ML20235G619
| ML20235G619 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 04/30/1983 |
| From: | Brandon R, Gridley R GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20235G546 | List: |
| References | |
| NEDO-30120, NUDOCS 8709300125 | |
| Download: ML20235G619 (18) | |
Text
i Attachment 6' NED0-30120 DRF L12-00604 l 83NED037 Class I April 1983 ANALYSIS OF OPERATION WITH ONE SAFETY / RELIEF VALVE OUT-OF-SERVICE 1
FOR
\
FITZPATRICK NUCLEAR POWER PLANT f
ff' ,
Approved:
R. .
S A-randon, Manager Approved:
R. . Gridl 1
, Manater
[J Nuclear Services Engineering Fuel and Se vices Licensing 4
NUCLEAR POWER SYSTEMS DIVISION + GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNIA 95125 GENER AL h ELECTRIC B709300125 870925 l PDR ADOCK 05000333 G PDR
~. _ _ _ - _ _ - _ _ _ _ _ _
a ..
-Attachment 6 NED0-30120 IMPORTANT NOTICE REGARDING.
CONTENTS OF THIS REPORT i
PLEASE READ CAREFULLY' This report was prepared by General Electric Company (GE) solely for The Power
- Authority of the State of New York (PASNY) for PASNY's use with the U.S.
Nuclear-Regulatory Commission (USNRC)'for amending ~PASNY's operating license of the James A. FitzPatrick Nuclear Power Plant.. The information contained in this report is believed by GE to be an accurate and true representation of the facts known,.obtained'or provided to GE at the time this report was prepared.
The only undertakings of GE respecting information in this document 'are con-
.I tained- in the contract between PASNY and GE as per GE Proposal No. 424-TY700--
! EPI, dated August 27, 1982. The use of this information except as defined by I 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 GE 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.
9 O
1 A _ . . _ _ _ _ _ . . . _ _ _ _ _ _ _
NEDO-30120 Attaciunent 6 CONTENTS P_agg ABSTRACT y
- 1. INTRODUCTION 1
- 2. LOSS-OF-COOLANT ACCIDENT 1 2.1 General Discussion 1 2.2 Plant Specific Analysis 2 2.3 Conclusions 2
( 3. TRANSIENTS 2 l 3.1 General Di'scussion 2 l 3.2 Limiting oCPR Event, Load Rejection Without Bypass 3 3.3 MSIV Flux Scram 4 3.4 Conclusion 4
- 4. OVERALL CONCLUSION 5
- 5. REFERENCES 5 4
111
Attachment 6 NEDO-30120 TABLES Table Title Page 1 FitzPatrick Loss-of-Coolant Accident Analysis One ADS Valve Out-of-Service 6 2 S/RV Valve Setpoint and Grouping 7 3 Results of Load Rejection Without Bypass Analysis 8 4 Results of MSIV Flux Scram Transient Analysis 9 ILLUSTRATIONS Figure Title Page 1 Water Level Inside the Shroud and Reactor Vessel Pressure Following a Small Recirculation Line Break, HPCI Failure, One ADS Valve Out-of-Service, Break Area = 0.07 ft2 10 2 Peak Cladding Temperature Following a Small Recirculation Line Break HPCI Failure, One ADS Valve Out-of-Service, Break Area = 0.07 ft2 11 3 Plant Response to Generator Load Rejection Without Bypass 12 4 Plant Response to MSIV Closure (Flux Scram) 13
.f .
l l
iv
NEDO-30120 Attachment 6 ABSTRACT The purpose of this report is to provide the technical basis for opera-tion of the FitzPatrick Nuclear Power Plant with one of the safety / relief valves (S/RVs) out-3f-service.
With one S/RV out-of-service, there will be no impact on the transient parameters that influence critical power ratio (ACPR) and no change in the calculated 6CPR. The peak vessel pressure for the main steam isolation valve (MSIV) flux scram event is well below the ASME code upset limit of 1375 psig. This analysis has demonstrated that these conclusions are valid for current GE fuel types utilized in the FitzPatrick plant.
1 l
1 e
l v
1 _ _ _
NEDO-30120 Attachment 6.
