))V178 None>> ) Opens Accum-Piston FC ulator <<" SGB-HV185 SGA-HV179 e 4. Eh&EB SGE-UV169 Piston C HSIS SGE-UV183 A & EB h & EB
. 't,4 1,2, Main SCE-UV1133 Solenoid 0 MSIS 3,4 steam SGE-UV1134 SCE UV1135h . R.C EB SGE-UV1135B EB solenoid 0 MSIS EB SGE-UV1136A SCE UV1136B Q ~ EB SGE-V603 Hand LC C C NA None NA Nh SGE-V611 Spare LC 0 C Nh None NA NA Deminer- DWE-V061 Hand M M None NA NA alized DME-V062 Hand M M LC 0 C Nh eater Nh None Nh Nh Fire FPE-V089 Hand M M LC C Nh None NA NA protec- FPE-V090 None h h C C tion 8,10 Feed SGB UV132 Hydraulic 0 C C NSIS Accumulator )
Mater SGB-UV137 SGA-UV174 SCA-UV177 Hydraulic 0 C NSIS )1;6 Accumulator .
Table 6.2.4-2 CONTAINMENT ISOLATION SYSTEM (Sheet 3 of 9) Valve Position Pri- Secoi maryt dary ) ESFib) Pene- Actua- Actua- Actua-tration Valve Valve tion tion Shot- Post- tion Time" ~ PoM.:i Number System Numbers Operator Mode Node Normal dovn Accident Failur Signal (Sec) Souse 8,10 Feed- SGE-V003 None NA None NA NA vdter SGE-V006 SGE-V007 None SGE-V005 None NA NA NA Radvaste RDA-UVC23 Motor R 0 FAI CIAS EA drain RDB-UV024 Diaphragm R C C CIAS EB RD-UV407 Solenoid R C FC CIAS EB 11,12 Feed- SGE-V652 None None NA NA vater SGE-V653 SGE-V642 0 None NA SGE-V693 None NA 11,12 Feed- SGB-UV130 EA and vater SGB-W135 Piston 0 MSI 5 EB SGB-HV200 Solenoid FC CIAS 5 SGA-UV172 EA and SGA-W175 Piston 0 HSIS EB SGB-HV201 Solenoid FC CIAS 13 HPSI SIE-V113 None h A 0 NA None NA NA SIB-UV616 Motor A R,N 0 FAI SIAS 10 EB SIA-UV617 Motor A R,M 0 FAI SIAS 10 EA 14 HPSI SIE-V123 None h C 0 NA None Nh NA SIB-UV626 Motor R,N C 0 FAI SIAS 10 EB SIA-UV627 Hotor R,N C 0 FAI 5 IAS 10 EA 15 HPS I SIE-V133 None h h 0 Nh None NA NA S I B-UV636 Motor A R,M 0 FAI SIAS 10 EB SIA-W637 Motor h R,M 0 FAI SIAS 10 EA 16 HPSI S IE-V143 None h h 0 NA None NA NA SIB-UV646 Motor A R,M 0 FAI SIAS 10 EB SIA-UV647 Hotor A R,M 0 FAI SIAS 10 EA
0 Table 6.2.4-2 CONTAINMENT ISOLATION SYSTEM (Sheet 4 of 9) Valve Position Pri- Seco~-) ESF(b) mary( dary Pene- Actua- Actua- Actua- Close~ tration Valve Valve tion tion Shut- Post- tion Time Power Number System Numbers Operator Mode Hode Normal dovn hccident Failur Signal (Sec) Soot 17 I PSI SIE-V114 None A 0 NA None NA HA SIB-UV615 Motor R,M 0 FAI ~ SIAS 10 " ll 18 LPSI S I E-V124 None A NA None NA HA S I B-UV625 Motor R,M FAI SIAS 10 HIS 19 I.PS I SIE-V134 None A NA None NA HA S I A-UV635 Motor R,M FAI SIAS 10 EA 20 LPSI SIE-V144 None A NA None Nh HA SIA-UV645 Motor R,M FAI SIAS 10 EA 21 CS SIA-V164 None C NA None Nh NA SIA-UV672 Motor LC FAI CSAS 10 EA 22 CS SIB-V165 None A C NA None Nh SIB-UV671 Motor A I,C FAI CSAS 10 23 SI S I A-UV673 Hotor A R,M FAI RAS g5'ga5 EA SIA-UV674 Hotor A R,M FAI RAS EA SIA-PSV151 Safety A NA C None Hh SIA-UV708 Solenoid A R FC CIAS 8 'l EA SI S IB-UV675 Motor R,M FAI RAS R5 EB SIB-UV676 Hotor R,M FAI RAS Z.5 EB SIB-PSV140 Safety NA C None HA 25A CB rad HCB-UV044 Solenoid '0OorC FC CIAS fZ. EB mon HCA-UV045 Solenoid or C FC CIAS EA
'f 25B CB rad HCB-UV047 Solenoid Oor C FC CIAS p EB mon HCA-UV046 Solenoid OorC FC CIAS l g.- EA 26 SDC SID-UV654 Motor R LC 0 OorC FAI None 80 ED (CESSAR SI 8-UV656 Motor M LC 0 OorC FAI None 80, EB
27) S IB-HV690 Motor H LC OorC OorC FAI None 30 EB.
SIB-PSV189 Safety HA C C C C None NA NA 27 SDC SIC-UV653 Motor R I.C 0 OorC FAI None 80 EC (CESSAR S I A-UV655 Motor H IC 0 OorC FAI Hone 80 EA
28) SIA-HV691 Motor H LC Oor C OorC FAI Hone 30 EA S I A-PSV179 Safety NA C C C C None NA NA
Table 6.2.4-2 CONTAINMENT XSOLATXON SYSTEM (Sheet 5 of 9) dary(aa) Valve Position trationn Pri- Secon-
,mary( ESF(b)
Pene- !Actua- Actua- ,'Actua Closure Valve Valve ltion tion Shut- Post- ltion gime(c) Powel' Number System Numbers Operator !Node Node Hormal down Accident Failure )Signal (Sec) Soul'a.'v I I < 28 SI S Ih-UV682 hir h I
' R C OorC C FC BIAS 5 Eh
's) I (CESSAR SIE-V463 None N LC OorC C NA None NA NA SIE-PSV474 Safety ! A I Nh C C C C None NA NA I 29 N2 CAE-VOI 5 None A ! A 0 ;-0 or C NA Hone NA
.'Solenoid j '0 GAA-UV002 A R 0 or C Fc CIAS EA I
30 N2 GAE-Vcll ,None
. A C Oor C NA None NA CAA-UV001 : Solenoid A C OorC FC EA I
31 Inst air IAE-V021 ! None h OorC C NA HA IAA UVG02 ~ 'Solenoid A I R i 0 ,Oor C C I Fc CIAS Eh I
~ -' l 32A CB press HCC-HV076 solenoid R ! 0 I 0 0 Hone l Opens Ec I l
mon I I l I I l 32B !Spare I I I 32C (Spare I I 33 Nuc NCE-V118 : I l CW NCB-UV401 ! None Motor h 0 0 0 0 C NA None NA Nh ~ l I FAI CIAS I
~
I l C EB ~ 1 34 Nuc CW NCB-UV403 'otor 1 I h 0 0 C FA I ! CIAS EB'A HCA-UV402 Motor 0 C FAI CIAS~ 5 g 35 CB hyd control HPA-UV001 HPA-UV003 HPA HV007h i Motor Motor Solenoid I A A R I k R C C C C C I l Oor C Oor C 0'or C FAI FAI FC . I CIAS None I
~
CI AS ~~ 1 5'g EA EA EA I ll 36 CB hyd HPB-UV002 Moto A C OorC FAI CIAS~ 5 '3 EB control IlHPB-UV004 lHPB-HV008h Motor Solenoid A R R R C C OorC OorC FAI FC I Clh~~ 5 f None ' I E6 EB
'I" 37A SG blow- ~~A I down -UV211 Solenoid h R 0 C C FC AFAS sample -UV22 8 Solenoid A R 0 C C FC SIAS I
aS OI CA
~
g VI
Table 6.2.4-2 CONTAINMENT ISOLATION SYSTEM (Sheet 6 of 9) Valve Position Pri- Seco maryi dary~a) ESF Pene- Actua- Actua- Actua- Closur)r tration Valve Valve tion tion Shut- Post- tion Time(c) Pouter Number System Numbers Operator Node Mode Normal dovn hccident Failure Signal (Sec) Sorlr
... ~ r 37B SG blou- SGA-UV204 Solenoid A C FC MSIS F.A doun SGB-UV219 solenoid AFAS A C FC sample SIAS 38 B hyd HPA-V002 None A h C OorC Nh None Nh 'NA control HPA-UV005 Notor A R C 0 or C FAI CIAS Eh HPA-HV0078 solenoid R R C OorC FC None Eh HPA-UV23 Solenoid A R C C FC ~
CI AS OorC 39 B hyd control HPB-V004 HPB-UV006 HPB HVOOSB No) re Motor Solenoid A A R C C C OorC Oor C Nh FAI FC None CIAS None NA 1
~ ))h EB EB 40 CVCS CHA UV516 hir 0 CIAS/ 5'I BIAS CHB-UV523 hir R,M 0 C FC CIAS EB CHB-VV924 solenoid R C C FC CIAS EB FV 41 Cs CHE-VM70 None A C Oor C OorC NA None NA Nh CHA-HV524 Notor R 0 0 0 FAI None Eh CHE-V854 Hand M C C C NA None NA NA 42h Sample SSA-UV204 Solenoid h C FC CIAS 5'7 SSB-UV201 Solenoid A C FC CIAS 42B sample SSA<<UV205 Solenoid A C FC CIAS SSB UV202 Solenoid h ~
C FC CIAS 42C ample SSA-UV203 Solenoid A FC CIAS Eh SSB-UV200 Solenoid A FC CIAS EB 43 VCS CHA-UV506 hir A R,M 0 OorC FC CIAS EA CHB-VV505 Air A R 0 OorC FC CIAS EB Cs CHA UV560 hir R OorC FC CIAS Eh CHB-UV561 Air R,N OorC FC CIAS EB Cs CHE-V494 None h A OorC C C NA None NA CHA UV580 Air A R,N OorC C C FC CIAS 5'l EA CH-UV715 solenoid A R C C C FC CIAS Sf EA 46 G blov- SGA-VVSOOP Diaphragm FC NSIS f.C EA down 'FC AFAS SGB-UVSOOQ Diaphragm SIAS .0 (p EB
Table 6.2.4-2 CONTAINMENT lSOLATlON SYSTEM (Sheet 7 of 9) Valve Position Pri- Secon-maryi dary <a) ESF~ Pene- Actua- Actua- Actua Closure 1 tration Valve Valve tion tion Shut- Post- tion 7'ime (c) Power Number System Numbers Operator Node Node Normal down hccident Failure Signa (Sec) Sourc . 47 SG blow- SGB-UVSOOR Diaphragm "A 0 FC 7i(r EB down 0 SGA-UVSOOS Diaphragm A FC SG blow- SGB-UV226 Solenoid 0 C C FC EB down sample SGA-UV227 Solenoid 0 C C FC EA 49 SG blow- -UV220 Solenoid A 0 C FC Eh down -UV221 Solenoid A 0 C FC EB sample 4g 50 Pool PCE-V071 Hand N 0 or C NA None NA cooling PCE-V070 Hand N OorC Nh None NA 51 Pool PCE-V075 Hand LC OorC Nh None Nh NA cooling PCE V076 Hand LC OorC Nh None Nh NA 52 CVCS GRA-UV001 Notor h R 0 0 'FAI CIAS Eh GRB-UV002 Solenoid A R 0 0 FC CIAS EB 53 Fuel tran Flange Nh NA Dot C C Nh None 54A CB press HCA-HV074 Solenoid 0 0 0 0 Hone Opens EA monitor 55A CB press HCB-HV075 Solenoid 0 0 0 None Opens EB monitor 56 CB purge CPB-UV003A Notor h R LC 0 FAI CIAS IK EB CPA-UV002h 'otor h R LC 0 C FAI CPIAS l+ EA ~ l'/ 57 CB purge CPA-UV002B Notor h R 0 FAI CIAS CPIAS ~IS I+ Eh ~ g CPB UV003B Notor h R 0 FAI EB )il i,i ~ r 58 CB test flange Nh NA NA NA None NA NA ~ l 59 Air IAE-V073 None h C OorC C NA None NA NA 0) IAE-V072 Hand N LC OorC C NA Hone Nh NA I s3 0)
Table 6.2.4-2 CONTAINMENT ISOLATION SYSTEM (Sheet- 8 of 9) Valve Position Pri- Secon ESF(b) mary>> dary(a) Pene- Actua- Actua- Actua Closure tration Valve Valve tion tion Shut- Post- tion Time <<) Po~er Number System Numbers Operator Mode Mode Normal down hccident Failure Signa (Sec) Soui c,l None 60 Chilled water WCE-V039 WCB-VV063 None Motor Nh PAI CIAS /59 NA E!3 Chilled WCB-UV061 Motor FAI CIAS 5l EB water WCA-UV062 Motor FAI CIAS 5R press HCD-HV077 Solenoid Hone Opens 62h CB monitor None Nh NA 62B CB test Flange Nh NA NA NA test Flange Nh NA NA Nh None Nh NA 62C CB 63h SG blow- SGB-UV224 Solenoid 0 FC EB down SGA-UV225 Solenoid 0 FC EA sample 63B SG blow- SGB-UV222 Solenoid C FC EB down SGA-UV223 Solenoid C FC EA sample SIB-V533 None C NA None Nh NA 67 SIS FAI None 10 ED (CESSAR SID-HV331 Motor LC
11) O 72 CVCS CHE-V835 None OorC Nh None Nh Nh 0y OorC
~
