ML20236Q056
| ML20236Q056 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/16/1987 |
| From: | Tiernan J BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| GL-87-12, NUDOCS 8711190159 | |
| Download: ML20236Q056 (7) | |
Text
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A
' BALTIMORE GAS AND 1
ELECTRIC 1
CHARLES CENTER R O. BOX 1475 BALTIMORE, MARYLAND 21203 JostPH A.TitRNAN I
- VicE PRESIDENT NUCLEAR ENERGY November 16,1987 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:
Document Control Desk
SUBJECT:
Calvert Cliffs Nuclear ' Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 i
Response to Generic Letter 87-12, Loss of Residual Heat Removal While the Reactor Coolant System is Partially Filled
REFERENCES:
(a) Letter from Mr. J. A. Tiernan (BG&E) to NRC Document Control Desk, dated September 14, 1987, same subject i
Gentlemen:
Reference (a) provided a description of Calvert Cliffs' plant operation during the approach to, and with, a partially filled Reactor Coolant System (RCS). In addition, we stated that we were working with the Combustion Engineering Owner's Group (CEOG) to j
review one concern presented in the Generic ' Letter. Specifically, the concern-is a l
loss of residual heat removal (RHR) leading to RCS ' pressurization and potential ejection of coolant via a cold leg opening (e.g.,
during maintenance 'on a RCP seal assembly).
Combustion Engineering (CE) has qualitatively assessed the above loss of RHR scenarios-4 and has issued a report which we have summarized in Enclosure (1).'. The report provides bounding conditions (i.e.,
injection capability and steam generator availability) for-CE design plants to show core protection based on the given type of RCP maintenance I
We are in the process of reviewing certain areas of the report using - plant specific data.
In the interim, we will conservatively maintain. a High Pressure Safety Injection pump available whenever we are in a MODE 5 partial filled condition,
'i i
.0 8711190159 871116 i
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ADOCK 05000317 PDRe L_~
M Document Control Detk November 16, l'987 N
Page 2 j
Should you have further questions regarding this matter, we will ' be pleased ' to discuss them with you.
1 Very truly yours, X
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l-STATE OFMARYLAND :
- TO WIT:
CITY OF BALTIMORE Joseph A. Tiernan, being duly sworn states that he is Vice President of the Baltimore Gas and Electric Company, a corporation of the State' of Maryland; that he. provides the foregoing response ' for the purposes therein set'. forth;. ' that.the'. statements.. made. are true and correct to the best of his knowledge, info'rmatio'n,~ and belief; and that he was.
authorized to provide the response on behalf of said Corporation.
WITNESS my Hand and Notarial Seal:
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Notary Publ[
My Commission Expires:
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Attachment i
JAT/LSL/ dim 1
cc:
D. A. Brune, Esquire J. E.
Silberg, Esquire i
R. A.Capra, NRC S. A.McNeil,NRC W. T. Russell, NRC T. Foley/D. C. Trimble, NRC i
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.--___._____L w
ENCLOSURE m CALVERT CLIFFS RESPONSE TO GENERIC LETTER 87-12 In our response to Generic Letter 87-12, we provided a description of Calvert Cliffs plant operation during the approach to, and with, a partial filled Reactor Coolant System (RCS). In addition, we stated that = we were working with the. Combustion Engineering Owner's Group (CEOG) to review one concern presented in the Generic Letter.
Specifically, the Generic Letter's concern is a loss of residual heat removal (RHR),
which leads sequentially to bulk boiling of coolant in the reactor vessel, RCS pressurization, and displacement of liquid level from the vessel and from the loop seals into the four discharge legs.. As the increasing water level in the discharge legs reaches the top of the leg where a Reactor Coolant Pump (RCP) is assumed to be partially disassembled for repairs, liquid is lost from the RCS. This liquid loss will continue until the decreasing level in the " pressurized" (i.e.,
vessel, hot legs and steam generator primary volumes) portion of the RCS is low enough to allow the steam-air mixture to " vent" through the loop seal to the leak location at the RCP.
The above venting will allow equalization of pressures between the pressurized portion i
of the RCS and the relatively lower pressure portion of the RLS. Therefore, vessel level may be at least partially restored as the liquid inventory in the RCP til;ch9rce legs can flow back into the Reactor Vessel. Following the initial venting, subsequent repressurizations and venting cycles may take place.
