ML20008E769
| ML20008E769 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 02/28/1981 |
| From: | Lyon R EG&G IDAHO, INC., EG&G, INC. |
| To: | Panciera V Office of Nuclear Reactor Regulation |
| References | |
| CON-FIN-A-6267 EGG-EA-5358, NUDOCS 8103090593 | |
| Download: ML20008E769 (15) | |
Text
i l
EGG-EA-5358 February 1981 EVALUATI0fl 0F OPERATION OF ZION STATION UNITS 1 AND 2 WITH ONE LOOP ISOLATED
/4 NRC Researci anc Tecinica #T d % o.fr[er s
R. E. tye, Assistance.Penn.rt i
\\
U.S. Department of Energy Idaho Operations Office
- Idaho National Engineering Laboratory
\\s
~
y..",hh,.(
g M m,r wa m,,,,;,,*,, Ch j
jr
,g 7 ;g 9
- Fr if
.h s. d y y M D b
~f
.g
- mm 7 r===,ww w, m.mme
~~c4!._"lMW1 MM w,
- - ---w2/a a..
L
&N 2^J 1. 55 5 $4.T M ?'"
r', W'N*""
cg$ 1?'
f
.s.
j,,,c; g
,--~~~e
- -~.-
,...y,
,,h s
~
T'
~
Q 4
D.
ilE
.id.:fS/24D.s bu ENd$dEY w
w m~
This is an informal report intended for use as a preliminary or working document Prepared for the U.S. Nuclear Regulatory Connission Under DOE Contract No. DE-AC07-76ID01570 Q
8103090593
i D
I p EGnG.... -
70ku IG4G ym en., ii v3 INTERIM REPORT i
Accession No.
l l
Report No.
EGG-EA-5358 I
Contract Program or Project
Title:
l (N-1) Loop Operation of Beaver Valley and Zion 1/2 Subject of this Document:
Evaluation of Operation of Zion Statior. 'Jnits 1 and 2 with One Loop Isolated
\\
Type of Document:
j Technical Evaluation Report i
Author (s):
I x,,'. rch a d Technical
" e Lion 0#0T[
~
Date of Document:
February 1981 v
Responsible NRC Individual and NRC Office or Division:
V. W. Panciera, NRC-DSI This dccument was prepared primarily for preliminary or internal use. It has not received full review and approval. Since there may be substantive changes, this document should not be considered final.
i EGaG Idaho, Inc.
Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
Under DOE Contract No. DE-AC07 761D01570 NRC FIN No.
A6267 INTERIM REPORT r
CONTENTS I N T R O D U CT I O N...........................................................
1 TECHNICAL DISC llSSION...................................................
2 Core and Coolant Boundary Protection Analysis.....................
2 Standby Safeguards Analysis.......................................
7 Loss-of-Coolant Accident..........................................
7
]
Technical Specification Char.ges...................................
10 i
S UMM AR Y AN D R E COMME N D AT I O N S............................................
11 REFERENCES.............................................................
12 TABLES 1.
Core and Coolant Boundary Protection Analysis.....................
3 2.
Standby Safeguards Analysis.......................................
8 3.
Loss-of-Coolant Accident..........................................
9 1
D l
i i
INTRODUCT ION The Zion Station Units 1 and 2, operated by Commonwealth Edison Co.,
are four-loop PWRs designed by Westinghouse. The reactor coolant systems are equipped with loop isolation valves which can isolate a steam generator and reactor coolant pump from the primary system, permitting continued Zion Units 1 and 2 operation at reduced power with one loop shut down.
have been operating with a condition in their operating licenses which Commonwealth Edison Co.
prevents them from operating with a loop isolated.
which has submitted a request for amendment to their operating license provides analyses to establish the response of the plant to transients and accidents when operating with one loop isolated. This report documents tne review which has been performed on those analyses and the acceptability of permitting Zion Station Units 1 and 2 to operate with one loop isolated, 1
i n
i TECHNICAL DISCUSSION Commonwealth Edison Co. has submitted Loss-of-Coolant Analyses and Technical Specification changes to support operation with one loop isola-ted.1 The Nuclear Regulatory Commission subsequently requested additional analyses of system response to transient conditions for isolated loop oper-ation.2 The reply to this request" was not responsive to the question 1
r and, thus, insufficient information exists to directly determine the accept-ability of operation with an isolated loop. Westinghouse has expressed the l
opinion that operation of a four-loop plant with one loop isolated is very similar to a three-loop plant.