- 1. INTRODUCTION The purpose of this report is to provide the technical bases for operation of the FitzPatrick Nuclear Power Plant with one of the 11 safety / relief valves
- ) (S/RVs) out-of-service.
The potential effect of one S/RV out-of-service is to change the pressure response of the reactor during transients and postulated accidents. This could conceivably impact the margins or safety limits for plant operation.
Accident and transient considerations for operation with one S/RV out-of-service are presented. The overall conclusion is given in Section 4.
- 2. LOSS-OF-COOLANT ACCIDENT 2.1 GENERAL DISCUSSION FitzPatrick has 11 S/RVs. All of these valves have a pressure-actuated safety function and seven of them have an additional pneumatically operated relief function automatically actuated by the Automatic Depressurization System (ADS). For loss-of-coolant accident (LOCA) evaluations, credit is taken for the safety function; however, not all of the S/RVs are actuated during a LOCA.,
If the out-of-service valve is one of the S/RVs with the ADS function, there can be a potential impact on the calculated peak cladding temperatures 2
(PCT) for small break sizes of less than approximately 0.2 ft . This may occur because, with a worst case postulated single failure of the High Pressure
. ~ .
~
Coolant Injection System (HPCI), the small break response is affected by the time required to depressurize the reactor to the operating pressure of the low pressure Emergency Core Cooling System (ECCS).
For larger postulated break sizes, the blowdown itself depressurizes the reactor vessel rapidly before the ADS actuates, and the number of S/RVs or actuation of the ADS has an inconsequential effect on the calculated PCT.
1
L Attachment 6 NED0-30120 l
l l For the FitzPatrick plant, the limiting LOCA is a large break (greater 2
than 1.0 ft in size), and the loss of the S/RV or ADS function has no effect on the calculated maximum average planar linear heat generation rate (MAPLHGR) limit.
2.2 PLANT SPECIFIC ANALYSIS A plant specific LOCA analysis has already been performed (Reference 1),
using the. approved Appendix K methods to determine the increase in small break PCT with one S/RV with an ADS function assumed to be out-of-service. The results of this anal'ysis are repeated in this report.
The results of the analysis are given in Table 1. The most limiting small 2
break in terms of PCT is the 0.07 ft recirculation suction line break. With one ADS out-of-service, the PCT is less than 1300*F, which is over 900*F below the 2200*F limit. This analysis was performed assuming the most limiting fuel type and exposure. The water level, pressure, and PCT for the worst case small break are shown in Figures 1 and 2.
2.3 CONCLUSION
S With one S/RV out-of-service, there is no impact on the calculated MAPLHGR limits for FitzPatrick, even if the out-of-service S/RV has the ADS function.
This conclusion is valid for all current GE fuel types utilized in the FitzPatrick plant.
1 l
- 3. TRANSIENTS )
3.1 GENERAL DISCUSSION
]
Operation with one S/RV out of service can affect the system response in the event of an abnormal operating transient. The decreased relief capacity can lead to higher transient pressures, which could affect the change in critical power ratio (ACPR) and the ASME code overpressure limits.
l 2
1 c -' - _____ ;
NED0-30120 Attachment 6 The following limiting abnormal operational transients were evaluated with one S/RV out-of-service using the NRC approved ODYN computer code for transient analysis:
- a. Load rejection without bypass at 104% power /100% flow, to evaluate the ACPR.
- b. Main steam isolation valve (MSIV) closure, flux scram, to evaluate the ASME code overpressure protection event.
The S/RV petpoihts and groupings for the analysis of these events are presented in Table 2. Note that the S/RV in the lowest setpoint group is conservatively assumed to be out-of-service in this analysis.
To justify that the conclusions of this analysis are cycle independent, transient analyses were performed using Cycle 5, Cycle 6 and an extrapolated end-of-equilibrium cycle (E0EC) nuclear data.
3.2 LIMITING ACPR EVENT, LOAD REJECTION WITHOUT BYPASS The transient analysis results for normal operation are reported in References 2 and 3 for Cycle 5 and Cycle 6 operation. The load rejection without bypass (LRNB) is the limiting ACPR event. A base case for the E0EC nuclear data was evaluated to assess the sensitivity of the analysis to cycle-dependent nuclear parameters.