CIIB"HV255 Motor FAI Hone 5 EB (CESSAR 57) AFE-V079 None h NA None NA tp 75 hux FW FAI i5 AFC-VV036 Motor R AFAS EC ~
~
p w FAI AFAS EB IlI Motor AFB-UV034 R None Nh NA
~ r 0 Nh 76 hux FW AFE-V080 AFB-UV035 None Motor 0 FAI AFAS ,f EB AFA VV037 Motor 0 FAI AFAS ~ >>5 EA SIA-V523 None C Nh None NA NA 77 5 I'CESSAR SIC-HV321 Motor IC FA I None 10 EC 12)
Table 6.2.4-2 CONTAINMENT ISOLATION SYSTEM (Sheet 9 of 9) Valve Position Pri- Secon-mary Ia) daryIa) ESFIb) I Pene- Actua- Actua- Actua- Closure trat'Ion Valve Valve tion tion Shut- Post- tion Time Ic) Power Number System Numbers Operator Hode Hode Normal down Accident Failure Signal (Sec) Source I 78 CB purge CPB-UVOOSA Hotor R OorC C C FAI CIAS 98 EB CPA-UV004A Hotor R OorC C C FAI (CPIAS EA 79 CB purge CPA-UV004B Hotor OorC C C FAI CIAS B CPB-UVOOSB Hotor OorC C C FAI CPIAS 8 L-1 Air locks NA None NA None NA NA L-3 L-2 Equipment NA None OorC None NA NA hatch O 0 K y t4
~ ~
I
PVNGS F SAR CONTAINMENT SYSTEMS The power access purge used during operation at power has 8-inch containment penetrations sized in accordance with the guidelines of Branch Technical Position CSB 6-4. The 8-inch diameter valves (CPA-UV$04A, 4B, CPB-UV005A, 5B) are designed to close in less that~seconds after receipt of a CIAS or a CPIAS. This minimizes the amount of containment atmosphere mass released to the environment in the unlikely event that a LOCA should occur with the power access purge valves open. n~~>>n- cJ~2
/c. ~
The 42-inch diameter refueling purge valves (CPA-UV002A, 2B, CPB-UV003A, 3B) are 'designed to close in less than~ seconds ~1+ The
'~+df after the receipt of a CIAS or a CpIAS~A~.~ a4 r '~l ~~~-+'.@.
C/+-~g M~~ ~ '~ ~+~ ~ ~ES~~~ setpoint for con ainment isolation is 5 psig and.purge
~g yea~~
isolation valve closure is initiated 0.9 second after the setpoint is reached. I Qa In addition, the following valve characteristics are specified: a ge., Power access valves (8 in.) are ANSI rated 150 lbs. Refueling purge valves (42 in.) are, ANSI rated 75 lbs. Power access valve bodies are hydrotested at 225 psi. Refueling purge valves are hydrotested at 112 psi. Power access valves seat leak test is 150 psi differ-ential. Refueling purge valves are at 75 psi differential. Operability tests will be in conformance with the requirements of section 3.9.
6. 2. 4-22
PVNGS FSAR ENGINEERED SAFETY FEATURE SYSTEMS Table 7.3-6 SAFETY,INJECTION ACTUATION SIGNAL ACTUATED DEVICES LIST (Sheet 1 of 2) Figure No. Description Function
6. 3-1 SI tanks No. 1 through 4 sample isolation valves (4) fill and Close

6. 3-1 SI tanks No. 1 through 4 check valve Close leakage line isolation valves (4)

6. 3-1 HPSI pumps and pump room essential Start cooling units (2)

6. 3-1 LPSI pumps and pump room essential Start cooling units\ (2)

6. 3-1 SI tanks No. 1 through 4 isolation Open valves (4)

9. 3-13 Letdown line isolation valve (1) Close

6. 3-1 Hot leg injection check valve leak Close isolation valve (2)

9. 2-4 Essential cooling water system Refer to and pump room essential cooling section units 7.4.1.1.5

9. 2-1 Essential spray pond system Refe" to section 7.4.1.1.4

9. 5-9 Diesel generator system Refer to section 7.4.1.1.1

9. 4-1 Control room essential Refer to filtration system table 7.3-9 and section

7. 3. l. l. 10. 10

9. 2-8 Condensate transfer system Refer to section 9.2.6
@.5-l C S l ~~p5 ~ad pw.~p s oo~ assen+.~l ++ c.o l;eg u.~ '+s (2 ') 7.3-18
PVNGS FSAR
.ONSITE POWER SYSTEMS powered bus and the diesel will be left running for a period of at least one hour. With an SIAS or AFAS signal present and no loss of preferred power, the operator can manually override the SIAS or AFAS signal from the control room. The diesel generator can then be manually shutdown from the control room or locally. On a subsequent loss of preferred power, load shedding and load sequencing will be initiated. The diesel generator can start accepting loads within 10 seconds of the AFAS/SIAS signal and be completely loaded within 60 seconds after closure of the diesel generator breaker. The ESF loads required for the operation of components within the CESSAR scope in table 8.3-1 will be sequenced on within 30 seconds after receipt of a starting signal as identified in table 8.3-3. Relays at the 'diesel generator detect generator rated voltage and frequency conditions and provide a permissive interlock for the closing of the respective generator circuit breaker. Upon loss of the preferred source of~power without-LOCA, the undervoltage system initiates the starting of the s .diesel generators and sheds all loads. The sequ'encer then automatically initiates the starting of the safe 'shutdown loads upon closure of the diesel generator breaker.
If the diesel generator is supplying power to the ESF bus, a subsequent accident signal initiates start'ng of the loads associated with the subsequent accident signal without. shedding any operating equipment. If offsite power is lost at some time after an accident and the required ESF equipment is running and the diesel generator i<<ribs 5 ~the l
,restart~of agpia~yfe 1
dwater ump 4 the safety injection pumps ir within 3th& w within seconds
~ seconds and such at:
A. Interrupted flow to the core is fully reestablished within 13 seconds. December 1980 8.3-43 Amendment 3
p y.~~~ 0 FSAR ~ L 1+ye.c~&
~~~5gp PVNGS ~ d.> ~4,~ 9 9 >,'//~
EtCC~$
~* ~H.>>
ONSITE POWER SYSTEM ( B. Interrupted auxiliary feedwater flow to the steam generator(s) is fully reestablished within seconds.
~1.
for testing requirements. f '
~ . /
3l During testing if an SIAS or AFAS occurs while the diesel generator is paralleled to the preferred power supply with the control switch in the REMOTE or LOCAL position, the diesel generator breaker will be automatically tripped by a momentary tripping pulse. The diesel generator will continue running and
, automatically revert to the isochronous mode. All non-critical protective devices're bypassed. If a non-critical trip occurs during testing, the diesel generator will trip. On a,subsequent SIAS or AFAS the diesel generator will automatically start and run in the isochronous mode.
The LOCAL control position is selected from the local control panel for diesel generator maintenance testing. A diesel generator LOCAL POSITION alarm will be annunciated in the control room. To prevent, any starting of the diesel generator during maintenance, the OFF position is selected at the local control panel and a DIESEL GENERATOR INOPERABLE alarm is initiated at the safety equipment status system annunciator. 8.3.1.1.4.6.,'I%/C If the preferred power source is lost while paralleled to the diesel generator during testing, the diesel generator will trip on overcurrent and the diesel generator breaker will trip automatically on a diesel generator shutdown signal. Upon detection of undervoltage on the Class IE 4.16 kV bus, load shedding and sequencing will be initiated as described in section 8.3.1.1.4.8 Diesel Generator Fuel Oil Stora e and Transfer
~S stems. Refer to section 9.5.4 for system description.
8.3.1.1.4.9 - Coolin and Heatin S stems. Refer to section 9.5.5 for system description. Amendment 3 8.3-44 December 1980
PVNGS F SAR
SUMMARY
DESCRIPTION Table 10.1-1 STEAM AND POWER CONVERSION SYSTEM DESIGN AND PERFORMANCE CHARACTERISTICS (Sheet 2 of 4)' System/Component Performance Characteristics Main Steam .System (Section 10.3) Main steam piping From each steam generator up to and t including the main steam isol'a-tion valves: ASME III, Code Class 2. (design pressure 1270 psia, design temperature 575F, Seismic Category I ) Balance of the main steam piping: ANSI B31.1.0 g.& Main steam Maximum closing time 9 seconds isolation valves after receipt o fClass signal. (1 per steam line) ASME III, Code 2 (design pressure 1270 psia, valves. design temperature 575F, Seismic Category I) Main steam safety Flow capacity equal to 105% of the valves maximum calculated steam genera-(5 per steam line) tor mass flow at setpoint pressure: ASME III, Code Class 2 valves. (design'pressure 1390 psia, design temperature 575F, Seismic Category I) (See CESSAR Table 5.4.13-2). Power operated Flow capacity equal to 15% of the atmospheric dump design main steam flow with 100% valves ~ redundancy: ASME III, Code (1 per steam line) Class 2 valves. (design pressure 1270 psia, design temperature 575F, Seismic Category I)* Turbine Bypass System (Section 10.4.4) Bypass Valves Flow capacity equal to 55% of Downstream of design steam flow: Piping ANSI Main Steam.Iso- B31.1.'0 (design pressure 1270 lation Valves psia, design temperature 575F, (Six piped to con-, non-Seismic Category I denser, two piped to atmosphere) 10.1-4
PVNGS FSAR
SUMMARY
DESCRIPTION.