Information on the response of Beaver Valley j
r Power Station Unit 1, a three-loop plant, to transient conditions with an isolated loop is available.4 A comparison of this information with results of the Prairie Island Nuclear Generating Plant (two-loop) analyses has been made to determine if this may be a valid assumption, and if the f
results can be applied to Zion Station. The discussion of the analyses is
. l broken into three categories, as was done for the Zion Final Safety Analysis Report (FSAR)5 and, in general, is carried to the same level of detail as was presented in the FSAR.
I Core and Coolant Boundary Protection Analysis Analyses in this area cover minor transients which should result in, i
at most, a reactor trip and which should not result in any loss of function of fuel cladding and reactor coolant system boundaries. The analyses which i
l have been reviewed are listed in Table 1.
The basic acceptance criteria for the majority of these transients, as given in the Standard Review Plan (SRP),0 are that the reactor system pressure remains below 110% of design pressure and that the DNB Ratio remains above 1.30; therefore, these param-eters are listed in Table 1 along with other parameters which may be of interest.
In most cases, these parameters are listed in terms of change in magnitude rather than absolute magnitude, to facilitate comparison between plants.
r A comparison of results of the Prairie Island transients with the corresponding results of Beaver Valley operating with one loop isolated 2
TABLE 1.
CORE AND COOLANT BOUNDARY PROTECTION ANALYSIS Beaver Valley Prairie Island Zion Three-loop Two-Loop Two-loop Four-loop Three-loop Transient Press DNBR Press DNBR Press DNBR Press DNBR Press DNBR b
b b
a D
Uncontrolled 0.75a 270 F 0.25a 180 F 0.75a 50 F 0.84 450 F Not analyzed RCCA bank withdrawal-subcritical Uncontrolled 100 0.35 50 0.44 160 0.45 130 0.37 Not analyzed RCCA bank withdrawal-at power RCCA 25
<0.4 Not analyzed N.G.
<0.4 N.G.
<0.4 Not analyzed misalignment Uncontrolled 57/129/15c Not analyzed 54/90/9c 91/140/16c Not analyzed boron dilution Partial loss N.G.
0.1 N.G.
0.32 N.G.
0.25 N.G.
0.07 N.G.
0.15 of forced reactor coolant flow Startup of 20
<0.4
>6d 100
<0.4 140 0.1 9d inactive reactor coolant loop Loss of load /
280 0.1 260 0.2 310 0.1 310 0.05 Not analyzed turbine trip
1 TABLE 1.
(continued)
Beaver Valley Prairie Island Zion Three-loop Two-Loop Two-loop Four-loop Three-loop Transient Press DNBR Press DNBR Press DNBR Press DNBR Press DNBR Loss of 220 1.30 110 1.le N.G.
1.50 N.G.
1.9e Not analyzed normal feedwater Excessive 50 0.05 I.V.
0.2 13 0.32 I.V.
0.07 Not analyzed heat removal, feedwater system malfunction Excessive 25 0.22 30 0.1 12 0.15 I.V.
0.2 Not analyzed load increase Loss of N.G.
N.G.
N.G.
N.G.
N.G.
N.G.
N.G.
N.G.
Not analyzed offsite power Accidental I.V.
<0.4 1.V.
<0.4 Not analyzed Not analyzed Not analyzed depressuriza-tion of main steam system Accidental I.V.
0.26 Not analyzed Hot analyzed Not analyzed Not analyzed depressuriza-tion of reactor coolant system r
TABLE 1.
(continued) a.
Thermal flux (fraction of nominal).
b.
Change in average clad temperature.
c.
Time to loss of shutdown margin (refueling /startup/at power).
d.
Time to loss of shutdown margin.
e.
Pressurizer volume (fraction of nominal).
m 1.
Pressure values are increase in pressure (PSI).
I.V. indicates that the pressure remains below the initial value.
2.
N.G. indicates that the information was not given in the analysis.
3.
DNBR values are decrease in ratio.
showed a moderately good correlation. With one exception, the pressure transients were less severe for Beaver Valley isolated loop than for Prairie Island. The changa in DNBR was not consistent but, due to the higher starting DNBR for Beaver Valley isolated loop, the absolute value was always higher for Beaver Valley. There was no clear trend as to the speed of the transient. For some of the transients, Beaver Valley responded faster, while for others, Prairie Island was faster.