The LRNB transient was performed with one S/RV out-of-service. The peak neutron flux, the peak heat flux, and the minimum critical power ratio (MCPR)
- remain unchanged for this event. This is because both the peak neutron heat and the peak heat flux occur before the S/RVs are actuated during this event. ;
Therefore, the ef fect of one S/RV out-of-service has no impact on the ACPR of the limiting transient LRNB event. While the time of peak neutron heat flux can be affected by the fuel nuclear parameters, there is sufficient margin between the peak time and actuation of the SRVs to assure that there is no effect on ACPR.
3
Attiachment 6: NED0-30120 The results of the transient' analyses are summarized in Table 3 and the
-time response of this transient is shown in Figure 3-for a typical case.
There was no change in the ACPR for the cases considered..
'The non-pressurization events (e.g., rod. withdrawal error) are independent of valve setpoint and valve capacity, therefore, the ACPR values are unchanged as a result'of one S/RV out-of-service.
'The adequacy of'the S/RV capacity based on ASME code requirements is demonstrated by the MSIV closure transient with high flux scram. The peak vessel pressure for this event increases a maximum of 15' psi as a result of one S/RV out-of-service, resulting in a peak pressure of 1290 psig for Cycle 5.
The Cycle 6 and E0EC cases show an increase in peak vessel pressure of 12 psi, with a maximum pressure of 1268 psi. Therefore, there is a large margin to the ASME code limit of 1375 psig.
The output parameter data for the MSIV flux scram transient considered' in this analysis is summarized in Table 4. The time response of key variables for this transient is shown in Figure 4 for a typical case.
3.4 CONCLUSION
With one S/RV out of service, there will be no impact on the transient parameters that influence ACPR and no change in the calculated 6CPR. The peak vessel pressure for the MSIV flux scram event is well below the ASME code upset limit of 1375 psig. This analysis has demonstrated that these conclu-
.' sions are valid for current GE fuel types utilized in the FitzPatrick plant.
. i l
1 L.
. 2-- - - _ -. _ _ _ _ _ _ _
Attac'hment 6' NEDO-30120
-1 o
- 4. OVERALL CONCLUSION
)
The operation of the FitzPatrick Nuclear Plant at full power with'one' 'f S/RV'out-of-service ~will have no impact on operating' limits. This conclusion' ') 1
.is valid for current General Electric fuel types, operating strategies and
]
{ analysis methods, as applied to the FitzPatrick Nuclear Plant. !
l- :
.5. REFERENCES
- 1. FitzPatrick. Nuclear Power Plant, Cycle 5, Analysis for Operating with One. Safety / Relief Valve Out of Service'. NEDO-22226, September 1982. !
- 2. Supplemental Reload Licensing Submittal for FitzPatrick Nuclear. Power-1
- Plant Reload 4, Y1003J01A25, August 1981. ;
I
- 3. Supplemental Reload Licensing Submittal'for FitzPatrick Nuclear Power Plant Reload 5, Y1003J01A56, March 1983.
O i
l
[
5
Attachment 6 NEDO-30120 Table 1 FITZPATRICK LOSS-OF-COOLANT ACCIDENT ANALYSIS ONE ADS VALVE OUT-OF-SERVICE Recirculation Line Break Upstream of Discharge Valve.
System Failed: HPCI Systems Remaining: 2 LPCS + 2 LPCI + 6 ADS Valvesa Peak Cladding Break Size Uncovery Time Reflooding Time Temperature (ft2) (sec) (sec) (*F) 0.05 , 324.6 412.4 1103 0.07 265.6 372.4 1271 0.10 232.1 330.5 1241 "Two of the LPCI systems inject into the broken loop and it is conservatively assumed that all of the injected water is lost through the break.
e l
l l
I 6
.~
I e .
