ML20064C147
| ML20064C147 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 09/30/1978 |
| From: | Zull L GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20064C138 | List: |
| References | |
| NEDO-24129-1-S01, NEDO-24129-1-S1, NUDOCS 7810170114 | |
| Download: ML20064C147 (12) | |
Text
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NEDO-24129-1 Supplement 1 Class 1 September 1978 RAISED SAFETY / RELIEF VALVE SETPOINT REANALYSIS FOR THE I
JAMES A. FITZPATRICK NUCLEAR POWER PLANT FOR RELOAD NO. 2 Prepared by:
L. M.
ull Sr. Licensing Engineer
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Approved by: f. ((#
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.O W p-I i
R. O. Brugge, anager Operating Licenses 11 NUCLE AR ENE RGY PROJECT 5 Dtvis10%. GENE R AL E LECTRIC Cow'ANY SAN JOSE. CALIFORNI A 9512s W
GENER AL h ELECTRIC
NEDO-24129-1 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for the Power Authority of the State of New York (The Authority) for The Authority's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending The Authority's operating license of tne James A. FitzPatrick Nuclear Power Plant.
The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared.
I The only undertakings cf the General Electric Company respecting information in this document are contained in the contract between The Authority and General Electric Company (GE letter G-EP1-3-121 dated August 11, 1978 and The Authority's P.O. NY0 6-77-20) and nothing contained in this document shall be construed as changing said contract.
The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.
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NE00-24129-1 1.0 Introduction and Summary One event that has a significant impact on Boiling Water Reactor (BWR) availability is the spurious openi.,g or failure to reclose of the dual function safety / relief valves. As described in Reference 1, the event from a safety standpoint has a relatively minor effect on the reactor core and reactor coolant pressure boundary.
- However, the event can result in a significant maintenance outage since the reactor must be shutdown, depressurized, and the valve repaired or replaced ba' ore the plant can be returned to service.
The cause of the majority of these spurious openings or failures to reclose of safety / relief valves is excessive leakage around the setpoint pilot valve.
Other causes of valve failures have been identified and corrective action has been taken. Operating data demonstrate that an increase in valve simmer margin (the differen-tial pressure between the valve setpoint and normal system operating I..
pressure at the valve) will reduce the probability of valve failure due to pilot leakage.
The Reload 2 licensing supplement (Reference 2) presented the results of the safety analysis for Cycle 3 using the following safety / relief valve (S/RV) groupings and setpoints:
2 @ 1090 psig + 1%
9 @ 1115 psig + 1%
Subsequent calculations performed to address the Commission's multiple safety / relief valve actuations request (Reference 3) indicated that multiple actuations could be minimized, while obtaining the benefit of an increased simmer margin, by using the following revised valve groupings and setpoints:
2 @ 1090 psig + 1%
2 @ 1105 psig + 1%
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7 @ 1140 psig + 1%
This document provides the results of a reanalysis of the h:ost limiting thermal and pressurization transients using the above revised S/RV setpoint groupings. The results indicate that the operating limits for 8x8R fuel presented in the Reload 2 licensing submittal (Reference 2) are still applicable. The operating limits for 7x7 and 8x8 fuel decreased by 0.01 between EOC3-2 Gwd/t and EOC3-1 Gwd/t.
The operating limit for 8x8 fuel is also reduced by 0.01 between EOC3-1 Gwd/t and EOC3.
The analysis also indicates that a 103 psig margin to the ASME vessel code limit of 1375 psig exists for the most severe overpressurization event, an MSIV closure with flux scram at EOC3.
These results indicate that additional safety / relief valve reliability is obtained without imposing additional restrictions on plant operation.
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NED0-24129-1 2.0 Safety Analysis 2.1 Introduction The safety analysis for FitzPatrick Reload 2 is provided in Reference 2.
The raising of the safety / relief valve setpoints only affects those events which result in valve operation to limit system pressure. The limiting events which require reanalysis are the most severe pressurization transient (generator load rejection with failure of the bypass valve), vessel overpressure protection analysis (closure of all main steam line isolation valves - flux scram) and loss-of-coolant accident (small break).
In addition, the capability of the reactor core isolation cooling (RCIC) and high pressure coolant injection (HPCI) systems were reevaluated for the higher safety / relief valve setpoints.
The results of the analysis which demonstrate the acceptability of the increased simmer margin are given in Sections 2.2 through 2.5.
