ML20247E742
| ML20247E742 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 09/08/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20247E714 | List: |
| References | |
| NUDOCS 8909150367 | |
| Download: ML20247E742 (8) | |
Text
-
bE - L.
'of.
UNITED STATES.
WZ ' 7
- NUCLEAR REGULATORY COMMISSION o
4W j$
i g
WASHINGTON, D. C 20555 y.
g j
1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED T0' AMENDMENT NO.118' TO FACILITY OPERATING LICENSE' NO. DPR-39
- .3 N
. AND AMENDMENT NO.107 TO FACILITY OPERATING LICENSE NO. DPR ! n C0pm0NWEALTH EDISON COMPANY' ZION NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-295 AND 304
1.0 INTRODUCTION
~
h
'By'1etter dated July 6, 1989, as supplemented August 4 August 10, and August 24, 1989, theCommonwealthEdisonCompany(thelicensee) requested amendment to Facility Operating License Nos. DPR-39 and DPR-48 for the Zion Nuclear. Power Station.. The proposed amendment would change the plant Technical.
. Operation (LCO)(in TS 3.15.2.C: for the "0"present Limiting Condition fordieselgener Specifications TSs) that would allow the from 7' days to 45 days during the fall of 1989, while Unit 1 is in refueling outage..The subject DG is shared between two units'and existing TS 3.15.1.B requires it to be operable whenever either unit is.in operation. Thus, the licensee.is not able to perform maintenance or modifications on the "0" DG which.take longer than the 7-day LCO-without shutting down both units.
The purpose of this amendment is to allow the licensee to perform the maintenance recommended by the manufacturer to be performed every 5 years and to modify DG "0" control and transfer capability when aligning to either unit's engineered safety feature (ESF) bus. The licensee has provided a list of.all the maintenance activities items to be performed and estimated manhours for activity in Reference'1. Since the service water and the component cooling (CC) pumps are cross connected between two units, the licensee has committed to make the DGs (IA and IB) of Unit 1 available during the 45-day LCO period. The staff has approved a similar request, in principle, in 1984 and 1987.
2.0 DISCUSSION The Zion "0" diesel generator is shared between Units 1 and 2.
Technical Specification 3.15.2.C permits the diesel to be inoperative for 7 days during power operation. The diesel manufacturer recommends an extensive maintenance overhaul every 5 years which takes more than 7 days. The licensee requested a TS change to allow the "0" diesel generator to be inoperative for this maintenance. The licensee plans to perform the diesel generator overhaul during the Unit I refueling outage scheduled from September 8 through November 20, 1989.
8909150367 890908 PDR ADOCK 05000295 P
PDC 4
r-
/,*
4 i
-2 l
Available information for the Zion station includes a probabilistic risk assessment (PRA) for power operation (Ref. 4) and a PRA for non-power operation (Ref. 5). The licensee based a portion of its justification for the requested TS change upon the power operation PRA. Other justification included analyses ofplantresponsetoselectedeventinitiations,adescriptionofsomeofthe equipment potentially affected by non-operability of the 0" diesel, and a description of selected actions the licensee plans to take during the period of non-operability. The licensee provided a list of ESF equipment powered from each ESF bus and identified the minimum number of components available under the worst system condition for Unit 2 while DG "0" is out of service. This list is shown in Table 1.
2.1 Unit 1 A comparison of the power operation PRA (Ref. 4) to the non-power operation PRA (Ref. 5) shows that the risk of a significant release during non-power operation may not be neglected. This concern is addressed in many sources, one of the most recent being Generic Letter (GL) 88-17 (Ref. 6). The licensee did not address Unit 1 in the original submittal dated July 6, 1989. After discussion with-the licensee, additional information was provided to correct this omission (Refs. 2 and 3).
The licensee stated that all fuel is to be removed from Unit 1 for most of the time the "0" diesel is unavailable (Ref. 2). Further, no cid-loop or reduced (Ref. 3)y operations, as defined in GL 88-17, are planned for this outage inventor This removes most of the risk associated with non-power operation.