'able 10.1-1 STEAM AND POWER CONVERSIQN SYSTEM DESIGN AND PERFORMANCE CHARACTERISTICS (Sheet 3 of 4)
System/Component Performance Characteristics Condenser See section 10.4.1 Condenser Air See section 10.4.2 Removal System Circulating Water See section 10.4.5 System Turbine Gland Seal See section 10.4.3 System Condensate and Main Piping in cain steam support Feedwater System structure (MSSS) to (Section 10.4.7) downstream feedwater isolation valvei-ASME III, Code Class 2. Design pressure 1875 psi, 500F; from downstream feed-water isolation valves to steam generators - ASME III, Code Class 2. Design pressure 1325 psi, 500F,, Seismic Category. I. All other piping ANSI B31.1. Balance of system piping: ANSI B31.1 Auxiliary Feedwater Two motor-driven auxiliary feed- q'5O System (Section 10.4.9) turbine-driven a
.water pump, eac 'ar water pumps, and one steam ,
feed-ga /mxn . delivered capacity and one 300,000 gallon minimum capacity condensate storage tank. 10.1-5
0 PVNGS F SAR MAIN STEAM SUPPLY SYSTEM pressurization of the steam generators. Also, sample con-nections are provided downstream of the steam for determination of steam quality. Branch piping generator'ozzles p'rovides steam to moisture separator reheaters, main and feedwater pump turbine gland steam sealing systems, the feedwater pump turbines, the auxiliary feedwater pump tur- ~ bine, and the bypass steam to the condenser and to atmosphere. The main steam lines are designed to permit preoperational cleaning to remove foreign material and rust. The design is such as to prevent entry of foreign material into either the steam generators or turbine generator. The main steam piping drops several feet immediately downstream of the steam generators prior to being routed to the turbine stop .valves. The lines are sloped in the direction of the turbine, and drains are provided at all low points to provide for flushing and drainage. Pertinent design parameters for the main steam piping out to the main steam isolation valve are presented in table 10.3-2. The main steam piping out to the first isolation valve is inspected and tested in accordance with ASME Code Sections III and XI. ANSI B31.1 piping is inspected and tested in:"accor-dance with paragraphs 136 and 137 of ANSI B31.1. 10.3.2.2.2 Main Steam Isolation Valves Each of the main steam lines is equipped with one quick acting main steam iso ation valve (MSIV). Each'valve has an actua-tion time of seconds or less and operates automatically in the event of rupture in the main steam piping or associated components either upstream or downstream of the MSIV': They 'prevent blowdown of more than one steam generator (assuming a single active failure). The valves are designed to close upon loss of electric power. 'Once isolation is initiated, in response to a main steam isolation signal, the"valves con-tinue to close and cannot be opened until manually reset. 10.3-7
0 0
PVNGS FSAR OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM e 2. Hot standby for 8 hours with a loss power and normal onsite power. of offsite
3. Reactor coolant system cooldown using the intact steam generator following a main steam line break or main feedwater'ine break inside the contain-ment with a loss of offsite power and normal onsite power.
G. Safety Design Basis Seven Each of the two Seismic Category I AFS pumps sh ll be to provide 100% of the required flow ( 5'esigned gal/ min) for decay heat removal. The head generated by each pump is sufficient to deliver feedwater into the steam generators at a pressure equivalent to the accumulation pressure of the lowest set safety valve plus system frictional resistance and static head. The non-Seismic Category I AFS pump shall meet this requirement with no miniflow. H. Safety Design Basis Eight In the unlikely event that the control room must be evacuated, the AFS shall be operated for shutdown from a remote shutdown station. Safety Design Basis Nine The Seismic Category I motor-driven AFS pump shall be located in a separate room designed to Seismic Cate-gory I requirements in the main steam support structure. The Seismic Category I steam turbine-driven AFS pump shall also be located in a separate room designed to Seismic Category I requirements in the main steam support structure. The non-Seismic Category I motor-driven AFS pump is located in the turbine building. Amendment 5 10.4-44 August 1981
PVNGS FSAR OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM steam isolation valves. The turbine controls and associated valves are powered from the dc bus. 10.4.9.2.2 Component Description Principal components are listed in table 10.4-6. 10.4.9.2.3 System Operation For emergency operation, normal flow is from the condensate tank to either the Seismic Category I motor-driven AFS pump or to the steam turbine-driven Seismic Category I AFS pump which are located in the main steam support structure. An alternate supply of water is provided by cross connections to the reactor makeup tank. A minimum flow recirculation system is provided on each pump discharge with recirculation to the condensate tank. Each pump can supply either steam generator with feedwater. Table 10.4-6 AUXILIARY FEEDWATER SYSTEM DESIGN DATA Design Factor Value Auxiliary feedwater pumps Quantity Motor-driven non-Seismic Category I Motor-driven Seismic Category I Steam Turbine-driven Seismic Category I Flow, gal/min, net 5 8'/54IC CAT'C'6r><< < Head, fl ~gl/ ~5&i, C epxie vm ft - Seismic Category I 3280 Head, ft - non-Seismic Category I 2960 Head, ft - non-Seismic Category I without 3200 miniflow bypass at 875 gal/m PLou -u>> Sai>><< ~4~<+>~~> 4<<f~~~ t7F
<htgko8~ Lp ASS g(y]ping~ -hlguu'SSI54'4 /pre Pg7.fgnhJ 5
yECI VfD ta >TF~Ah +gA/EIZ+og, 10.4-47 Amendment 5 August 1981
0 PVNGS FSAR OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM P Condensate recirculation lines are provided downstream of the control stations to allow for system testing up to the feed-water isolation valves. One auxiliary .,feedwater path 'to the steam generators is provided for,the non-Seismic Category I,,... motor-driven auxiliary feedwater pump through the feedwater header, with manual operation of feedwater valves possible during emergency operation. 1 I The two Seismic Category I auxiliary feedwater M pumps only pro- ~ ~ vide flow to the downcomer feedwater I nozzles I on each steam I generator. Either Seismic rCategory I ~ r W I auxiliary feedwater pump
~
can supply the necessary feedwater. for reactor decay heat .-.. removal and reactor cooldown to 350F. At a reactor coolant temperature of 350F,,the .shutdown cooling system is placed in operation.,
~ ~ ~ ~
minimum flow ath is provided for each gum . " Approximatelv,."
~ ' ~
5EiSIHiC eATEr uRY? @goof rHE uouSE Mute CAltkoA
~u recz.rculated Z kd~PdhPP iTV % of the<pump capacitya.s bac to t:ne condensate tank whenever a pump is operating.". The minimum flow line is provided to prevent pump overheating in the event the pump discharge line is shut off. If a break is postulated o occur in the recirculation line, since the flow restriction orifice limits the flow to 13% of pump capacity flow,,system operation is not affected. The pump still delivers required, flow to the:
steam generators. The water inventory of the condensate tank i has been calculated to include thei possibility of 13% flow I w p ~ ~ water loss through the recirculation line while maintaining a sufficient quantity of water to provide the required cooling as described in section 10.4.9.3. ~ ~ T ) I,A One motor driver is powered from a separate engineered, safety features (ESF) bus which is powered by the load group 2 diesel generator. The Seismic Category I steam turbine-driven pump's associated valving is powered from the dc bus as discussed in
~ ~
PVNGS FSAR OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM F. Safety Evaluation Six The AFS is designed to maintain an adequate water level in the steam generators under the following operating mo'des and accident conditions:
1. Reactor cooldown at a maximum administratively con-trolled rate of 75F/h from hot standby to 350F with a loss of offsite power and normal onsite power.

2. Hot standby for 8 hours with a loss of offsite power and normal onsite power.

3. Reactor coolant system cooldown using the intact steam generator following a main steam line break or main feedwater line break inside the contain-ment with a loss of offsite power and normal onsite power.
Only the Seismic Category I pumps are started on an AFAS. However the non-Seismic Category I motor-driven pump can be manually started from the control room. Only the non-Seismic Category I motor-driven pump is used for startup, hot standby, and normal shutdown. G. Safety Evaluation Seven Each of the three AFS pumps is capable of delivering 150 net gal/min against a back pressure in the steam generator equivalent to the accumulation pressure of the lowest set safety valve plus system frictional resistance and static head. Safety Evaluation Eight The AFS can be operated from either the control room or from a remote shutdown station. Safety Evaluation Nine Each Seismic Category I AFS pump is installed in a 0 separate room designed to Seismic Category I require-ments. These rooms are in the main steam support structure. Amendment 5 10.4-54 August 1981
PVNGS FSAR 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.1 DECREASE IN FEEDWATER TEMPERATURE I10 Refer to CESSAR Section 15.1.1 T.HE Prhl& Fs~epwgg
<><<~ yggT ISOZyriom V'WL V E ~ <<<<~~
15.1.2 INCREASE IN FEEDWATER FLOW Se<o~gg, gEFFI ~ ~<<<"~ yF
~'caawapp Refer to CESSAR Section 15.1.2. ~ ~ lo Fog hxl5dd5$ /ON Hq~ 7qiy PE'NA7-ION Po'<< ~
15.1.3 INCREASED MAIN STEAM FLOW QF+Eg~ 76E 48658~ +~ Refer to CESSAR Section 15.1.3. h&A~~~I5 15.1.4 INADVERTENT OPENING OF A STEAM GENERATOR RELIEF OR SAFETY VALVE 10 Refer to CESSAR Section 15.1.4> 15.1.5 STEAM SYSTEM PIPING FAILURES I
@HE PRwIazRy PFF ga.ruF y HE'SE CONTAINMENT ~up Pv~h5 sP~I~i<<.oHWNAE5 << @XX iW orner. CEFk r A~y I~5FKZ~
15.1.5.1. Identification of Event an aus 7 RE +~n.cgWAg ggy~1p
~ ~
zap ryp pppsar~ sr Refer to CESSAR Section 15.1.5.l. except that the limiting stealli line break, CESSOR Case 1 (SKPLOP), is reanalyzed for the PVHGS specific feedvater isolation valve closure time and the ruptured generator delta pressure isolation (lockout) setpoint. R Cases 2-1P are bounded by the Case l 15.1.5.2 e uA'lt:e o'z'r".'veBcF "a'r'id S stems 0 eratxon THE 5EQ~AC8 Nf V&)F5 A'g ( H 5h JP C4 Sf gefer to CESSAR Section 15.1.5.2. Also, Pa/o Verde uta. x J> SL I= ZO 15 10 other reactor trip signal - steam generator dP low flow~+ vhpA8 oUFRfLElc', rIIK sia+Ais4KE +~~a+~ res c pc r@psidr~ac. yap 12 The effect of th1sEreactor trxp signals has been included in section 15.1.5.3. Additionally, CIAS or SIAS signals will actuate control room and fuel building essential ventilation systems. See sections 6.4 and 9.4 for details. CIAS or SIAS 10 will terminate the containment power access purge as described in section 9.4.