A comparison of results of the Zion transients with the corresponding results of Beaver Valley operating with all loops also gave a good correla-tion.
Pressure transients were, in all cases, less severe for Beaver Valley than for Zion, while the DNBR was almost the same.
For all four cases analyzed, the maximum pressure was a result of the loss of load with turbine trip, and the minimum DNBR occurred as a result of the uncontrolled rod cluster control assembly at power.
The analyses for the Zion FSAR did include a small number of transients which were analyzed for an isolated loop. The partial loss of forced reac-tor coolant flow resulted in a slightly greater change in DNBR for isolated loop operation, but the actual value was lower for normal operation due to the different starting point. This is the same response which was seen for Beaver Valley. Startup of an inactive reactor coolant loop was also analy-zed, with a resulting time of 9 min to loss of shutdown margin if the intcr-locks were violated.
There is one area where operation with an isolated loop may not meet current acceptance criteria. The acceptance criteria for an uncontrolled boron dilution at power, as given in SRP 15.4.6, is that a minimum of 15 minutes must be available from the time an alarm makes the operator aware that the dilution is in progress until loss of shutdown margin occurs. For normal operation, the FSAR states that 16 minutes is the time for recriticality after reactor trip. For operation with an isolated loop, this time would be somewhat reduced because of the smaller volume being diluted.
l l
t 6
l
J -
Standby Safeguards Analysis Analyses in this area cover transients that are more severe but very infrequent and may lead to a breach of fission product barriers. Transients in this category that have been considered are listed in Table 2.
There are two transients in this area where isolated loop analyses were performed for the FSAR. For the reactor coolant pump locked rotor transient, the resulting reactor coolant system pressure was higher for isolated loop operation than for normal operation, but peak clad temperature was lower for isolated loop operation.
For the complete loss of forced reactor cool-ant flow, isolated loop and normal operat' ion transients resulted in similar plant response. Results of these analyses compare favorably with the results noted for Beaver Valley, and acceptance criteria have been met for these two transients.
The remaining transients in this classification did not yield suffici-ent data for any comparisons.
Loss-of-Coolant Accident The Westinghouse approach to analysis of isolated loop operation for large and intermediate loss-of-coolant accidents (LOCAs) has been reviewed and approved by NRC on a generic basis.
The approval was conditional on two items; first, that at least two nodes are used in modeling the inactive loop break and, second, that momentum flux is accounted for between the two cold leg nodes. The model described in the Commonwealth Edison Co. submit-tal uses only a single node in the inactive loop model and thus, does not comply with Condition 1.
Compliance with Condition 2 is dependent on com-pliance with Condition 1.
A summary of the results of the loss-of-coolant analysis for the Zion plant with one loop isolated are given in Table 3.
Both the peak clad temperature and metal water reaction are within the acceptance criteria of 10 CFR 50.46 for all breaks analyzed. The analysis considered both an active and an inactive loop break and confirmed that the active loop break is the limiting case. The inactive loop break results would be somewhat 7
TABLE 2.
STANDBY SAFEGUARDS ANALYSIS Beaver Valley Prairie Island Zion Three-loop Two-Loop Two-loop Four-loop Three-loop Transient __
Press DNBR Press DNBR Press DNBR Press DNBR Press DNBR Steam gener-Only Not analyzed Only Only Not analyzed ator tube radiological radiological radiological rupture consequences consequences consequences Steam pipe I.V.
<0.4 I.V.
<0.4 I.V.
<0.4 I.V.
<0.4 Not analyzed rupture Rod ejection N.G.
2565a N.G.
20388 H.G.
N.G.
200 2500a Hot analyzed accident Reactor 425 2031a 525
<20316 480 1680a 150 1760a 160 1280a 03 coolant pump locked rotor Complete loss N.G.
0.22 N.G.
0.18 N.G.
0.32 N.G.
0.3 N.G.
0.3
~
of forced reactor coolant flow a.
Peak clad temperature.
TABLE 3.
LOSS-OF-COOLANT ACCIDENT Break Size Peak Cald Metal / Water and Tempgrature Reactor Location
( F)
(%)
Active loop 1880 2.7 1.0 DECL Active loop 1996 4.1 0.6 DECL Active loop 2061 5.1 0.4 DECL Inactive loop 1748 1.5 0.6 DECL higher if two nodes were used in modeling the break; however, this has been 8
shown to only be about! 13 F.