The licensee also addressed numerous items that provide additional insight and that are consistent with prudent operation and maintenance of " defense in depth" during the outage (Refs. 2 and 3). These include:
1.
The Unit I residual heat removal (RHR) pumps are not powered by the "0"
- diesel, 2.
Both Unit I high pressure injection systems will be available during cooldown (a system consists of a charging pump and an intermediate head safety injection pump),
3.
The IB charging pump receives power from bus 149 and will be available whenever there is fuel in the reactor vessel, 4.
Both intermediate head safety injection pumps will be available until the reactor vessel head has been removed, 5.
Gravity flow from the refueling water storage tank will always be available when the water level is below the reactor vessel flange and there is fuel in the vessel,
h,
?
1
.s 6.
General maintenance will be performed on emergency safeguards features buses 148 and 149 only when the "0" diesel is operable (these buses are not associated with the "0" diesel),
7.
The 18 and IC service water (SW) pumps will be available throughout the "0" diesel outage, and 8.
The component cooling water (CCW) pumps will be operated in accordance with technical specifications during the "0" diesel outage, j
We have no concerns with cooling the core in the reactor vessel if there is no I
fuel in the vessel. Maintaining the reactor vessel water level at or above an elevation that is three feet below the reactor vessel flange, as discussed in GL 88-17, removes most of the concern associated with loss of core cooling.
Provision of water addition capability' and cooling water support as discussed above provides additional assurance that the core can be kept covered and cooled should RHR be lost. These measures provide sufficient assurance that Unit I risk is not significantly impacted by the proposed amendment.
1 2.2 Unit 2-Removal of the "0" diesel means that only one centrifugal charging pump (part of the SI system) and one intermediate head SI pump are provided with emergency power via the remaining diesels. A loss of offsite power with the additional loss of one of the two remaining Unit 2 diesels would also cause loss of either the remaining charging pump or the remaining SI pump. The auxiliary feedwater pumps and the RHR pumps are not directly affected by the status of the "0" diesel.
Removal of the "0" diesel removes one SW pump and one CCW pump from an emergency power source. Two-SW pumps remain, as do two CCW pumps, with separate emergency power supplies. Loss of an additional diesel does not cause a complete loss of either SW or CCW. A similar condition exists for
-containment spray pumps and containment fan coolers in that loss of an additional diesel will not cause complete loss of the function (Ref. 1).
The ' licensee has addressed these conditions with additional action statements and by providing a justifiG; ion from the probabilistic views.
i The staff has reviewed the auxiliary power system which provide:, power to ESF equipment necessary for a safe shutdown, and noted the following:
1.
The switchyard maintains six-345kV offsite power sources which significantly exceed the requirement of 10 CFR Part 50, Appendix A, GDC 17.
The switchyard design (Ring-Bus Arrangement) can locate and isolate system faults rapidly. As a result, Zion has not experienced a complete loss of all offsite power (LOOP) since their criticality in 1973.
Consequently, the Zion Probabilistic Safety Study (ZPSS) reflects that the probability of occurrence of a total LOOP at Zion is considerably less than the industry average (i.e., 0.068 vs 0.194 Events / Site year),
i
l-
\\
- (
i The power sup(plies to ESF buses are available from either Unit Auxiliary 2.
i Transformer 241) or System Auxiliary Transformer (242) of the offsite power source as well as the adjacent unit's (Unit 1) offsite power source through its System Auxiliary Transformer (142).
If a backfeed capability through the Unit Auxiliary Transformer (241) is considered, a total of four offsite power supply paths to ESF buses are available. This exceeds the minimum requirement of GDC 17.
1 3.
The licensee made a commitment that while the "0" DG is out of service, DGs (1A and IB) of Unit I will be available to power two additional service water pumps and component cooling water pumps, and they will perform surveillance test of the remaining Units 1 & 2 DGs prior to placing "0" DG in the LCO. Also, the licensee proposed compensatory measures for the remaining DGs for the duration of 45 days. The compensatory measures consist of additional surveillance and more stringent action statements for the DGs.
4.