~55/p a+~$ 5 ~Q 4/pgE Q+oQEAJ To C'&IJSERVAT>ueI Py 5<+ 1AXi~)eC + ~g~T.ig( Fot WE4eptzgr1oa oFFcEL. PFRFNbi~ag +n o H<<
THE (o~~ gE)( g~ ~+a'(ASS
~,~~> gg~ggpZuy TH+ ATMT<4C F~N FV >~ W <~ g)r CiASE ~,
lIipti.VEaT dRriOu. <F65AIr- CASE AAoV+T 4h qgyqgg frbDR VO W1'~1 <~ ~<S<d7S r~W+ i~agEgSE ~ (i~ rgb
+C,r Hda P4r H l tf E uF AYiw W5'LA N~~E eE Is ~o RPPREuid Zc.E l~ ~cvupE~shre REiFR5Eu ro 7RE Wr~nsPgERE, An-(virv nF r&E" u o ~ >FFSnE'os'E PUG vox.HE WEhsiclg<p CENSER C'ASKING Awe C A~V6'~or O'E'&V CBVISafM.
4<F~<<cz ~<<7 o~ /.9(.2.+. (O FOR FagI-gag SEr~u.g. February 1984 15.1-1 Amendment 12
PVNGS FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.5.3 Anal sis of Effects and Conse uences Refer to CESSAR Section 15.1.5.3 since the 0 to 2 hour X/Q value presented in section 2.3 is smaller than that assumed in the,CESSAR radiological consequence analysis. A. Mathematical Model Refer to CESSAR Section 15.1.5.3.A. B. Input Parameters and Initial Conditions Refer to CESSAR Section 15.1.5.3.Bg<CEPy ~~1 /8 VoL 5'F1>>> o/soRT Rd5 Is //o oo 17/E '5/fFEIY Frr A/O EclvCiA/ C/MES ISEt oRE' FoR Fvus6 eo/ pdzEp re izo Frs FoR d E+sAR E C. esults Case 1: Large Steam Line Break During Full Power Operation with Concurrent Loss of Offsite Power (SLBFPLOP) l>SEXI O~ Refer to CESSAR Section 15.1.5.3.C. Additionally due to decreasing core flow following loss of power to the reactor coolant pumps conditions exist for a steam generator hP low flow trip. Also for Palo Verde, FSAR table 15.1-1 is used in place of CESSAR table 15.1.5-1. Case 2: Large Steam Line Break During Full Power Operation with Offsite Power Available (SLBFP) Refer to CESSAR Section 15.1.5.3.C.. Additionally after a steam line break a trip Si6 <RC.j
~can OR a:.had be initiated FOR.SCS'~$ .'XWSIDE'CO~~Aj~~b~jdf bv a variable overpower 0
trip~ pigh
~ ~
containment pressure trip~. Amendment 12 15.1-2 February 1984
0 Insert E Additionally case 1 utilizes the PVNGS specific feedwater isolation valve closure time of 9.6 seconds and the steam generator bP isolation setpoint of 325 psid.
e The dynamic Insert C to FSAR page behavior of the salient NSSS parameters 15.1-2 following the SLBFPLOP is presented in Figures 15.1-1 through 15.1-16. Concurrent with the steam line break, a loss of offsite power occurs. At this time,, an actuation signal for the emergency diesel generators is initiated. An auxiliary feedwater actuation signal (AFAS) is conservatively assumed to be immediately initiated for the ruptured steam generator. Due to decreasing core flow following loss of power to the reactor coolant pumps, conditions'xist for a Steam Generator hP Low Flow or low DNBR trip. Additionally, for inside t inment breaks a trip can be initiated by a high containment pressure trip signal. At 0.6 seconds, a trip signal'second is initiated. At 0.75 seconds, thee reactor trip breakers open. After a 0.34 coil decay delay, the CEAs
'ain begin to drop into the core at 1.09 seconds.
t ri p. a'uillotine Turbine trip occurs upon reactor steam flow is not affected since the SLB is a double-ended break of a main steam line. At 7.6 seconds, voids begin to form
'n the upper head of the reactor vessel.
At 9.1 seconds, the steam generator pressure drops below the main steam isolation signal (HSIS) analysis setpoint of 810 psia. This results in generation of an MSIS at 10.1 seconds. The MSIS initiates closure of the main steam isolation valves (HSIVs) and main feedwater isolation valves (MFIVs). The HSIVs close by 14.7 seconds and the MFIVs close by 19.7 seconds. At 28 seconds, the pressure difference between the steam generators reaches the analysis setpoint of 325 psid for lockout of auxiliary feedwater (AFW) to the ruptured steam generator, Isolation of AFW from the ruptured steam generator begins at 29 seconds. Ry 44 seconds switching of the AFW to the intact. steam generator is complete; AFW to the ruptured steam generator has been isolated and the AFW valves to the intact steam generator are fully open. At 94 seconds, the pressurizer empties. At 139 seconds, the pressurizer pressure has dropped below the analysis setpoint of 1600 psia for safety injection actuation, resulting in initiation of a safety injection actuation signal (SIAS) at 140 seconds. Within 30 seconds of SIAS, the HPSI valves are fully open and the operable high pressure safety injection (HPSI) pump delivers full flow. At 230 seconds, the affected steam generator empties. Safety injection boron begins to reach the core at 249 seconds. At 276 seconds, the maximum core reactivity (-0.03~ a P) occurs. As shown by Figure 15.1-16 the values of DNRR remain above those for which fuel damage would be indicated. At a maximum of 30 minutes, the operator commences a controlled plant cooldown via the appropriate atmospheric dump valves, assuming that offsite power has not been restored. Shutdown cooling is ini tiated when the RCS reaches shutdown cooling entry conditions.
PVNGS FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Case 3: K,arge Steam Line Break During Zero Power Operation with Concurrent Loss of Offsite Power (SLBZPLOP), K Refer to CESSAR Section 15.1.5.3.C. pppfg p 5ypp~ gj)gg ggEpg p/(r9 7Hz A 055 bF off5lrg8hlb A zR t f'Q Am. gE IwirIArEn/y PJQ + P jP PP g gg gP k'$ /A/5I PC C'O~l RIMMEwer~
$ /@slow'trap p
SSCoaZL. h gigh containment pressure trip Case 4: Large Steam Line Break Zero Power Operation with Offsite Power Available (SLBZP) See CESSAR Section 15.1.5.3.C. Case 5: Small Steam Kine Break Outside Containment During Full Power Operation with Offsite Power Available (SSLBFP) See CESSAR Section 15.1.5.3.C. Case 6: .Large Steam Line Break Outside Containment from Zero Power Operation with Loss of Offsite Power (SLBZPLOPD) See CESSAR Section 15.1.5.3.C.
't 15.1.5.4. Conclusion Refer to CESSAR Section 15.1.5.4.
February-1984 15.1-3 Amendment 12 )iz
I PVNGS. FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.1-1 SEQUENCE OF EVENTS FOR A LARGE STEAM LINE BREAK DURING FULL POWER OPERATION WITH CONCUI&ENT LOSS
- -OF OFFSITE POWER (SLBFPLOP)
T~ime s Event Set oint or Value 0.0 Steam Line Break an Loss o Offsite Power Occur 0.6 Steam Generator bP Low Flow Trip Condition Occurs, % of nitial Full Power Mass Flow 70 o Low DNBR Trip Condition Oc rs, Projected DNBR 1.19 l' 0.75 Trip reakers Open gq4 g 1.09 CEAs Be 'n to Drop 8.0 Voids Begi to Form in RV Head 8.3 Main Steam Iso ation S'gnal, 810 psia 13.3 MFIVs Close Compl ly 13.3 MSIVs Close Comp t ly
13. 3 EFW Initiated o Inta Steam Generator 120 Pressurizer mpties 178 Safety I ection Actuation 'nal, 1600 psia 208 Safet Injection Flow Begins 237 Aff cted Steam Generator Empties 259 M xjmum Transient Reactivity, +0.09 0 b,p 277 Minimum Post-Trip DNBR 2.7 280 Safety Injection Boron Begins to Reach Reactor Core 1800 Operator Initiates Cooldown Amendment 12 15.1-4 February 1984
0 Insert D to FSAR page 15.1-4 0.0 Steam Line Break and Loss of Offsite Power Occur AFAS Assumed to be Generated for Ruptured Steam Generator 0.6 CPC ow DNBR Trip Si na'I or Steam Generator P~ 1.19 Low Flow Tri Signal Generated,~Looff TInnit al Full Power Mass F ow 70 0.75 Reactor Trip Breakers Open 7.6 Voids Begin to Form in RV Upper Head 9.1 Steam Generator Pressure reaches Hain Steam Isolation 810 Signal (HSIS) Analysis Setpoint, psia 10.1 HSIS Generated 14.7 HSIVs Completely Closed 19.7 HFIVs Completely Closed 28.0 Difference Between Steam Generator Pressures reaches 325 Analysis Setpoint for Lockout of AFW to Ruptured Steam Generator, psid 29.0 Signal to Isolate AFW from Ruptured Steam Generator Generated 44.0 AFW Isolated from Ruptured Steam Generator; AFW Valves to Intact Steam Generator Fully Open 94 Pressurizer Empties'ressurizer 139 Pressure reaches Safety Injection Actuation 1600 Signal (SIAS) Analysis Setpoint, psia 140 SIAS Generated 170 Safety Injection Flow Begins 230 Affected Steam Generator Empties 249 Safety Injection Boron Begins to Reach Reactor Core 276 Maximum Transient Reactivity, 10 -0.03 285 Minimum Post-Trip ONBR 5.4 1800 Operator Initiates Cooldown
PVNGS FSAR INCREASE IN HEAT REMOVAL THE SECONDARY SYSTEM Table 15.1-2 QUENCE OF EV1MZS FOR A LARGE STEAM LINE B DURING FULL POWER OPERATION WITH OFFSITE
- -POWER AVAILABLE {SLBFP.) ~