Combustion Engineering (CE) has qualitatively assessed the above loss of RHR scenarios and has issued a report which we have summarized below. The outcome of the. above scenario is a function of decay heat, plant geometry, the size of the ejection flow path in the cold leg, and the dynamics of the venting and level equalization process.
The intent of CE's assessment was to qualitatively provide the bounding conditions which showed core protection for all CE design plants. Their assessment considered a matrix of 16 possible cases in terms of steam generator (SG) availability, integrity of the discharge leg via RCP integrity, RCS injection via High Pressure Safety Injection (HPSI) or charging, and loop seal elevations relative to the top of the active core.
The analyse used generic CE data which is typical for all CE plants since their RCS volume to reactor power ratios are essentially equal. One-half percent decay heat, corresponding to about one day after reactor trip at full power and equilibrium fission products, was assumed.
Based on the report, the bounding conditions for CE design plants to provide core protection in a partial filled condition are shown in Table
- 1. They are presented in
{
the format of necessary injection capability and SG availability for either no RCP maintenance or RCP seat maintenance. Maintenance beyond RCP seals is discussed later in this report.
l
_1-
ENCLOSURE (1)
CALVERT CLIFFS RESPONSE GENERIC LETTER 87-12 f
TABLE 1 Degree of RCP Maintenance No RCP Maintenance RCP Seal Maintenance Necessary Injection I charging pump and I charging pump and Capability a SG available a SG available and SG Availability
- or -
- or -
I I HPSI pump 1 HPSI pump For the case in which a charging pump is used for injection during MODE 5 partial t
I filled conditions, the study showed that if RHR were lost, an available SG as a heat sink would maintain RCS pressure less than 50 psia. This will eliminate potential leakage paths through either the PORV at the Low Temperature Overpressure Protection (LTOP) setpoint (400 psia), or the RCP, if the seal assembly it removed. RCP leakage l
via the removed seals is a torturous path and is available only when RCS pressure l
exceeds the lift pressure
(~ 83 psig) for the rotating assembly to allow leakage past I
the RCP shaft stop seal. This stop seal provides a metal-to-metal seat upon uncoupling of the motor and pump shafts. For cases in which maintenance beyond a RCP seal is being performed, the exact flow path size at whi.ch a charging pump with a SG is I
sufficient was not rigorously quantified.
For CE plants with elevated loop seals, such as Calvert Cliffs, the presence of water
)
in the RCP discharge leg from HPSI injection infers core coverage. Therefore, a HPSI pump would have adequate flow to maintain the core covered for all size RCP openings.
An available SG was not necessary to assure core cooling.
To assure its applicability to Calvert Cliffs, we are reviewing certain areas of the report on a plant specific basis. In the interim, we will conservatively maintain a HPSI pump available whenever we are in a MODE 5 partial filled condition. As discussed j
in our September 1987 response, Calvert Cliffs maintains SG levels as follows for proper SG chemistry. For short-term shutdowns, SGs are filled to the top of the indicating range; for long-term shutdowns, SGs are filled above~ the moisture separators. If maintenance is being performed on the secondary side of the SGs, normally only one SG at a time is drained.
Combustion Engineering also made a preliminary dose assessment of the thyroid and whole body doses at the site boundary associated with a loss of RHR event. Enclosures (2) and (3) provide the assumptions (which include no containment integrity) as well as RCS
~
specific activities. The analysis conducted summed the time dependent contribution from all appropriate isotopes at the site boundary. The dose results were 31 mrem (I-131) and 15 mrem (Beta-Gamma) for the thyroid and whole body, respectively. _ _ - _ _ _
ENCLOSURE (1)
CALVERT CLIFFS RESPONSE GENERIC LE' ITER 87-12 1
l i
i Although there is no established acceptance criteria.for a loss of RHR accident, the above doses are small when compared to typical FSAR doses for Design Basis Events.
Therefore, a loss of RHR which does not produce fuel damage is not considered a-threat to the. health and safety of the public.