The licensee has reviewed and verified their existing emergency operating procedures (E0P) for their applicability during a LOCA with only one ESF bus available. The licensee stated that their E0P procedures allow for unit shutdown following a LOCA using either a single charging pump or a single SI pump.
In addition, the licensee based a portion of the amendment justification on PRA arguments that addressed the likelihood that a LOCA would cause a core melt accident (Ref. 1). Our review established that other potentially significant i
contributors to core melt were not addressed in the submittal, and that the licensee used risk interchangeably with core melt, which can be misleading and incorrect.
Seismic events with a loss of AC power are the dominant contributors to a large release (6E-6 per reactor year) (Ref. 4), and thus to risk. The seismic event involves failure of the SW system due to crib house failure and failure of all diesels and related switch gear due to auxiliary building structural failure. The "0" diesel outage has no calculated influence on these failures (Ref. 3).
The next largest contributor to a large release (2E-7) is a turbine trip due to a loss of offsite power. The first order electrical impact would involve the electrically driven auxiliary feedwater pumps. But the power supplies for these pumps do not involve the "0" diesel. Smaller effects involve additional 1
simultaneous failures such as LOCA or steam generator tube rupture, where an additional (third simultaneous) failure of a diesel or involving the SI system J
could influence water injection into the RCS. This is negligible in comparison to the seismic effect.
interfacing system LOCA (IE-7), LOCAs with failure 1
The above are followed by(small at 2E-9, large at SE-10, and medium at SE-10) of recirculation cooling and others. These are small in comparison to the above and are negligible j
i l
f, i
v.~
.5
.'when one adds another. failure, which is required for'the "0" diesel to have an.
L
' influence. -The licensee also states that failure of one additional diesel willnot.leadtoacoremelt(Ref.1).
The ordering of events leading to core melt are small LOCA (2E-5), Seismic-
~ ith loss of AC power (6E-6), large LOCA (SE-6), medium LOCA (SE-6), and a w
numberofothersintheE-6 range,withtheLOCAswelistedincludingfailure t
of. recirculation cooling. The seismic event is' unaffected by the'"O diesel for the reasons previously discussed.
For any of the core melt sequences to occur and be influenced by the "0" diesel outage one first must have a total loss of offsite power which has never occurred at Zion (Ref. 1).. During the "0" diesel. outage, there is one charging pump and
.one intermediate head.SI pump operable if there is no offsite power and both Unit 2 diesels are running. This is the minimum coverage provided under the design basis. Loss of one of the Unit 2 diesels would cause loss of one of these pumps. The possibility also exists that one of the pumps will fail, that some other failure will occur in an SI system, or that a LOCA involves one of-
. the SI : systems. Thus,'there are several failure possibilities, albeit with a small. likelihood..that leave Unit 2 with one pump capable of injecting water-into the RCS at intermediate pressure, and with no pumps capable of injecting water at high pressure. This.has potential implications for LOCA and transient scenarios.
The licensee addressed some aspects of these failures by considering LOCA
- conditions.- The licensee concluded that the snall break was the accident of most' potential' concern since for'large breaks, the RCS pressure will be low and low head pumps will provide water (Ref. 1). We agree. The reduced. number of pumps following loss of offsite power will not significantly contribute to increased risk for the large breaks. The same conclusion is valid for the intermediate size LOCAs which the licensee did not address in probabilistic analysis.
The design basis success criteria for small breaks include an operable charging pump. The licensee estimated that application of this success criterion to the "0" diesel outage would increase the likelihood of a core melt due to a small break by about a factor of ten. The licensee then addressed whether this was realistic, and noted that the success criterion in Reference 4 was one of four (charging or intermediate head) pumps. A number of analyses using the MAAP code were discussed (Ref. 1) from which the licensee concluded that the impact
.of the "0" diesel outage was calculated to be negligible to two significant figures. We are not convinced the MAAP code is fully adequate for these studies, but the licensee's conclusions appear reasonable when contrasted to other' studies.
(See,forexample, Reference 7.) We agree with the licensee's conclusion that the "0" diesel outage does not significantly impact upon core melt frequency due to LOCA.