T~ime s Event S t oint. or Value 0.0 Steam L ne Break Occurs 6.95 Variable ver Power Trip '16 Condition ccurs, % Ex-Core Neutron Fl or Low DNBR Trip Condition 0 urs, Projected 1.19 DNBR 7.10 Trip Breakers pen 7.44 CEAs Begin to D p 11.9 Voids Begin to Fo in V Upper Head 13.5 Main Steam Isolatio Signal, 810 psia 18.5 MFIVs Close Comp etely 18.5 MSIVs Close Co pletely 18.5 EFW Initiate to Intact S am Generator 67 Pressuriz r Empties 90 Safety jection Actuation Si al, 1600 psia 120 Safe Injection Flow Begins 149 Af cted Steam Generator Empties 151 M xjmum Transient Reactivity, -0.18 0 d,p 151 Minimum Post-Trip DNBR 26 160 Safety Injection Boron Begins to Reach Reactor Core 18 0 Operator Initiates Cooldown February 1984 Amendment 12
0 PVNGS FSAR INCREASE lN HEAT REMOVAL
~
BY THE SECONDARY SYSTEM Tab3.e 15.1-3 SEQUENCE OF EVENTS FOR A LARGE STEAM LINE BREAK DURING ZERO POWER OPERATION WIT+ CONCUMENT OSS OF OFFSITE POWER (SLBZPLOP AND SLBZPLOPD) T~ime s Event Set o'nt or Value 0.0 Ste Line Break and Loss of Of site Power Occur 0.6 Steam G nerator dP Low Flow Trip Con ition Occurs, % of Initial ll Power Mass Flow or Low DN Trip Condition 70 Occurs, Pr 'ected DNBR 1.19 0.75 Trip Breaker Open 1.09 CEAs Begin to rop 5.7 Main Steam Isol ion Sign 810 psia 10.7 MFIVs Close Comple l 10.7 MSIVs Close Complet 10.7 EFW Initiated to ntac Steam Generator Safety .Injec on Actuatio Signal, 1600 psia 55 Voids Beg'o Form in RV Up r Head 59 Press izer Empties 75 Saf y Injection Flow Begins 120 S fety Injection Boron Begins to each Reactor Core 189 Maximum Transient Reactivity, -0.06 10 dp 1240 Affected Steam Generator Empties Operator Initiates Cooldown Amendment 12 15.1-6 February 1984
150 125 C) 100 75 50 0 t00 ZQG 300 400 500
't" SECONDS FULL POWER LARGE STEAM LINE BREAK WITH Figure CONCURRENT LOSS OF OFFSIDE POWER CORE POWER vs TIME 15,1-1
0 150 125 100 75 5G 0 100 200 300 400
'!!'I" . SKCONGS FULL POWER LARGE STEAM LINE BREAK WITH Fig vre CONCURRENT LOSS OF OFFSITE POWER CORE HEAT FLUX vs TIME 15,1-2
2500 2000 1500 1000 5GG 100 200 300 400 500 TT.N . HEI.ONES FULL POWER tARGE STEAM LINE BREAK WITH Rgure CONCURRENT LOSS OF OFFSITE POWER RCS PRESSURE vs TIME 15,1-3
4SOOQ. 40000 36000
,30000 25000 a
20000 C) TOTAL, 15000 AFFECTED SG 100QO INK. INTACT SG SGOO 0 0 a>0 200 300'00 SCG T SEt:ONGS FULL POWER LARGE STEANl L1NE BREAK WITH Figure CONCURRENT LOSS OF OFFSETE POWER REACTOR COOLANT FLOW RATE vs TEME 15.1-4
600 CORE OUTLET gpRE-AVERA&E CORE INLET 300 100 200 300 400 gl 0 TlI'lE. SECQNGS FULL POWER LARGE STEAM ONE BREAK WITH CONCURRENT LOSS OF OFFSITE POKIER REACTOR COOLANT TEMPERATURES 5) vs TIME
600 TMTAcT S6 HOT Q'6 ch 550 INTAC< $ 6 COLD LESS 8 SGQ g ~ AFRcTaa S6 Ho~ ~q CC Q Leal 4QQ PFFED'ED SQ COLO LC&S 300 100 200 300 400. TIME . SECONDS FULL POWER lA RGE STEAM LINE BREAK WITH Rgure CONCURRENT LOSS OF OFF SITE POlNER 15,1-58 REACTOR COOLANT TEMPERATURES (B) vs TIME
10 Ho DERAToR DOPP l~R ToTAV sAFsv( mzcnroe 00 3GO 400 Cn 100 T Iii= . SECOiMGS FULL POWER LARGE STEAM LINE 8REAK NETH CONCURRENT LOSS OF OFFSOE POWER REACTIYITYCHANGES vs TIlNE
1200 1000 800 200 100 ".00 300 400 EGG T<ilE, SECONDS FULL POWER LARGE STEAM LINE BREAK WITH CONCURRENT LOSS OF OFF SITE POWER PRE SSURIZER WATER VOLUME vs TIME
~ 1200 1000 INTACT STEAM GENERATOR BGO 600
~ AFFECTED STEAM GENERATOR 100 200 300 . 400 SGQ TINK SKCQiRGS FULL POWER LARGE STEAM LINE BREAK WITH CONCURRENT LOSS OF OFFSITE POWER STEAM GENERATOR PRESSURES vs TIME
6000 i ~ N SOGO 4000 O-3000 CO 2000 O AFFECTED STEAM GENERATOR 1000 INTACT STEAM GENERATOR,
.'00; i!, GQ t i 300 St=PGqqg FULL PONE R LARGE STEAM LINE BREAK IlttITH Figure CONCURRENT I OS S -OF OFFSIDE POKIER 15,1-9 STEAM GENERATOR BLOWDOWN RATES vs TIME
2000 1500 1000 500 INTACT STEAM GENERATOR
~ AFFECTED STEAM GENERATOR 00 100 200 '00 TIME, SECONDS FULL POVJER LARGE STEAM LINE BREAK WITH figure CONCURRENT LOSS OF OFF SITE POMIER 15.1-1 FEEDWATER FLOW RAKS vs TIME
0 o= 300 200 100 0 200 300 500 TIME, SECONDS FULL POWER NRCE STEAM LINE BREAK WITH Figure CONCURRENT LOSS OF OFFSIDE BREAK FEEDWATER ENTHALPY vs TIME 15,1-11
300000 250000 O 200000 150000 INTACT STEAM GENERATOR 100000 iGQQO ~ AFFECTED STEAM GENERATOR 10" 200 3GO 4GG TIi'!= SECQiF~S FULL POWER NRCE STEAM UNE SREAK WITH Figure CONCURRENT LOSS OF OFFSITE POWER 15,1-12 S~M GENERATOR MASS INVENTORIES vs TIME
350000 300000 250000 u 200000 150000 Ch g 100000 0 0 1 GQ ZQG 30G 400 T I HE SECONDS FULL POWER LARGE STEAM UNE BREAK WITH Fig vre CONCURRENT LOSS OF OFFSITE POWER 15 1-1z NTEGRATED STM MASS RELEASE THRU BREAKvsTIM
200 f60 120 C) 80 40 100 200 3rp 400 TINE SECQNGS FUU. POWER fARQE STEAM UNE BREAK WITH CONCURRENT LOSS OF OFFSITE POWER SAFETY INJECTION FLOW vs TIME
2000 TOP OF REACTOR VESSEL l50O LIQUIO VOLUME l000 a'00 TOP OF HOT KG 0 t00 happ 300 4pp
'tIMFSECONDS FULL POSER LARGE STEAN LINE BREAK WITH Figure CONCURRENT LOSS OF OFFSITE POMIER 15,1-15 REACTOR VESSEL LIQUID VOLUME vs TIME
10 6 5 N 4 200 300 TIME, SECONDS FULL POWER URGE STEAM LINE BREAK LOSS OF OFF SITE POWER WITH'ONCURRENT MINIMUMPOST-TRIP DN8R vs TIME
PVNGS FSAR lS.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2. 1 LOSS.-'OF .EXTERNAL LOAD Refer to CESSAR Section 15.2.1. 15.2.2 TURBINE TRIP Refer to CESSAR Section 15. 2. 2. 15.2.3 LOSS OF CONDENSER VACUUM 15.2.3.1 Identification of Event and Causes Refer to CESSAR Section 1S.2.3.1. 15.2.3.2 Se uence of Events and S stems 0 eration Refer to CESSAR Section 15.2.3.2. Additionally, CIAS or SEAS, signals will actuate control -room and fuel building essential ventilation systems. See sections 6.4 and 9.4 for details. A CIAS or SIAS will terminate the containment power access purge, as described in section 9.4. 2ivsE.iz TA The auxiliary feedwater system is described in section 10.4.9. ~ 15.2.3.3. Anal sis of Effects and Conse uences Refer to CESSAR Section 15.2.3.3. 15.2.3.4. Conclusions Refer to CESSAR Section 15.2.3.4. 15.2.4. MAIN STEAM ISOLATION VALVE CLOSURE Refer to CESSAR Section 15.2.4. 15.2.5.
~ ~ ~ STEAM PRESSURE REGULATOR FAILURE Re fer to CES SAR Section 15 . 2 . 5 .,
December ]982 1") . 2-1 Am> nlm~nt lA
PVNGS FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.6 LOSS OF NON-EMERGENCY A C POWER TO THE STATION AUXILIARIES Refer to CESSAR Section 15.2.6, 10 15.2.7 IOSS OF NORMAZ FEEDWATER FJOW Refer to CESSAR Section 15.2.7. 15.2.8. FEEDWATER SYSTEM PIPE BREAKS to CESSAR Section 15.2.8 and CESSAR A 12 Section xcept th xs taken for the high contai ressur rmining the peak RC ssure. Qc.~K~K, Geiger i~c l wscar'Yea uKw'M'swR SEczi~ Co~wm~K~ iu ~a~ ~s~~ Q i+~6K P~~'. Amendment 12 15.2-2 and>>3/-4 Deleted Februarv 1984
Insert "A" With reguards to the loss of condenser vacuum analysis, the PVNGS Auxiliary Feedwater System deviates from the CESSAR assumed response in that the auxiliary feedwater flow delivery has been reduced from 875 to 750 gpm per pump, and that the motor dri ven and steam driven auxiliary feedwater pump start times have been increased from 10 to 22 and 29 seconds respectively. See section 1.9.2.4.10 for a discussion of how these deviations do not affect the conclusions of the CESSAR Safety Analysis.
15.2.8 FEEOWATER SYSTEM PIPE BREAKS 15.2.8.1 Introduction Refer to CESSAR Appendix 158. In addition, refer to section 1.9.2.4.10 for a discussion of how the PVNGS auxiliary'eedwater performance deviates from the CESSAR assumed response. This section provides an analysis documenting the adequacy of the PVNGS auxiliary feedwater system to maintain long term RCS heat removal during the limiting Chapter 15 transient, the feedwater line break. For this transient, the maximum RCS pressures reported in Appendix 15B of CESSAR will not be impacted as a result of the reduction in auxiliary feedwater flow from 875 to 750 gpm, because the maximum RCS pressure during a feedwater line break occurs prior to the deli very of auxiliary feedwater flow. The feedwater line break transient is the limiting Chapter 15 transient with respect to long term RCS heat removal because of its analysis assumptions, which include:
1) An instantaneous total loss of all main feedwater to both steam generators at the event initiation and throughout the transient,

2) A reactor trip on a conservatively delayed low level trip signal which minimizes the steam generator inventory available for RCS heat removal,

3) 'he single failure of one of the two available safety related auxiliary feedwater pumps,

4) All auxiliary feedwater which is diverted to the ruptured steam generator throughout the event is not credited for heat removal, and

5) The auxiliary feedwater temperature is the maximum value within the allowable range of 40 F to 180 F.