Neither the Commonwealth Edison submittal nor the Westinghouse generic analysis has considered a small break LOCA. Westinghouse has taken the 9
position that the response of a four-loop plant operating with one loop isolated is very similar to a three-loop plant. They further state that, "since all N-loop small break analysis show a great deal of margin to the regulatory limits, there is adequate assurance that small break LOCA analy-ses for plants operating in the (N-1) loop condition with loop stop valves closed will not be limiting." This would appear to be an adequate justifi-cation; however, there is one are of concern. Analyses presented for Beaver 4
Valley show that isolated loop analyses result in higher peak clad temper-atures and metal water reaction than the corresponding values for normal operation.
In addition, it appears, from the limited data available, that this difference increases for smallcr break sizes.
If this trend continued into the smaller breaks, the statement by Westinghouse may not be adequate justification.
l 9
Technical Specification Changes Several changes in the reactor protection system are required for isolated loop operation. These are as follows:
1.
Change the P-8 interlock setpoint to <80% of rated thermal power 2.
Reduce the dT overtemperature trip setpoint 3.
Trip the following channels associated with the inactive loop:
a.
Overpower dT b.
Overtemperature dT c.
T
- U*~IU" avg d.
Steam line pressure e.
Differential pressure between steam lines.
These actions are to be completed with the reactor subcritical.
These requirements are listed in the Technical Specifications, Section 3.2.2.8.4.
i f
1 10
SUMMARY
AND RECOMMENDATIONS Commonwealth Edison Co. has not submitted sufficient information to entirely support their request for license amendment to allow operation of Zion Station Units 1 and 2 with one loop isolated. Review of additional information tends to support the adequacy of the system when operating in this mode; however, a final decision will be dependent on additional infor-mation in several areas, as follows:
1.
Since most of the information for non-LOCA transients is based on extrapolation of Beaver Valley data, open items on that licensing action, particularly the applicability of the LOFTRAN code, need to be resolved.
2.
The applicability of the 15-minute time to loss of shutdown margin for the boron dilution event needs to
'e established for Zion.
If it is determined to be a
applicable, Commonwealth Edison Co. should be requested to provide the appropriate analysis for isolated loop operation.
3.
It is recommended that Commonwealth Edison Co. again be requested to provide analyses of at least the limiting moderate frequency transients with the isolation of one loop to allow better confirmation of the applicability of results from the Beaver Valley analyses to Zion.
4.
It is felt that insufficient information exists to determine the adequacy of the plant response to a small break LOCA.
It is recommended that the appropriate analysis be requested from Commonwealth Edison Co. to verify the response of the plant to a small break LOCA.
With this additional information, it is felt that sufficient informa-tion will exist to justify allowing operation of the Zion Station Units'l and 2 with an isolated loop.
11
REFERENCES 1.
Letter from R. L. Bolger, Assistant Vice President, Commonwealth Edison Co., to B. C. Rusche, Director, Office of Nuclear Reactor Regulation, USNRC, dated January 27, 1977.
2.
Letter from A. Swencer, Chief, Operating Reactors Branch No. 1, USNRC, to C. Reed, Assistant Vice President, Commonwealth Edison Co., dated July 7, 1978.
3.
Letter from W. F. Naughton, Nuclear Licensing Administration, Commonwealth Edison Co., to Director of Nuclear Reactor Regulation, USNRC, dated September 27, 1978.
4.
Letter from D. N. Dunn, Vice President, Operations, Duquesne Light, to A. Schwencer, Chief, Branch No. 1, Division of Operating Reactors, dated October 27, 1978.
5.
Final Safety Analysis Report, Zion Station Units 1 and 2.
6.
U.S. Nuclear Regulatory Commission, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, NUREG-75/087, March 1979.
7.
Letter from J. F. Stolz, Chief, Light Water Reactors Branch No. 1, Division of Project Management, to T. M. Anderson, Manager, Nuclear Safety Department, Westinghouse Electric Corp., dated February 28, 1979.
8.
Letter from C. Eicheldinger, Manager, Nuclear Safety Department, Westinghouse Electric Corp., to J. F. Stolz, Chief, Light Water Reactors r>roject, USNRC, dated September 7,1977.
9.
Letter from C. Eicheldinger, Manager, Nuclear Safety Department, Westinghouse Electric Corp., to J. F. Stolz, Chief, Light Water Reactors Project, USNRC, dated June 27, 1977.
l l
I 12 w_
_