The licensee did not perform a similar assessment of the potential impact upon other transients that are predicted to contribute to core melt likelihood (Ref. 4). Although such events are calculated to be smaller contributors than LOCA, they are not negligible when compared to LOCA. The licensee stated that
l 7
- f y
1 f
-6.
the core damage from such events requires the failure of the three train auxiliary feedwater system in addition to failure of the appropriate SI systems (Ref. 2).-l As we previously identified, the: auxiliary feedwater system is essentially unaffected by the "0" diesel outage. The licensee also stated that i
the core could be cooled by feed and bleed using' one of four (two charging, two
.l intermediate head) ~ pumps even.if all auxiliary feedwater was lost, -a conclusion that appears reasonable from the trends evident in Reference 7.
]'
d Based on the above, the staff. determined that simultaneous loss of all offsite power (i.e., 6 offsite lines and 4 paths to ESF buses) and a LOCA concurrent with additional single failure of-a DG is a very unlikely event over the 45-day LCO period while the "0" DG is out for maintenance and, the risk and likelihood of core melt are not significantly.affected by Unit 2 operation while the "0" DG is out for maintenance..However, the staff had several discussions with the licensee regarding the prudency of such repeat-requests every five years for extended maintenance.
In the August 24,.1989 letter, the licensee stated that it does not plan to ask for this type of relief in the future, and it will perform a study to determine alternate methods to perform the preventive maintenance of "0" diesel. These alternatives would include, but not limited to those that would modify existing plant design or operational changes to plant evolutions. The licensee is planning to discuss its findings with the staff in'the near future.
The staff finds that the proposed extension of the allowable out-of-service period to allow for maintenance of the "0" diesel generator is acceptable and poses no undue risk to the public health and safety. This conclusion is based upon numerous compensatory actions planned by the licensee during this outage (e.g., removal of fuel from Unit I for most of the time the "0" diesel is unavailable and performance of maintenance on emergency safeguards features buses only when the "0" diesel is operable), and the likelihood of the event that may cause a significant damage to the core.
Since the licensee has indicated that specific plans may undergo modification during the overall refueling outage, the staff requires that the licensee i.
specifically evaluate the impact of any modifications on plant risk during the l
"0" diesel generator maintenance period.
3.0 FINDING OF NO SIGNIFICANT IMPACT Pursuant to 10 CFR 51.32 an environmental assessment and finding of no i
significant impact has been prepared and published in the Federal Register on September 7,1989,1989 (54 FR 37170
).
Based upon the environmental assessment, the Commission has determined that the issuance of this amendment will have no significant effect on the quality of the human environment and that no environmental impact statement need be prepared.
4.0. CONCLUSION We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such
l {.-
3 l..
activities will.be conducted in compliance with the,Comission's regulations,
'and (3) the issuance of these amendments will not be. inimical to the common c
defense and security or to the health and safety of the public.
5.0' REFERENCES
'1.
G.. E. Trzyna, " Zion Nuclear Power Station, Units 1 and'2, License Nos..
DPR-39 and DPR-48, Proposed Amendment to Technical Specification, Section.
3.15 " Auxiliary Electric Power System," letter to Director of Nuclear Reactor Regulation from Commonwealth Edison, July 6,;1989.
2.
G. E. Trzyna, " Zion Nuclear Power Station, Units 1 and 2. Proposed Amendment to Technical Specifications, Section 3.15 ' Auxiliary Electric Power System', License Nos. DPR-39 and DPR-48, NRC Docket Nos. 50-295 and 50-304" letter to Director of Nuclear Reactor Regulation from Commonwealth Edison, August-4.1989.
3.-
S. C. Hunsader, " Zion Nuclear Power Station, Units 1 and 2, Proposed Amendment to Technical Specifications Section 3.15, ' Auxiliary Electric 1
Power System,' License Nos. DPR-39 and DPR-48, NRC Docket Nos. 50-295 and 50-304," letter to Dr. Thomas E. Murley (NRC), August 10, 1989.
-4.
" Zion Probabilistic Safety Study," Commonwealth Edison, Se stember 1981 (affiliation..and date provided from knowledge of report, tiey are not containedonthe.coverandtitlepage).