Another analysis assumption which makes feedwater line breaks limiting is the treatment of the steam generator break discharge enthalpy. In order to maximize the primary 'system pressurization, the enthalpy is adjusted to maximize the mass release rate while minimizing the energy removal rate. The CESSAR Appendix 15B analyses assume that the break discharge enthalpy is that of saturated liquid until the ruptured steam generator dries'ut. At this point the discharge enthalpy is assumed to switch from saturated liquid to that of saturated vapor. This analysis assumption is made with breaks in the steam generator economizer lines in mind due to their very low elevation. For breaks in the downcomer lines this assumption is very conservative. If a break were to occur in the downcomer feedwater line, the discharge would be mainly steam prior to steam generator dry out due to the very high discharge elevation (the line discharges above the lower level of the moisture separators). Therefore, a break in the downcomer line will appear very much like a small steam line break. The downcomer nozzle is 6" in diameter which results in a break no greater than 0.2 sq ft. The CESSAR FSAR Appendix 158 analyses assumed that, following actuation of auxiliary feedwater on low steam generator water level, 875 gpm would be delivered to the intact generator throughout the transient. For the PVNGS
design, auxiliary feedwater actuation to the ruptured steam generator will divert auxiliary feedwater flow away from the intact steam generator at a rate which is dependent on the pressure difference between auxiliary feedwater lines to'ach steam generator until the ruptured steam generator is identified and isolated. This flow diversi'on is specifically accounted for in this evaluation. 15.2.8.2 Discussion Refer to CESSAR Appendix 15B. In addition, in order to demonstrate the adequacy of long term RCS heat removal with the PVNGS auxiliary feedwater system deviations, the full spectrum of feedwater line breaks were evaluated. It was determined that for breaks larger than the limiting break size presented in Appendix 158 of CESSAR (0.2 sq ft) the main steam isolation signal on low steam generator pressure would occur very early in the transient. Once main steam isolation occurs for these larger breaks, the 325 psi difference between the steam generators established very quickly. The ruptured steam generator is then isolated, and full auxiliary feedwater flow is provided to the intact steam generator. For example, the 0.6 sq ft feedwater line break transient would reach the main steam isolation setpoint on low steam generator pressure within approximately one minute. Credit is not taken for main steam isolation on high containment pressure which could come very early in the transient, especially for the larger breaks. In addition to the speed with which a main steam isolation signal would be generated, the larger break sizes would deplete the ruptured steam generator inventory more rapidly due to the increased break discharge rate, causing an earlier trip on low steam generator level. This results in the intact steam generator having more liquid inventory in it at the time of the reactor trip, yielding greater inventory for long term RCS heat removal. Breaks smaller than 0.2 sq ft were also evaluated with respect to long term RCS heat removal. It was determined that, for the full spectrum of small feedwater line breaks, the RCS will not overpressurize in the long term. Very small feedwater line breaks will trip on low steam generator water level in the intact steam generator before the conservatively delayed low level trip signal in the ruptured steam generator. With the additional liquid inventory available in the ruptured steam generator at the time of the reactor trip, long term heat removal is ensured. In order to demonstrate the adequacy of the PVNGS auxiliary feedwater design, the limiting CESSAR Appendix 158 feedwater line break transient is presented in this section with the auxiliary feedwater syste~ changes discussed previously. 15.2.8.3 method of Analysis 15.2.8.3.1 Hathemati cal Model s S ee CESSAR A ppeendix 158, except that the analysis of the feedwater line break event was performed using the CESECIII computer program descri b e d in CESSAR Section 15.0.3. The method chosen for analyzing the'feedwater line break presented herein is the. same as that used in CESSAR Appendix 158.6.2, except for the treatment of
auxiliary feedwater. In order to evaluate the PVNGS auxiliary feedwater performance, the CESECIII code was modified to include the PVNGS specific auxiliary feedwater system piping resistances and pump head-flow curves. This allowed the crediting of the greater auxiliary feedwater flow which will be delivered to the steam generators at pressures lower than the main steam safety valve opening setpoint. The Appendix 158.6 feedwater line break analysis assumed that the limiting single failure occurred with offsite power available (the failure to fast transfer). The analysis presented in this section assumed that the loss of offsite power occurred coincident with the turbine generator trip. This assumption degrades the RCS flowrate to the largest extent possible which minimizes the heat transfer rate from RCS to the steam generator. In addition, this analysis assumes a single failure of one of the two safety related auxiliary feedwater pumps. 15.2.8.3.2 Input Parameters and Initial Conditions The initial conditions of this transient are the same as those discussed in CESSAR Appendix 15B.6.2.2 and listed in table 158-3 with the exception of the initial RCS and steam generator pressures. The initial RCS pressure assumed in the PVNGS analysis is 2300 psia instead of 2115 psia. As was done in CESSAR Appendix 15B, the initial RCS pressure was adjusted to achieve a coincident reactor trip on high pressurizer pressure and the conservatively delayed low level trip in the ruptured steam generator. The reason that the initial pressurizer pressure is different is due to the RCS metal heat transfer model in CESECIII. This model reduces the pre-trip RCS pressurization rate. Therefore, the initial RCS pressure must be raised in order to achieve the coincident trips. In addition to the initial RCS pressure increasing, the initial steam generator pressures increase sl,i ghtly from 1026 to 1038 psia. This slight increase is due to the increased RCS initial pressure. 15.2.8.4 Results See CESSAR Appendix 158. In addition, the dynamic behavior of the important NSSS parameters following the 0.2 sq ft feedwater line break with the loss of offsite power is presented in figures 15.2-1 through 15.2-8. The sequence of events provided in table 15.2-1 summarizes the important results of this event. A 0.2 sq ft rupture in the main feedwater line is assumed to instantaneously terminate main feedwater flow to both steam generators and establish critical flow from the generator nearest the break. This causes a decrease in steam generator inventory as shown in Figure 15.2-8. At 27.05 seconds coincident reactor trip conditions are reached on high pressurizer pressure and low steam generator water level in the ruptured steam generator. By 28.2 seconds the reactor tri p breakers open. Due to the rapid decrease in steam flow as a result of the turbine generator trip, the auxiliary feedwater actuation condition is reached in the intact steam generator at 29.4 seconds. By 73.1 seconds auxiliary feedwater reaches the intact steam generator as shown in figure 15.2-7.
'I Both steam generator pressures decrease gradually until the main steam isolation signal is generated at 241.2 seconds. As shown in figure 15.2-6, it is at this point that the pressure of the two steam generators begin to deviate significantly. Between this time and the isolation of auxiliary feedwater to the ruptured steam generator, auxiliary feedwater to the intact steam generator is assumed to equal 0 gpm. At 247 seconds the intact steam generator dries out and remains empty until auxiliary feedwater is reinitiated to it at 285.3 seconds. During the time when both steam generators are dried out there is a slight increase in the RCS temperature, pressure and pressurizer water volume as shown in figures 15.2.2, 1'5.2-4, and 15.2-5 until auxiliary feedwater is re-established. After the isolation of the ruptured steam generator, the intact steam generator's pressure gradually increases to the main steam safety valve setpoint. This causes the RCS temperatures, pressurizer level and pressurizer pressure to gradually increase. Ry 700 seconds .the main steam and pressurizer safety valves are maintaining primary and secondary pressures. By 900 seconds th 1 t is in a stable condition with the pressurizer level less than 1200 cubic feet and the intact steam generator liquid inventory at approximate 1 y 51000 ibm. This. analysis demonstrates that, even with the very severe restrictions associated with the feedwater line break analysis, the CESSAR deviations described in section 1.9.2.4.10 do not adversely impact long term RCS heat removal. 15.2.8.5 Conclusion See CfSSAR Appendix 15B. In addition, it has been determined that, for the full spectrum of feedwater line break sizes, and therefore all Chapter 15 transients, the PVNhS auxiliary feedwater system deviations described in section 1.9.2.4. 10 are acceptable with respect to longiterm RCS heat removal concerns.
0 Table 15.2-1 SEQUENCE OF EVENTS FOR A 0.2 SQ FT FEEDWATER LINE BREAK WITH THE LOSS OF OFFSITE POWER AND THE SINGLE FAILURE OF ONE AUXILIARY FEEOWATER PUMP Time sec Event Setpoint or Value 0.0 Rupture in the Main Feedwater Line, ft 0.2 O.O Complete Loss of Feedwater to Both Steam Generators O.O Initial Steam Generator Break Flow, ibm/sec 1987 27.05 Pressurizer Pressure Reaches Reactor Trip 2475 Analysis Setpoint, psia 27.05 Steam Generator Water Level Reaches Low Level Reactor Trip and Auxiliary Feedwater Actuation Signal Analysis Setpoint, Ruptured Generator (Heat Transfer degradation begins) 28.05 High Pressurizer Pressure Trip Signal Generated
28. 05 Auxiliary Feedwat'er Actuation Signal Generated 28.20 Reactor Trip Breakers Open 29.4 Steam Generator Water Level Reaches Auxiliary 10 Feedwater Actuation Signal Analysis Setpoint, Intact Steam Generator, 'f. wide range 31.2 Maximum RCS Pressure (uncorrected for Reactor 2615 Coolant Pump and elevation heads), psia 35.6 Ruptured Steam Generator Dries Out 73.1 Auxiliary Feedwater Initiated to Intact Steam Generator 240.2 Steam Generator Pressure Reaches Main Steam 810 Isolation Signal (MSIS) Analysis Setpoint, psia 241.2 MSIS Generated 245.8 MSIVs Completely Closed 247.7 Intact Steam Generator Dries Out e 269. 3 Difference Between Steam Generator Pressures Reaches Analysis Setpoint for Lockout of AFW to Ruptured Steam Generator, psid 270.3 Signal to Isolate AFW from Ruptured Steam Generator Generated.