Attachment 6 NED0-30120 l
l Table 2 l S/R VALVE SETPOINT AND GROUPING New (One S/RV l Previous S/RV Grouping Out-of-Service) Grouping ]
Setpoint S/RV Setpoint S/RV No. No.
l (psig) Target Rock (psig) Target Rock 1090 + 1% 2 1090 + 1% 1 1105 + 1% 2 1105 + 1% 2 1140 + 1% ,
7 1140 + 1% 7 I
l l
7
Attachment 6 NEDO-30120 i
Table 3 RESULTS OF LOAD REJECTION WITHOUT BYPASS ANALYSIS Peak Neutron Peak Heat Time to Time for First Power / Flow Flux Flux Peak Heat S/RV to Open Case (%) (%) (%) Flux (sec) (sec)
Cycle 5 104/100 653.1 125.17 1.0337 1.3302 Cycle 6 104/100 623.1 127.80 1.0103 1.3126 E0EC 104/100 - 596.1 127.6 1.0249 1.3116 O
8
~.
-o .
Attachment.6_ NEDO-30120 Table 4 RESULTS OF MSIV FLUX SCRAM TRANSIENT ANALYSIS Peak Vessel Power / Flow Pressure Change in Pressure Case (%) (psig) bbrgin (psi)
(Relative to Base Case)
Cycle 5 104/100 127$
Base Case Cycle 5 104/100 1290 - 15 1 S/RV Out !
of Service i
Cycle 6 104/100 1256 Base Case Cycle 6 104/100 1268 - 12 1 S/RV Out i of Service i
E0EC 104/100 1256 f Base Case-E0EC 104/100 1268 - 12 1 S/RV Out of Service i
9
i !
5
>tt3OW9 rrD }tr O T
Z$ogoCo I_Jm3ag$<E 0 0 4 2 O0 0
0 n
- 1 o i
t a
- l u
c
- i r
c e
- R2 t
lf l
a7 m0 0 S I 0 0 8
a
=
g na F
A
-AF i e wr T B oA
- l l k oa
- F e eB r
r
- u se sc ei D
U O
0 0
6 rv P r e
R H
S
- lS e -
sf E so
)
D c e e-I S h Vt N E u
- rO I
L M I o -
E T t e V
E cv al L ea RV E
R "w - 0 dS nD U I 0
aA S 4 S
E
/ d e R un P oO L- r E h ,
S S e S r E
V eu hl t i
- a d
eF
- iI sC nP 0 I H
- I 0
2 l ek va
- ee L r B
- r ee t n ai WL
- - - 1
- - e 0
- - - -- O r
u g
0 O 0
2 0
8 M
4 i _
1 F _.
3$ wE3 ea ,N s _
o
- l 0
0 8
- )
F 2t f
r h
/
u
- 8
(
t T
- N E
I C
I l 0
F 0 F
6 E
O C
~ T I
- R E
5 2
=
F h S
M-
)
I L N c e
T A (
s R E T
C-P T M I
A T
- E H
0 l 0 4
f 0
=
h 0
- , I
, 2 0
- 0 0
0
_ 0
_ 1
_ =
- h
- - - ( 0
_ 0 0 0 O
_ 0 0 0 0 0 0
_ 3 2 1
_'oW ?4ba OEokJu V4*
_ C
- i
I S
P
( m T F(
WuL 1
I 4
3 T ITY EI t I
iL YVTT T[II R tt t SYvVf C: t L TI E VvVi Et i
4 8 24 ITVV VCII IATT TECC KFT F
ERRl EU ;
t[ /
2
% 6 1
L r 't$ RE TrS LM TLiHl Sf EOt tl 3
/PAE
!PR O0 VSfU i V0 6 I23 u 2 12t '
_ 31 1
> 4 C 2 E S
I L
E
\
M 4 3
- 4. T 2
I D % 8.
0I u
2 A 2 2 4
/ 1 0
4 3
2 1 . 1
. _ '0, 0*
. .0 - * - .