- 5..
L. A. Bowen, et. al., " Zion Nuclear Plant Residual Heat Removal PRA,"
report prepared by Pickard, Lowe and Garrick, Inc. for the Nuclear Safety Analysis Center, NSAC-84, July 1985.
6.
D. M. Crutchfield, " Loss of Decay Heat Removal (Generic Letter No. 88-17),
10 CFR 50.54(f)," NRC letter to all holders of operating licenses or constructionpermitsforpressurizedwaterreactors(PWRs), October 17, 1988.
7.
B. E. Boyach, et al., "Los Alamos PWR Decay-Heat-Removal Studies, Sumary Results and Conclusions," NUREG/CR-4471, LA-10637-MS, Los Alamos National Laboratory, March 1986.
8.
G.E. Trzyna, " Zion Nuclear Power Station, Units 1 and 2 Proposed TechnicalSpecificationAmendment,Section3.15,'AuxillaryElectric Power System; License Nos. DPR-39 and DPR-48, NRC Docket Nos. 50-295 and 50-304," letter to Director of Nuc' lear Reactor Regulation from Commonwealth Edison, August 24, 1989.
Principal Contributor:
W. Lyon, NRR/ DEST P. Kang, NRR/ DEST C. Patel, NRR/DRSP Dated: September 8, 1989 u _ _ _ ___ _ ----------
l
,J n,.
a2
~
1 n
m m$
p
+
z I' 3
v'.
_J t '
u~
o h G' B 1
1 0
2 1
3' 3
2 tD2
~
i W
t'uo=
hGA 1
O 1
2 1
3 3
2 tD2 i
W g
n u'f i
^
=
d g
e
)
'f h
r~
5 o
c
- )
~
6 2'-
i
(
(
unS 1
1 r
1 1
12 3
3 qIS o
4 e
P y
1 R
Z n
I A
S V
. N O
~
.O M
I g
n T
d n
)
n o
~
U e
i
. (4 e
i B
r s
p s
I inn 12 1
1 1
1 2
2 3
'o s
R uI e e
T q
c o
r T
S e.i t
p sD 9.
p I
R L
u 4
s R
2
~
E e
A
.W t
s r
C O
u~
u u
O r
P o
) -
)
)
)
B s
L hG0 1 ~
2 3
3 s
m
.e r
T tD"-
(
(
(
(
N i
2 1
1 3
2 4
4 4
o r
o E
W r
p f M
f
~
P r
s I
r o
p U
e f.
m Q
w u
E o
.s p
p p
.m 2
F B
S 2s9 s
.u E
u4 X
X X
X X
X X
e p d GB2
'i 2
a r
n D
2 u
- r q
d)
~
t e
nSe i
r 1aSw 1
n
)
Po U
t%tlZp E
u0ia(
L A-b0nv
.e B
2s8 1U o yt A
u4 X
X X
X X
X X
(
mdi T
GB2 n
m e u s.
D eportf vmr Sf i ufe o
rp ny d
eitf nldeo s
Gebof s
p Dval as p
m il Ss m
u srir o
s u
p s
idaocl p
s p
p r
vfi m
p l
m e
sma tr u-m g
s a
u l
pa pso p
u n
p v
p s
o mesmif V'
p i
m o
p o
ut pul
)
y g
u m
g m
c p s m pi p 0
a n
r p
e n
u u
bm 6
r o
a r
i p
n o a p1 a u 4
s -
i h
r l
a w
bp 1
p t
c e
t o
r f
tssso-(
c t
a o
e eeer1 t
e l
a e
c t
t fddrP -
s n
j a
w h
a n
ouui s
t e
n g
d t
w e
ll und n
m i
u e
l n
m eccqoe e
n f
e a
e e
n nnneie y i f
u n
c i
OIIRZN n
i o
a t
r d
o i
a p
t e
t i
p v
t m
n f
n x
s m
r n
))))))
o o
a e
u e
o e
o 123456 C
C S
C A
R C
S C
((((((
- :. : u lr.jl