Table 15.2-1 (Continued) SEQUENCE OF EVENTS FOR A 0.2 SQ FT FEEOWATER LINE 8REAK WITH THE LOSS OF, OFFSITE POWER ANO THE SINGLE FAILURE OF ONE AUXILIARY FEEOWATER PUMP 285.3 Isolation of Ruptured Steam Generator Compl eted, Auxi 1 i a ry Feedwater Re-establ i shed to Intact Steam Generator gno Pressurizer Water Volume at End of 1178 Simulation, cubic feet 900 Intact Steam Generator Liquid Inventory 51000 at End of Simulation, ibm
120 100 LLJ 80 60 - Q 40 20 0 0 180 360 540 720 900 TIME. SECONOS O.g ~ FT F'EEO~ATEg LINE BR~A% WITH THE LoSS aF GFFSiTE POmEA RNDT'HK Gt<GL~ Fichu~ L, FAILURE'F OH< AVXL(ARY FEECMIATEIZ PVYl P 1S.2- t CORE PowER VS Wt ME
0 0
670 o 630 610 ColPE ovr~E7 590 I- ~ CoRE AVENGE C.oRE 2NmT 570 550 0 180 360 540 720 900 TINE SECONDS AUXIN'AD.Y 0.2. ~ FT FKKDmAT'E~ UHE BREAK w>7'H THE LoSS OF OFF S IT'E- POWER AND T'HE 'S<MGC-< FAlLVRE'F Fig~ e. OH< FE'EChnlATER PVYl P J5.2 -2
'CoRE conu4N7 TempERAm<as vs t
20CCC 16CCC 1200C 4. 0 0 ~ 8600 (V'000 ~~ LOOP 4J <TH RV p'PJRED 8+Ca.~
~P Mi?H ZwTRcT QE,NER.R70tZ 0
18C 360 . 540 -
. 720 .900 T I NE.'Et.OMDS AUXILIARY 0.2. ~ FT FEEOmATER, UHE BRFAK WITH THE ~isa'iape LOSS OF OFFSl<E, POWER AMDT'HK Gtt4GLE. FattuRE OF OH< FE'EChnlRTE'R PVM P'CS l5. 2.- 3 LOOP FLQMS vs
2900 2700 2500 2300 2100 1900 1700 0 180 360 540 720 900 TINE SECONDS 0.2, ~ FT FE-ED~AT'E~ UHG BREA,K W]TH THE LOSS OF OFF S I TK POTE'g 4ND QH E Slhl g'LC, F4,LUQQ Figure. OF OH< AUX lL(ARY FE'GOD'AT+g PV~ P
>y Z~
RC+ t A.E'ssv<E vs ( /Me=
1' 18GC 16CC 14CC Z 120C 1COG 8CC 18C 36C ~4C ",20 9CC T I l'1" S":CCNt:S 0.2. ~ FT FEEO~ATER,, LINE BREAK w(7g THE'oSS OF OF FS t T'E. POWE'g RHD T'E SiNG(-< FAtLURE'F Figut L, ORE AUX'<(ARY FE'EDvVR7E'R POD P RCSSU<l'i ER RATER VO ~Urrlh VS Tl< c IS. 2. -S
1500 1250 100C Z~hc.T St.c.a.~ gp~~gq7-gg 750 Pv H VlZE9 ~4<a rn QE'N gZ+ l-og. 500 250 0 360 540 720 900 TII'lE. SECONDS 0.2. Gt FT FZED~ATER LIHF. BR<AK W'17H THg LaSs OF OFFSET TE, PoWE'R RN9+Hf S~agC.E, FALuRE OF OCR'VX L (ARY'E'EOvJS)Tgg, p0~ p rs. 2-C. Meum QegpazTo iz. PR,cssuRe
18C 1HC 1~O 9G F60 3CQ 18C 360 54C 72G 9GG. TII'1:> S"CONCS O.E ~ FT FKE,OMATER, LIHE BREA,H WiTH THE'oSS OF OFFSIT'E POv48R AMDT'HK S<MGL< FAILURE'F CN< A Vali-(AD FEEDvVATE'R PU& P
~~$ 9 + Qpgg m / Qt ~g)fh MiQ AU5 FEED F ow V5 I l ~F-
0 Q
18CCOC }iCGCC 1"CGCC BCGCC
'6GCCC Z~Tgt"~ GfC~ G KNEAD<~
pgpqgwU S te.arv) QGNERRToR 30000 0 180 360 540 ~n0 9GC T I i"!": ~ S":CONt'S O.E ~ FT FZE,OMAN'ER, UHG BRE4K WiTH THE LoSs OF OF FS t TE PoWE'g Pt,NDTHK StMGLc- FaimRE Figure. OF 0<6 AUXILIARY FE'~ChnlATER PcJYl P
~tEaM QEN&pqTo k J I au> g Volga'. vs
PVNGS FSAR 15.3. DECREASE IN REACTOR COOLANT FLOWRATE 15.3.1, TOTAL LOSS OF REACTOR COOLANT FLOW V Refer to CESSAR'.Section 15.3.1. Additionally, CIAS or SIAS signals will actuate control. room and fuel building essential ventilation systems. See sections 6.4 and 9.4 for details. CIAS oz SIAS will terminate the containment power access purge as described in section 9.4. The auxiliary feedwater system is described in section 10.4.9. '15.3.2 FLOW CONTROLLER MALFUNCTION CAUSING A FLOW COASTDOWN Refer to CESSAR Section 15.3.2. 15.3.3 SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOSS OF OFFSITE POWER 15,3.3.1 Identification of Event and Causes Refer to CESSAR Section 15.3.3.1. 15.3.3.2 Se ence of Events and S stem 0 eration Refer to CESSAR Section 15.3.3;2 except for the following: A. The CEAs begin to drop into the core and the turbine/ generator trip occurs at 1.165 seconds. B. The minimum transient DNBR of 0.808 occurs at 1.3 seconds. C. The loss of offsite power occurs at 4.165 seconds. rZi s~r+ '>
PVNQS FSAR W ~ 15.3 DECREASE IN REACTOR COOLANT FLOWRATE 15.3 1~ TOTAL LOSS OF REACTOR COOLANT FLOW Refer to CESSAR..Section 15.3.1. Additionally, CIAS or SIAS signals will actuate control room and fuel building essential ventilation systems. See sections 6.4 and 9.4 for details. CIAS or SIAS will terminate the containment power access purge as described in section 9.4. The auxiliary feedwater system is described in section 10.4.9. 15.3.2 FLOW CONTROLLER MALFUNCTION CAUSING A FLOW COASTDOWN Re fer to CESSAR Section 15.3.2. 15.3.3 SINGIE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOSS OF OFFSITE POWER 15.3.3.1 Identification of'vent and Causes Refer to CESSAR Section 15.3.3.1. 15.3.3.2 Se ence of Events and S stem 0 eration Refer to CESSAR Section 15.3.3;2 except for the following: A. The CEAs begin to drop into the core and the turbine/ generator trip occurs at 1.165 seconds. B. The minimum transient DNBR of 0.808 occurs at 1.3 seconds. C. The loss of offsite power occurs at 4.165 seconds. rZN 5 c.c-+
Insert "8" D. With respect to the single reactor coolant pump rotor seizure analysis, the PVNAS auxiliary feedwater system deviates from the CESSAR assumed response in that the auxiliary feedwater flow delivery has been reduced>from 875 to 750 gpm per pump. See section 1.9.2.4.10 for a discussion of how this deviation does not affect the conclusions of the CESSAR Safety Analysis.
PVNGS FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.8.2 Se ences of Events and S stem 0 eration Refer to CESSAR Section 15.4. 8. 2. Additionally, it is postu-lated that the containment power access purge is in operation at start of this event. The purge will be terminated within 5 seconds after either CPIAS or CIAS is generated, as discussed in section 6.2.4. Dual ESF grade area radiation monitors are provided at the outside face of the containment adjacent to the ducting for the power access purge exhaust. These monitors will initiate CPIAS should a significant amount of radio-activity be purged prior to the generation of CIAS at 40.2 seconds. Monitor response time to purged containment radioactivity is approximately 1 second. Therefore, the power access purge will be terminated by CPIAS 6 seconds after significant amounts of radioactivity begin to be purged from the containment, or by CIAS at 46 seconds after event initia-tion, whichever is sooner. Since the power access purge fans, filters, and ducting outside the containment are, considered non-safety'related, the follow-ing system operation has been assumed coincident with event initiation: The supply fan and ducting remains intact and in operation. The exhaust ducting suffers a rupture outside of the auxiliary building. As a result of these assumptions, the purge is continued, but, HEPA and charcoal filtration by either the purge exhaust filter unit or the auxiliary building exhaust unit is not ass Should there be an appreciable increase in containment pressure, CSAS would be generated to initiate containment sprays. How-ever, in assessing the radiological consequences of this event, no credit has been taken for iodine removal by sprays. The auxiliary feedwater system is described in section 10.4.9. Amendment 10 15.4-2 December 1982
PVNGS FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES The CIAS or SIAS signals will actuate control room and fuel building essential ventilation systems. See sections 6.4 and 9.4 for details. Containment combustible gas control is manually initiated, if required for the event, as described in section 6.2.5. 15.4.8.3 Anal sis of Effects and Conse uences Refer to CESSAR Section 15.4.8.3 for information other than radiological consequences. Although it is unlikely that the entire radioactivity noted in CESSAR Table 15.4.8-6 would be instantly released from the core and the RCS, measures have been incorporated in the PVNGS design to keep offsite doses well below 10CFR100 limits should such a release ace. The assumptions u x ized in the analysis of this event, noted in section 15.4.8.2, rovide that the ower s urge is promptly terminate Credit for iodine removal by sprays has not been assumed. Leakage from recirculation components outside the containment, as well as O.l volume % leakage from the containment for the first 24 hours, has been assumed. Table 15.4-1 presents the estimated offsite doses at the exclusion area and low population zone boundaries. 15.4.8.4 Conclusions Refer to CESSAR Section 15.4.8.4 for non-radiological consequences. Should there be a rupture of a CEDM housing that ejects a CEA, 2-hour EAB doses will be less than rem O cg. inhalation thyroid and MQ rem whole body gamma. Thirty day LPZ doses will be less than rem inhalation thyroid and
o. I
+-.:~rem whole body gamma. These exposure's are well within 10CFR100 limits and assume no credit for various available protective features as described in section 15.4.8.2. December 1982 15.4-3 Amendment 10
PVNGS FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4-1 RADIOLOGICAL CONSEQUENCES OF A CONTER) ELEMENT ASSEMBLY EJECTION ACCIDENT
. Thyroid Dose Whole Body Dose Dose (rem) (-rem)
Exclusion area boundary 2-hour consequences 10I Containment leakage 29' 0.10 Recirculation leakage 0.12 2.7 (-4)
~ ~
10l Secondary release $ .9 Total u, g. Low population zone 30-day consequences 12I 1oI Containment leakage 60.5 6.9 (-2) Recirculation leakage 0.28 2.4 (-4) 10I Secondary release P '~ o.S I.l(4 Total o. og
a. Assumes no credit for containment sprays or non-ESF filtration.
Amendment 12 15.4-4 February, 1984
PVNGS F SAR 15.6 DECREASE IN REACTOR COOLANT SYSTEM INVENTORY 15.6.1 INADVERTENT OPENING OF A PRESSURIZER SAFETY/RELIEF VALVE Refer to CESSAR Section 15.6.1. 15.6-2 DOUBLE-ENDED BREAK OF A LETDOWN LINE OUTSIDE CONTA I NMENT Refer to CESSAR Section 15.6.2. Additionally, CIAS or SIAS signals will actuate control room and fuel building essential ventilation systems. See sections 6.4 and 9.4 for details. A CIAS or SIAS will terminate the containment power access purge, as described in section 9.4. The auxiliary feedwater system is described in section 10.4.9. t 15.6.3 Refer to 0 STEAM GENERATOR TUBE RUPTURE CESSAR Section 15.6.3. Additionally, CIAS or SIAS signal will actuate control room and fuel building essential ventilation systems. See sections 6.4 and 9.4 for details. A CIAS or SIAS will terminate the containment power access purge, as described in section 9.4. The auxiliary feedwater system is described in section 10.4.9. The 0 to 2 hour X/Q value for PVNGS is less than that used in CESSAR. Accordingly, refer to CESSAR Sections 15.6.3.1.3.2 and gay" 15.6.3.2.3.2 for conservative radiological dose conclusions. g~ R ) 15.6.3.1 Steam Generator Tube Ru ture With a Loss of Offsite Power And Sin le Failure 15.6.3.1.1 Identification of Event and Causes Refer to CESSAR Section 15D.1. Additionally, it is assumed that the atmospheric dump valve (ADV) of the affected steam generator remains open throughout the transient, since PVNGS does not have block valves. February 1984 15.6-1 Amendment 12
Insert "Z" With respect to the steam generator tube rupture analysis, the PVNGS auxiliary feedwater system deviates from the CESSAR assumed response in that the auxiliary feedwater flow delivery has been reduced from 875 to 750 gpm per pump. See section 1.9.2.4.10 for a discussion of how this deviation does not affect the conclusions of the CESSAR safety Analysis.