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All analyses were performed using the same input parameters as documented in Reference 2 with the exception of safety / relief valve setpoints and capacity, which were as follows:
Valve No. of Setpoint ASME Capacity
- at Setpoint Group Valves (psig)
(Per Valve, 105 lbm/hr) 1 2
1090 psig + 1%
7.986 2
2 1105 psig + 1%
8.091 3
7 1140 psig + 1%
8.343
- Capacities include the 2.3% reduction due to use of Schedule XXS inlet piping.
The use of the maximum S/RV setpoint tolerance of +1% results in valve actuations at 1101 psig (Group 1); 1116 psig (Group 2';
and 1151 psig (Group 3) in the analysis.
The total capacity of all 11 S/RVs at the lowest setpoint is 84.2% of NBR steam flow.
2.2 Generator Load Rejection With Bypass Failure This transient produces the most severe reactor isolation event during Cycle 3.
Fast closure of the turbine control valves is initiated whenever electrical grid disturbances occur which result in a significant loss of generator load.
The turbine control valves close rapidly to prevent overspeed of the turbine generator rotor. This closing, concurrent with the failure of the bypass valve system, causes a sudden reduc-tion in steam flow which results in a nuclear system pressure increase. The pressure increase causes a significant void reduction which yields a pronounced positive void reactivity effect. The net reactivity is sharply positive and causes a rapid increase in neutron flux until the net reactivity is forced negative by a scram initiated from closure of the 2
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NE00-24129-1 turbine control valves and by a void increase after the safety /
relief valves have automatically openec on high pressure.
The results of the load rejection analyses are given in Table 1 and shown in Figures 1 and 2.
The peak vessel and steamline pressure are both approximately 10 psig above the results reported in the Reload 2 licensing submittal, but are still well below the ASME vessel code limit of 1375 psig.
In regard to the fuel thermal margins, the load rejection without bypass event determines the MCPR operating limit from E0C3-2GWd/t to EOC3.
The GETAB transient analysis results indicate a decrease in the change in critical power ratio (ACPR) such that the operating limits for 7x7 and 8x8 fuel are reduced by 0.01 between EOC3-2GWd/t and EOC3-1 Gwd/t.
The operating limit for 8x8 fuel is also reduced by 0.01 between E0C3-1 Gwd/t and EOC3. The analysis also indicates that the operating limits presented in the Reload 2 licensing submittal g
(Reference 2) for 8x8R fuel are still applicable.
From BOC3 to E0C3-2GWd/t the rod withdrawal error transient determines the operating limit for 7x7 and 8x8 fuel, and the loss of 80 F feedwater heating transient determines the operating limit for 8x8R fuel. These transients are not affected by the revised safety / relief valve setpoints since the safety / relief valves are not actuated during these tran-sients.
Therefore, the operating limits for B0C3 to E0C3-2GWd/t presented in the Reload 2 licensing submittal (Reference 2) are still applicable. The MCPR operating limits for Cycle 3 are summarized in Table 2.
2.3 Vessel Overpressure Protection Analysis I
The pressure relief system must prevent excessive overpres-(
surization of the primary systen process barrier and the pressure vessel to preclude an uncontrolled release of fission products.
The James A. FitzPatrick Plant pressure relief system includes 11 dual function safety / relief valves located on the main steamlines within the drywell between the reactor vessel and the first isolation valve.
These valves provide the capacity to limit nuclear system overpressurization.
The ASME Boiler and Pressure Vessel Code requires that each vessel designed to meet Section III be protected from the l
consequences of pressure in excess of the vessel design i
pressure:
A peak allowable pressure of 110% of the vessel design (a) pressure is allowed (1375 psig for a vessel with a design l
pressure of 1250 psig).
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NEDO-24129-1 (b) The lowest qualified safety valve setpoint must be at or below vessel design pressure.
The highest safety valve setpoint must not be greater (c) than 105% of vessel design pressure (1313 psig for a 1250 psig vessel).
The James A. FitzPatrick Plant safety / relief valves are set to 1090, 1105 and 1140 psig, satisfying (b) and (c) self-actuate at Requirement (a) is evaluated by considering the most severe above.
The safety /
isolation event with failure of the direct-scram trip.
relief valves are assumed to activate in their safety valve mode of operation.
The event which satisfies this specification is the closure of all The main steamline isolation valves with indirect (flux) scram.
results of the analysis of this event using the revised safety /
b relief valve setpoints are given in Table 1 and shown in Figure 3.