In
- addition, as stated in our j
September 14, 1987 response, our Abnormal Operating Procedure-3B '(Loss of Shutdown l
Cooling) requires Technical Specification. 3.9.4,. regarding containment integrity, to be implemented an'vtime Shutdown Cooling is lost, l
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l _ _ _ _ _ _ _ _ _ _ _ _
J
I ENCLOSURE (2)
-)
PRELIMINARY DOSE ASSESSMENT ASSUMPTIONS FOR A LOSS OFRESIDUAL HEAT REMOVAL The below assumptions were used in.CE's preliminary dose assessment for - thyroid and whole body doses at the site boundary, associated with a loss of RHR.
o RCS Integrity; none l
0 Containment Integrity:
none o
Time after Reactor Trip:
- 1. day o
Initial Liquid Inventory in the RCS: at mid-loop o
RCS Specific Activities: CE System 80 data used, based on ANSI N237
(" Specific Activity of Radioactive Waste Materials in Principal Fluid j
l Streams of Light Water Cool Nuclear Power Plants"), a copy of the specific j
activity data considered is provided at the end of this section o
Site Meteorology; per Reg. Guide 1.4 o
Credit Taken for Purification or. Gas Stripping in the CVCS. During. the Shutdown Process: none 1
o Boiloff Rate of Coolant: 15 lbm/sec l
o Volatile Fission Products Released at a Constant Rate
~
1 o
Release Source Model: point source o
Point of Absorption Analysis via Standard Semi-Infinite Cloud Model with Dispersion j
j
h ENCLOSURE (3)
~
y l
I l
l Reactor Coolant System Specific Activities Ouring Normal Operations Including Anticipated Operational Occurrences Specific Activity Specific Activity Nuclide 970*F uCf/cc Nucifde 970*F uCf/cc KR-83e 2.1 (-2)"
Y-93 3.4 (-5)
I KR-85e 1.1 (-1)
ZR-95 6.0 (-5)
I KR-85 1.5 (-1)
NS-95 5.0 (-5) 1 KR-87 6.0 (-2)
MD-99 8.4 (-2).
KR-88 2.0 (-1)
TC-99m 14.8 (-2)
KR-89 5.0 (-3)
RU-103 4.5 (-5)
XE-131a 1.1 (-1)
RU-106 1.0 (-5)
XE-133a 2.2 (-1)-
RH-103m 4.5 (-5) i XE-133 1.8 (+1)
RM-106 1.0 (-5)
XE-135e 1.3 (-2)
TE-125e 2.9 (-5)
XE-135 3.5 (-1)
TE-127m 2.8 (-4)
XE-137 9.0 (-3)
TE-127 8.5 (-4)
XE-138 4.4 (-2)
TE-129m 1.4 (-3)
BR-83 4.8 (-3)
TE-129 1.6 (-3) i BR-84 2.6 (-3)
TE-131m 2.5 (-3)
BR-85 3.0 (-4)
TE-131 1.1 (-3)
I-130 2.1 (-3)
TE-132 2.7 (-2) i I-131 2.7 (-1)
BA-137a 1.6 (-2) l l
I-132 1.0 (-1) 8A-140 2.2 (-4) 1-133 3.8 (-1)
LA-140 1.5 (-4)
I-134 4.7 (-2)
CE-141 7.0 (-5)
I-135 1.9 (-1)
CE-143 4.0 (-5)
R8-86 8.5 (-5)
CE-144 3.3 (-5)
RS-88 2.0 (-1)
PR-143 5.0 (-5)
C5-134 2.5 (-2)
PR-144 3.3 (-5)
C5-136 1.3 (-2)
NP-239 1.2 (-3)
C5-137 1.8 (-2)
N-16 1.4 (+2)**
CR-51 1.9 (-3)
H-3 1.0 ( 0)
MN-54 3.1 (-4)
SR-89 3.5 (-4)
FE-55 1.6 (-3)
SR-90 1.0 (-5)
FE-59 1.0 (-3)
SR-91 6.5 (-4)
CO-58 1.6 (-2)
Y-90 1.2 (-6)
CD-60 2.0 (-3)
Y-91m 3.6 (-4)
Y-91 6.4 (-5)
- Nuncers in parentheses denote powers of ten
- This is the specific activity of d-16 at the reactor vessel outlet nozzle
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