PVNGS F SAR DECREASE IN REACTOR COOLANT INVEN 15.6.3.1.2 Sequence of Events and Systems Operation This section is the same as CESSAR Section 15D.2 th the following differences: A. The operator opens one ADV in each earn generator in order to cool the RCS to 550F at e PVNGS plant procedure specific maximum cooldown rate f 75F/h. This cooldown rate translates into a 10.5% 'pening of one ADV of each steam generator. B. When the RCS temperature 's below 550F the operator unsuccessfully attempts o close the ADV of the affected steam generator. The V of the affected steam generator is assumed to remain pen for the remainder of the transient. 12 C. The operator initi es auxiliary spray flow in order to regain pressur' evel two minutes after attempting to close the a ec ed V. D. A cooldown te of 75F/h is assumed until 30 minutes after the ttempted closing of the affected steam generato ADV. After this period a cooldown rate of 30F/h 'ssumed until shutdown cooling entry conditions are r ached. E. Sin e steam is being continuously released through the s ck open ADV, concern regarding affected steam gen-ator overfill due to the primary to secondary leak does not arise. Table 5.6-A1 presents a chronological sequence of events which occu during the steam generator tube rupture event with a loss of ffsite power and stuck open ADV, from the time the operator
12) Amendment 12 15.6-1A February 1984
PVNGS FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6-A1 SEQUENCE OF EVENTS FOR A STEAN GENERATOR E RUPTURE WITH A LOSS OF OFFSITE POWE AND STUCK OPEN ADV Time (sec) Event Setpoj. or /Clue t Success Path 460 Operator Initiates Plant Reactor Heat Cooldown by Opening Removal One ADV on each SG 546 Pressurizer Empties 570 Safety Injection Act a- 1578 :Reactivity tion Signal, psia Control 620 Safety Injection F ow ';Reactivity Initiated Control 2700 Operator Attemp to 550 'Secondary Isolate the D aged System Generator, RC Integrity Temp., (F) 2820 Operator I tiat Primary System Auxili y Spray Flow Inventory 3400 Operat r Controls 20 Primary System Au liary Spray Flow, Integrity B kup Pressurizer eater Output, and SI Flow to Reduce RCS Pressure and Control Subcooling (F) 2 8,8 OIShutdown Cooling Entry 400/350 Reactor Heat Conditions are Assumed Removal to be Reached, RCS I Pressure, psia/Temp., (F) T sequence of events for the first 460 seconds rs the s e as that presented in Table 15D-1 of CESSAR. e February 1984 15.6-1B Amendment 12
PVNGS FSAR DECREASE IN REACTOR COOLANT INVENTORY takes control. Prior to operator in ervention the sequence of events is the same as that presented in Table 15D-1 of CESSAR. The major operator actions are listed in figure 15.6-1. 15.6.3.1.3 Analysis of Effects and Consequence 15.6.3.1.3.1 Core and S stem Performance A. Mathematical Model Refer to CESSAR Section 15D.3. .A. B. Input Parameters and Initi Conditions Refer to CESSAR Section D.3.1.B. C ~ Results The dynamic response of the plant parameters for the first 460 seconds o the transient, i.e., from the time of rupture un il the operator takes control of the plant, is the me a that described in CESSAR Section 15D.3.1, . The ynamic behavior of important NSSS paramete s after operator intervention will be presented i figures 15.6-2 to 15.6-15 (to be provided). At 460 se onds the operator takes control of the plant and ope one ADV on each SG to cool down the plant. At 270 seconds the RCS has been cooled to 550 F. The opera / or isolates the auxiliary feedwater to the affe 'ted generator, closes the main steam isolation valises of both steam generators, and attempts to close th ADV of the affected generator. The operator r cognizes that the ADV did not close. The ADV of the steam generator remains open for the remainder 'fected of the transient. The operator then initiates an orderly cooldown and depressurization using the atmospheric dump valves, auxiliary feedwater flow to the unaffected steam generator, pressurizer Amendment 12 15.6-1C February 1984
PVNGS FSAR DECREASE IN REACTOR COOLANT INVENTORY auxiliary sprays, pressurizer heaters and safety injection flow. The maximum RCS and secondary pressures do not exce 110% of design pressure following a steam gener or tube rupture event with a loss of offsite piner and stuck open ADV, thus, assuring the integr'ty of the RCS and the main steam system. t A 460 seconds, when operator actioo's assumed, no more than 41,500 1bm of steam from the damaged steam generator and 41,470 ibm from e intact steam generator are discharged via the main s earn safety valves. Also, during the same time period pproximately 17,560 ibm of primary system mass is lea ed to the damaged steam generator. Subsequently, the operator begins a plant cooldown at the PVNGS pl nt procedure specific maximum cooldown rate (75F/h) u oth steam generators, the atmospheric dump. al , and the emergency feed-water system. Thj, ty minutes after attempting to isolate the affected steam generator, it is assumed that the operator educes the cooldown rate to 30F/h. For the first tw hours following the initiation of the n event, 488,,~r000 ibm of steam are released to the environ-thro'ugh the atmospheric dump valves. For the two-4'ent to eight,-hour cooldown period, an additional 982,000 ibm of ste are released via the atmospheric dump valves.
/'5.6.3.1.3.2 Radiolo ical Conse ences A. Ph sical Model T e physical model is the same as that discussed in ESSAR Section 15D.3.2 with the following exception.
After 460 seconds, the operator initiates a plant cooldown at the PVNGS plant procedure specific maximum cooldown rate of 75F/h. Thirty minutes after attempting Februar 1984 15.6-1D Amendment 12
PVNGS FSAR DECREASE IN REACTOR COOLANT INVENTORY to isolate the affected steam generator, the operator lowers the cooldown rate to 30F/h. B. Assumptions and Conditions The assumptions and conditions employed f the evalua-tion of radiological releases are the c~same as those discussed in CESSAR Section 15D.3.2<B. with the excep-tions of assumptions 7 and 10. They are:
7. During the period whe the water level in the affected steam generator is above the top of e U-tubes,'hat l
portion of the aking primary fluid which flashes t steam upon entering the steam gen e ator is assumed to be t released to e atmosphere with a decontaminat'he factor (DF) of 1.0. porti e leaked fluid that does n flash, mixes with the liquid in t steam generator and is released to he atmosphere with a DF of 100. D ring the period when the water level s below the top of the U-tubes, it is assumed that all the leaking primary fluid escapes to the atmosphere with a DF of 1.0. No credit is taken for the presence of steam separators and dryers which would retain a part of the escaping primary liquid in the steam generator. b uar 1984
PVNGS FSAR DECREASE IN REACTOR COOLANT INVENTORY
10. The atmospheric dispersion factors employed in the analyses are: 3.1 x 10 sec/m 3 for the exclusion area boundary and 5.1 x 10
-5 sec/m 3 for the low popula-tion zone.
C. Mathematical Model Refer to CESSAR Section 15D.3.2.C. D. Results The two-hour exclusion area by ndary (EAB) and the eight-hour low population z he (LPZ) boundary inhalation doses for both the genera d iodine spike (GIS) and the pre-existing iodine ike .(PIS) are presented in table 15.6-A2. The ca ulated EAB and LPZ doses are well within the accep ance criteria. Ta le 15.6-A2 RADIOLOGICAL CONSEQ NCES OF THE STEAM GENERATOR TUBE RUPTURE WI LOSS OF OFFSITE POWER TU OPEN ADV Offsite Dose (rem) cation GIS PIS
l. Excl sion Area Boundary 0-2 hr. Thyroid 116 148

2. L Population Zone Outer 222 121 undary 0-8 hr. Thyroid 5.6.3. .4 Conclusions

he ra 'ological releases calculated for the SGTR event with a
"-oss o offsite power and a stuck open ADV are within the OCFR 00 guidelines. The RCS and secondary system pressures
PVNGS FSAR DECREASE IN REACTOR COOLANT INVENTORY are well below the 110% of the design pressure limits, thus, assuring the integrity of these systems. Additionally, no violation of the fuel thermal limits occurs, since the minimum DNBR remains above the 1.19 .value throughout the duration of the event. 15.6.3.1.5 References Refer to CESSAR Section 15D.5. In addition, add the following reference: Palo Verde Nuclear Generating Station Manual, "Steam Generator Tube Leak", Procedure No. 41RO-1ZZ06, Revision 0. 15.6.4 RADIOLOGICAL CONSEQUENCES OF MAIN STEAM LINE FAILURE OUTSIDE CONTAINMENT (BWR) Not applicable. bruar 1984
Delete Figures 15.6-1 thru 15.6-25 SGTR with partially stuck open ADV.
PVNGS FSAR DECREASE IN REACTOR COOLANT INVENTORY The power access purge will actually be terminated before SIAS. CPIAS will be promptly generated by the power access purge area radiation monitors as soon as significant quantities of radioactivity are purged. Thus, containment purge releases are minimized
~~
0 j10
~ Q-gg~~~Q~QII~Q'/de 7QSI~
2 IZ/, g4
~
15.6.5.3
~ ~ ~ Anal sos o ~
ec s an Conse ences
~ ~~~ /p~~ ~P~g )10 Refer'o CESSAR Section 6.3.3 for information ot er ra xo-logical consequences. Section 15.4.8 presents an ana yszs o the, effects and consequences of the bounding radiological release through the power access purge. The bounding case has the same isolation characteristics as the small break LOCA d it has a higher release concentration Section 15.6.5.6 t10 presents the analysis of the efX'ects and consequences of the large break LOCA bounding radiological release through contain-ment and recirculation leakage or through secondary releases.
The large break LOCA release is bounding as accident source terms are equal to or greater than those predicted for a small break LOCA. Offsite dose consequences of a small break LOCA are summarized in table 15.6-1. The consequences are well within 10CFR100 limits. 15.6.5.4 Identification of Event and Causes - Lar e Break LOCA [10 Refer to CESSAR Section 6.3.3. 15.6.5.5 Se ence of Events and S stems 0 eration - Lar e )10 Break LOCA Refer to CESSAR Section 6.3.3. Containment power access purge through an 8-inch penetration will be terminated within econds 'after generation of CIAS or SIAS as described in section 6.2.4. December 1982 1'5. 6-3 Amendment 10
PVNGS FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6-1 RADIOLOGICAL CONSEQUENCES OF A SMALL-BREAK LOCA
'hyroid Dose Whole Body Dose Dose (rem) (rem)
Exclusion area boundary 2-hour consequences Power access purge Containment leakage and secondary release 104 1.7 Recirculation leakage 0.12 2.7 (-4) Total IOQ ~~}- l. 7 Low population Zone 30 day consequences Power access purge 0$ Containment leakage and secondary release 155.1 0.8 Recirculation leakage 0.28 Total ling
a. Assumes no credit for non-ESF filtration or for spray removal of particulates or organic iodine.
Amendment 12 15.6-4 Februa
PVNGS FSAR DECREASE IN REACTOR COOLANT INVENTORY representative large break LOCA, SIAS is generated 9.43 seconds after event. initiation. Thus, the purge will be terminated no later than 35 seconds after the event starts. ss noted in CESSAR Section 6.3.3.2.6, clad rupture will not occur until reflood. CESSAR Table 6.3.3.2-1 indicates that the earliest instance of reflood for the spectrum of large breaks will not begin until at least 20.8 seconds a'fter event initiation. Therefore, the purge is isolated before the earliest possible rupture is postulated to occur.
,SIAS will generate a control room essential filtration actuation signal (CREFAS) and a fuel building essential ventilation actuation signal (FBEVAS). CREFAS will initiate a switch to the filtered recirculation and filtered makeup mode of control room ventilation as discussed in section 6.4. A FBEVAS will initiate filtered ventilation of the lower region (below 100'l) of the auxiliary, building as discussed in ~
section 9.4. Since recirculation loop equipment and piping
~ ~ ~ ~
for safety injection and containment sprays in the auxiliary
~
building is located below the 100 foot elevation, leakage from active recirculation equipment is filtered prior to release to the environment. 15.6.5.6 Anal sis of Effects and Conse ences Lar e )~0 It is assumed
~su that< ~cL Qg x M~A(4AQ loO~/~
maes~el
~
instantaneously mixed with the containment atmosphere and available for release via the power access purge. :ace tahe
/>0 The containment airborne radioactivity inventory will be affected by three factors: leakage, radioactive decay, and sprays.
e No credit has been taken for spray removal of organic iodine or particulates. Refer to section 6.5 for a discussion of spray effectiveness. It is assumed that the containment leaks December 1982 15.6-5 Amendment 10
PVNGS FSAR DECREASE IN REACTOR COOLANT INVENTORY at the maximum rates allowed by the technical specifications, i.e., 0.1 vol %%uJ'd for the first 24 hours and half of that rate thereafter. This leakage, when combined with initial releases will result in potential doses offsite, and in the control room. These doses are listed in table 15.6-2. Additionally, there will be doses offsite and in the control room from the filtered release of recirculation leakage. The calculated leakage is based on the containment sump inventory listed in CESSAR Table 6.3.3.6-1. The doses from recircula-tion releases are listed in table 15.6-2. The total combined doses to an individual offsite and to control room operators following a postulated large break LOCA are also presented in table 15.6-2.
10) 15.6.5.7 Conclusions Even with the use of very conservative assumptions regarding spray effectiveness, containment ventilation, and leakage, as well as conservative fuel failure*models, the offsite doses presented in table 15.6-2 for LOCA are substantially below 10CFR100 limits.
]0I Amendment 10 15.6-6 December 1982 0
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ML17298B471

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ML17298B471

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