An abrupt pressure and power rise occur as soon as the reactor is Reactor shutdown is initiated when the neutron flux isolated.
reaches the 120% high flux scram setpoint.
The safety / relief valves open to limit the pressure rise at the bottom of the vessel to 1272 psig. This response provides a 103 psi margin to the vessel code limit of 1375 psig. Thus, requirement (a) is 5:tisfied and adequate overpressure protection is provided by the pressure relief system.
2.4 Loss-of-Coolant Accident Analyses The revised safety / relief valve setpoints have no significant effect on the large break LOCA analysis results. This is because the system depressurizes so rapidly from the break that the safety / relief valves are not actuated.
k A new analysis of the previously limiting small break was performed.
The small break models described in Reference 4 were used in the analysis. The results of the analysis showed a 33*F decrease in the calculated peak cladding temperature 2 recirculation (PCT) for the most limiting small break (0.07 ft suction line break with HPCI f ailure) from the 1285*F PCT reported in Reference 5.
The small break LOCA PCT is therefore reduced from 1285*F to 1252*F. The reason for this decrease is that with the revised S/RV setpoints the vessel depressurizes faster during the small break LOCA, allowing the Core Spray and LPCI systems to come on earlier, resulting in earlier cladding cooling and a lower PCT.
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NE00-24129-1 2.5 HPCI and RCIC Capability One of the design requirements for the HPCI and RCIC systems is that they be capable of providing design flow at the lowest safety / relief valve setpoint. These systems still meet the design requirement of full flow discharge to the core at 1120 psig with the increase in the lowest safety / relief valve setpoint to 1090 psig + 1%.
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3.0 References 1.
James A. FitzPatrick Nuclear Plant, Final Safety Anelysis Report, Docket 50-333, November 1971.
2.
" Supplemental Reload Licensing Submittal for the James A. FitzPatrick Nuclear Power Plant for Reload No.
2,"
June 1978 (NEDD-24129).
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3.
Letter, Victor Stello (NRC) to Licensees, " Multiple-Subsequent Actuations of Safety / Relief Valves Following an Isolation Event," dated March 20, 1978.
4.
" Loss-of-Coolant Accident Analysis Report for the James A. FitzPatrick Nuclear Power Plant (Lead Plant),"
July 1977 (NEDO-21772-2).
5.
" Supplemental Licensing Submittal for the James A. FitzPatrick Nuclear Power Plant for Reload-1 Cperation Between EOC2-Gwd/t and EOC2," March 1978 (NEDO-21619-1).
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m TABLE 1 FITZPATRICK CYCLE 3 EVENT DATA
SUMMARY
Core Power Flow Q/A P
P 3t y
(%)_
E
{%)
(%) (psig) (psig) Response Event load Rejection - No Bypass, Trip Scram from EOC3-2GWd/t to EOC3-1GWd/t Using Previous S/RV Setpoints (NE00-24129) 104 100 320 113 1174 1221 Figure 8 (NEDO-24129)
Using Revised S/RV Setpoints (This Analysis) 104 100 320 112 1185 1232 Figure 1 h
From EOC3-1GWd/t to EOC3
?
Using Previous S/RV Setpoints (NEDO-24129) 104 100 376 115 1178 1225 Figure 12 (NEDO-24129)
%C 3
104 100 376 114 1187 1235 Figure 2 Using Revised S/RV Setpoints (This Analysis)
MSIV Closure, Flux Scram, E003 104 100 1217 1264 Figure 16 (NEDO-24129)
Using Previous S/RV Setpoints (NEDO-24129) 1226 1272 Figure 3 104 100 Using Revised S/RV Setpoints (This Analysis)
P
- Peak suamh Pnssun h
- Peak Neutron Flux (% Initial) 3t P
- Peak Vessel P nssu n
[)/A-PeakHeatFlux(% Initial) y
TABLE 2 FITZPATRICK CYCLE 3 OPERATING LIMITS WITH REVISE 0* S/RV SETPOINTS i
MCPR Operating Limit **
Exposure Range 7x7 8x8 8 x BR From BOC3 to EOC3-2GWd/t 1.21 1.22 1.20 From E003-2GWd/t to EOC3-1GWd/t 1.25 1.33 1.33 From E003-1GWd/t to EOC 3 1.30 1.37 1.37 w
d
- The valve groupings and setpoints are:
2 @ 1090 psig, 2 @ 1105 psig, and 7 @ 1140 psig.
- The Safety Limit MCPR is 1.07.