0- 0 O 3 0 1 2-0 0 0 -
3 2 1 i
eh -
0bu " lgx T
R I
X K U S L
F
" - P E W T SWO AW - OL E0 FLFW dti Xtf UE LCi tFt r ON t t Q 1 4 8
1 EFMO lMAL fREF HET NS5U C1 I ET 4
1 3
8 4
0LI (LNA RS LEIH T E ESBD VSRE N
Util Ed0 EEUE 1CC 1 LVTF 6 3 6 123 12345
- 3) 3 C
' E S 1
(
E M
- 4. TI 4 2 2
- I\ 3 b '
4 _
jI q 2 1
lgIu 2 1
l
'I l
- 3 l
_ ~ _
0 3, - . .0 -
0 0 O 0
9 0 0 O 0 0 0 0 5 1 1 2 1 1 22 a e_
e t
l i
S P . T ,
(WHW 3 T EOCO IYY SLLL YVTT IFFF III R VV EEE V I SVVV ELLL u 4 I
EfTET ERR 4 LTF5 RE -
ETE5 LML SEIA DPAH SFLP 3 IPRT EAE1 OOCU VSRB VDSI 123456 8 123 8 41 q
1 u 1C 2 E 1 S
(
E M
I 3
- 2. T 2.
3 1I j 2 1
6 6 0
d1 .
4
- u. 3 r.
i 2
1 0
0
. .0 - * . .-
0 0
0 O 3 O. 1 2 0 0 - -
3 2 1 5 $ g6,6U g 3"; . gy T
R I
X K U S L - -
F - P ,
UE HL F
X~ 4 EFMO RMRL
- AEF
- 4 j
LCI HET n FFE FL 6 C15 NS U 6 NRN I EI 2 OUI ( LNR RS LEIW T E ESBU UER EVO NAC 12345 X 1
-8 41 C
E VSRL EEUL LVTP 12345 u
w 8 4
~
S a
I 2-E M
1 sf
- 2. T 3
I 2(
Wm -
2 3
~_
)
.f' y
)
6 1
1 2\.
W -1 6
~
_ 8 - '
. I
.0 - .
_:0 .-
o*
e 0
0 0 0*
0 0
0 n 0 0
g 1 2 1 1 8 ab -
u l
1 4 4 ..- j
-Attachment'7 )
EVALUATION OF.SRV OPERATION DURING THE -l '
JULY 19, 1985 SCRAM i
The plant transient commenced at 14:33:17 on July 19, 1985 following a ~ problem associated with the transfer of electrical i buses. A turbine trip and reactor scram occurred at 14:33:58.
MSIV's closedeat 14:34:02. MSIV closure isolated the reactor from bypass valve pressure regulation. Since the reactor had been operating. at .100% power, decay heat removal and pressure control is. accomplished by the SRVs. There are three instrument ,
systems that monitor SRV operation. The systems rely on j secondary effects like noise 'or temperature. Temperature j response is slow and provides no- additional information on 1 subsequent SRV operations. The sonic monitor provides fast response and continuous monitoring but may provide erroneous. data from crosstalk. The quality of instrumentation made the analysis very difficult. The reactor level / pressure chart was used to.
analyze the transient. This is a dual speed recorder with chart
. speeds of one inch per hour and one inch per minute. The chart j s3eed is changed manually. The shift to fast speed was done a'aout eight minutes after the scram.
Using the. level / pressure . chart the eight-minute transient was plotted in the (1).one inch per minute timeframe using post-trip data (chart available on site). Time reference marks were established on the chart, once time reference was made, a real time event could be established.
1 At 14:34:47, or 50 seconds after the scram, SRVs K, L, J,H, and A opened for 29, 11, 9,- 9, and 2 seconds respectively. The '
-post-trip data shows torus level changing at 14:34:48 which j supports SRV operation.
The chart shows all the valve operations for the remainder of the transient. It is interesting to note the coincidence of SRV valve operation with each level and pressure upset. At no time during the transient did the pressure exceed 1120 psig. Accord-ing to sonic indication, the following valves operated: K,L,J, H,A,E,G, and F. According to SRV tailpipe temperature alarms, the following valves operated: A,K,L,H,J, F, and G.
The attached table provides the recent setpoint data for the SRVs.
Based on this table, if one wanted to use the reactor pressure dome indication and known Wyle Laboratory tested SRV data as the determinant factor when the SRVs should have lifted, then the only relief valves that should have lifted would have been C, K, '
and L (assuming its setpoint is still in the 1095-1098 region).
l
Conclusion:
Attempting to use reactor dome pressure as a measure of SRV setpoints is too prone for error.