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f ATTACHMENT A Power Authority of the State of New York s
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License No. DPR-59 l
Docket No. 50-333 l
PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS
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S JAFNPP 1.2 REACTOR COOLANT SYSTEM 2.2 REACTOR COOLANT SYSTEM APPLICABILITY:
APPLICABILITY:
Applies to limits on reactor coolant Applies to trip settings of the instru-system pressure, ments and devices which are provided to prevent the reactor coolant system safety limits from being exceeded.
OBJECTIVE:
OBJECTIVE:
To establish a limit below which the To define the level of the process integrity of the Reactor Coolant System variables at which automatic protective is not threatened due to an overpressure action is initiated to prevent the safety condition.
limits from being exceeded.
SPECIFICATION:
SPECIFICATION:
1.
The reactor coolant system pressure 1.
The Limiting Safety System setting shall not exceed 1,325 psig at any shall be specified below:
time when irradiated fuel is present in the reactor vessel.
A.
Reactor coolant high pressure scram shall be 5 1,045 psig.
B.
Reactor coolant system safety / relief valve nominal settings shall be as follows:
Safety / Relief Valves 2 valves at 1090 psig 2 valves at 1105 usig 7 valves at 1140 usio The allowable setpoint error for each safety / relief valve shall be + 1 percent.
27 Amendment No. [2, [
JAFNPP 1.2 and 2.2 BASES The reactor coolant pressure boundary ANSI Code permits pressure transients up to integrity is an important barrier in 20 percent over the design pressure (120% x the prevention of uncontrolled release 1,150 = 1,380 psig). The safety limit of fission products. It is essential pressure of 1,375 psig is referenced to the that the integrity of this boundary lowest elevation of the Reactor Coolant System.
be protected by establishing a pres-sure limit to be observed for all The analysis in NEDO-24129, " Supplemental Reload operating conditions and whenever Licensing Submittal for the James A. FitzPatrick there is irradiated fuel in the Nuclear Power Plant for Reload No.
2", June 1978, reactor vessel.
as amended by NEDO-24129-1, Supplement 1, September 1978, shows that the main steam isolation valve The pressure safety limit of 1,325 psig transient, when direct scram is ignored, is the as measured by the vessel steam space most severe event resulting directly in a reactor pressure indicator is equivalent to coolant system pressure increase. The reactor 1,375 psig at the lowest elevation of vessel pressure code limit of 1,375 psig, given the Reactor Coolant System. The in FSAR Section 4.2, is above the peak pressure 1,375 psig value is derived from the produced by the event above. Thus, the pressure design pressures of the reactor pres-safety limit (1,375 psig) is well above the peak sure vessel and reactor coolant system pressure that can result from reasonably expected piping. The respective design pressures l
overpressure transients. Figure 3 in NEDO-24129-1 are 1250 psig at 5750F for the reactor presents the curve produced by this analysis.
vessel, 1148 psig at 568 F for the Reactor pressure is continuously indicated in 0
recirculation suction piping and 1274 the control room during operation.
psig at 5750F for the discharge piping.
The pressure safety limit was chosen A safety limit is applied to the Residual Heat as the lower of the pressure tran-Removal system (RHRS) when it is operating in sients permitted by the applicable the shutdown cooling mode. When operating in design codes: 1965 ASME Boiler and the shutdown cooling mode, the RHRS is included Pressure Vessel Code,Section III for in the reactor coolant system.
the pressure vessel and 1969 ANSI B31.1 Code for the reactor coolant system pip-ing.
The ASME Boiler and Pressure Vessel Code permits pressure transients up to 10 percent over design pressure (110% x 1,250 = 1,375 psig), and the t
Amendment No. 15, 25, 35 29
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JAFNPP 3.1 LIMITING CONDITIONS FOR OPERATION 4.1 SURVEILLANCE REQUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM Applicability:
Applicability:
Applies to the instrumentation. and associated Applies to the surveillance of the instru-devices which initiate the reactor scram.
mentation and associated devices which initiate reactor scram.
Objective:
Ob_lective:
To assure the operability of the Reactor Protection System.
To specify the type of frequency of a
surveillance to be applied to the protection Specification:
instrumentation.
A.
The setpoints, minimum number of trip Specification:
systems, minimum number of instrument channels that must be operable for each A.
Instrumentation systems shall be position of the reactor mode switch shall be functionally tested and calibrated as as shown on Table 3.1-1.