The event described in Attachment 8 also supports thia.
o s s p 7 7 7 7 7 7 t 8 8 8 8 8 n n 8 e 7 / / / / 7 / i i / 7 St 8 3 4 5 3 8 5 4 8 n ' 0 0 0 0 ' 0 t t 0 '
"p ve / / / / / u u /
d 2 2 2 2 d 2 p p 2 d oE e 0 0 0 0 e 0 ( ( 0 e P t t t
" g s n n n n s n 7 7 n s n e o o o o e o 8 8 o e ti t t ' '
t sw ro t
4 4
0 2
5 4
3 4 t 1
3 d d 3
8 t il o 1 1 1 2 o 1 e e 0 o Fl N 1 1 1 1 N 1 t t 1 N o s s dF e e e t t te ) )
st t5 t5 ea o8 o8 TD N' N' d
n a
t tn ne 5 3 5 5 5 5 3 3 3 3 5 iv 8 8 8 8 8 8 8 8 8 8 8 oE / / / / / / / / / / /
p 4 6 6 5 0 9 6 8 0 5 8 to 0 1 2 2 1 2 1 2 1 1 0 et / / / / / / / / / / /
S 4 7 3 3 4 3 7 1 5 7 5 r 0 0 0 0 0 0 0 0 0 0 0 lo ai n n n n n n n n n n n ur o o o o o o o o o o o tP c 8 8 0 2 3 0 5 5 6 3 8 At 3 4 4 1 0 4 3 3 3 9 9 s 1 1 1 1 1 1 1 1 1 0 0 ne 1 1 1 1 1 1 1 1 1 1 1 wT - - - - - - - - - -
o 4 0 5 0 8 3 0 3 9 5 nf 3 4 3 0 3 3 3 3 8 9 Ko 1 1 1 1 1 1 1 1 0 0 1 1 1 1 1 1 1 1 1 1 te st ea BD lt an) ni g 0 0 0 5 5 0 0 0 0 0 0 i oi 4 4 4 0 0 4 4 4 4 9 9 mps 1 1 1 1 1 1 1 ] 1 0 0 otp 1 1 1 1 1 1 1 1 1 1 1 Ne(
S l
a 5 8 3 0 6 7 2 7 0 7 2 i 4 8 5 8 5 9 1 8 1 4 6 r 0 0 0 0 0 0 0 0 1 0 0 e 1 1 1 1 1 1 1 1 1 1 1 S
V R A B C D E F G H J K L S
' Attachment 8 !
SUMMARY
OF JiNUARY 17, 1983 SCRAM AND SRV OPERATION
Background
On January 17, 1983 while at 86.3% rated power, a reactor. scram
^
and isolation occurred due to an apparent low water. level due to testing. The peak reactor pressure per the post trip log was 1120.9 psig and 1170 psig based on the strip chart recorder. The. ;
following relief valves automatically lifted as indicated by the sonic detectors: L, J, G, and H. Relief valves'K and J were pulled for testing due to a suspected setpoint problem.
SRV Performance Nominal Date of Test Setpoint at Wyle Lift Pressure of !
SRV & Serial # (psig) Following Scram First " Pop" at Test K 1056 1090 01/24/83 1115 psig i L 1062 1090 06/29/83 1124 psig D 1080 1105 06/30/83 1145 psig E 1050 1105 04/12/85 1122 psig J 1087 1140 01/26/83 1129 psig '
G 1012 1140 06/29/83 1110 psig H 1051 1140 03/21/85 1211 psig A 1045 1140 03/20/85 1209 psig B 1088 1140 06/29/83 >1197 psig 1
C 1052 1140 04/11/85 1207 psig 1 i
E 1013 1140 06/83 1174 psig Observation: If one used reactor dome pressure to determine i relief valve setpoints, the following valves should have lifted: 4 i
a) Based on 1170 pressure peak - K, L, D, E, J, G l b) Based on 1120 pressure peak - K, G l l
These actual test results when compared to which !
valves actually did lift, support the discussion ;
of Attachment 3. )
i l
- - _ _ _ - _ _ _ _ _ _ _ l