The design system indicated in Tables 4.1-1 and 4.1-2 response time from the opening of the sensor respectively.
contact to and including the opening of the trip actuator contacts shall not exceed 100 mscC.
B.
Minimum Critical Power Ratio (MCPR)
B.
Maximum Fraction of LLuiting Power Density (MFLPD)
During reactor power operation at rated power and flow, the MCPR operating The MFLPD shall be determined daily during ihmits shall not be less than those shown below:
reactor power operation at 2r25% rated thermal power and the APRM high flux scram FUEL MCPR OPERATING LIMIT FOR INCREMENTAL and Rod Block trip settings adjusted if TYPE CYCLE 3 CORE AVERAGE EXPOSURE necessary as required by Specifications 2.1.A.l.c and 2.1.A.l.d, respectively.
BOC3 to 2CWd/t EOC3-2GWd/t EOC3-1GWd/t before EOC3 to EOC3-1GWd/t to EOC3 7x7 1.21 1.25 1.30 8x8 1.22 1.33 1.37 8x8R 1.20 1.33 1.37 Amendment No. 1[e, 1[, [, f, M, 30
G N
JAFNPP 3.1 BASES (ctnt'd)
Turbine control valves fast closure initiates a scram based on pressure switches sensing electro-hydraulic control (EHC) system oil pressure. The
-s switches are located between fast closure solenoids and the disc dump valves, and are set relative (500 < P < 850 psig) to the normal ERC oil pressure of 1,600 psig so that, based on the small system volume,
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s they can rapidly detect valve closure or loss of hydraulic pressure.
The requirement that the IRM's be T
inserted in the core when the APRM's read 2.5 indicated on the scale in the startup and refuel modes assures that there is proper overlap in the neutron monitoring system functions and thas, s
that adequate coverage is provided for all ranges of teactor operation.
B. The limiting transient which determines the required steady state MCPR limit de-pends on cycle exposure. The operating limit MCPR values as determined from the transient analysis for Cycle 3 (NEDO-24129 m,
and NEDO-24129-1, Supplement 1) for various s
core exposures are given in Specification 3.1.B.
T The ECCS performance analysis assumed reactor operation will be limited to MCPR s
of 1.18.
However, the Technical Speci-fications ILmit operation of the reactor to the more conservative MCPR based on consideration of the limiting transient as given in Specification 3.1.B.
yk,yd,}[,7,7$
b Amendment No.
m JAFNPP 3.3 and 4.3 BASES (cont'd) resulting from a turbine stop valve closure later, control rod motion is estimated with failure of the turbine bypass system.
to actually begin. However, 200 msec Analysis of this transient shows that the is conservatively assumed for this time negative reactivity rates resulting f rom the interval in the transient analysis and scram (NEDD-24129-1 rigures 1 and 2) with this is also included in the allowable
]:
the average response of all the drives as scram insertion times of Specification j
given in the above Specification, provide the 3.3.C.
The time to de-energize the required protection, and MCPR remains greater pilot valve scram solenoid is measured
'1 than the Safety Limit, during the calibration tests required by Specification 4.1.
The numerical values assigned to the specified
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scram performance are based on the analysis of The scram times generated at each data from other BWR's with control rod drives refueling outage and during operation the same as those on JAFNPP.
when compared to scram times generated during pre-operational tests The occurrence of scram times within the limit',
demonstrate that the control rod drive but significantly longer than the average, _
scram function has not deteriorated.
should be viewed as an indication of a system-In addition, cach instant when control atic problem with control rod drives especially rods are scrap timed during operation if the number of drives exhibiting such scram or reactor tripa.. individual evaluations times exceeds eight, the allowable number of
, shall be performed to insure that control j
inoperable rods.
rod scram times have not deteriorated.
In the analytical treatment of the transients, D.-
Reactivity Anomalies 290 meec are allowed between a neutron sensor reaching the scram point and the start of motion During each fuel cycle, excess operative of the control rods. 7 his is adequate and con-reactivity varies as fuel depletes and as servative when comparea to the typical time delay any burnable poison in supplementary control of about 210 maec estcasted from the scram test is burned. The magnitude of this exces, results. Approximattely 90 msee of each of these reactivity may be inferred from the critical intervals result from the sensor and the circuit rod configuration. As fuel burnup progresses delay, at this point, the pilot scram valve anomalous behavior in the excess reactivity solenoid de-energizer. Approximately 120 msee may be detected by comparison of 103 I
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yk,If,gk,@f,Jb Amendment No.
.