ML20207M350
| ML20207M350 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 10/10/1988 |
| From: | Robert Williams PUBLIC SERVICE CO. OF COLORADO |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| P-88312, TAC-67896, NUDOCS 8810180235 | |
| Download: ML20207M350 (10) | |
Text
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O Public Service ~
nh October 10, 1988 o.nd$'cEso201os40 Fort St. Vrain Unit No. 1 RO WILUAMS, JR.
P-88312 VCE PRES:OENT NUCLE AR OPERAf TONS U.S. Nuclear Regulatory Commission ATTN:
Document Control Deck Washington, D.C.
20555 Docket No. 50-267
SUBJECT:
PSC Response to NRC Request for Additional Information
REFERENCE:
- 1. NRC letter, K.L. Heitner to R.0, Williams, dated 7/19/88 (G-88285)
- 2. PSC letter, R.0, Williams, Jr.
to Document Control Desk, dated 3/18/88 (P-88098)
Gentlemen:
The purpose of this letter is to submit Public Service Company of Colorado's (PSC) responses to NRC's four recuests for additional information contained in Reference 1.
PSC understands that this request for information is required to complete NRC's review of the Core Support Floor Casing Leak Safety Evaluation which was submitted by Reference 2. provides PSC's responses to NRC's requests.
If you have any questions concerning Attachment :. please contact Mr. M. H. Holmes at (303) 480-6960.
Very truly yours, 9
/IW R. O. Williams, Jr.
Senior Vice President Nuclear Operations i
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P-88312 October 10 1988 Attachment cc: Regional Administrator, Region IV ATTN: Mr. T. F. Westerman Chief, Projects Section 8 Mr. Robert Farrell Senior Resident Inspector Fort St. Vrain l
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Attschment 1 P-88312 Page 1 PSC RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION NRC Letter K.L. Heitner to R.O. Williams, dated 7/19/88 (G-88285)
NRC REQUEST 1 r
The licensee's safety evaluation utilized a value of 20 lb/hr for the leak rate through the core support floor vent under accident cont" tions.
This value is based on doubling the highest measured i
lea rate. Reference is made to the design capacity of the core support floor vent of 14 lb/hr under design circulating activity conditions. Since a leak rate of 20 lb/hr would not be acceptable i
under design circulating activity conditions, what actions, if any, will be taken if the measured core support floor leak rate exceeds 14 lb/hr?
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PSC RESPONSE 1 i
FSAR Section 11.1.1 states, "The release of radioactive materials in l
plant effluents will be within the limits specified in 20.106 of 10CFR20."
Technical Specification ELCO 8.1.1 Basis states, "...the licensee is permitted flexibility of operation, compatible with t
considerations of health and sa'ety, to assure that the public is provided a dependable source of power, e en under unusual operation conditions, which may temporarily result in releases higher than small fractions of, but still within, limits specified in 20.106 of 10CFR20".
j FSAR Section 11.1.3.4, in the discussion of core support floor (CSF) leakage, states, "The maximum allowable leak rate has been determined based on continuous radioactive release from the plant at the MPC
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levels given by 10CFR20 This results in an allowable leak rate of f
1.0 Ci/ minute, which would be equivalent to 14 lbs/hr at design primary coolant activity." PSC requested that General Atomics (GA) perfom calculations to confirm that a continuous CSF leak rate of 14 I
lb/hr at design levels of circulating activity would result in j
10CFR20.106 maximum permissible concen+ rations (MPC) allowed. GA's calculations have detemined that a continuous CU leak rate of 14 l
lb/hr results in reaching 42% of the 10CFR20.106 MPC levels.
A l
continuous CSF leak of 33 lb/hr at design levels of circulating l
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P-88312 Page 2 activity results in reaching the 10CFR20.106 MPC levels. The major assumptions used in these new calculations are:
CSF vent gas is comprised of design primary circulating gas-borne activity of FSAR Table 3.7-1, 879 MW(t) design column.
No credit is taken for filtration in the Gas Waste System.
The Reactor Building Ventilation Exhaust stack filters have filtration efficiencies equal to the minimum allowed by Technical Specification SR 5.5.3.
7370 lb of helium in the primary circuit.
No credit is taken for decay during time to reach unrestricted areas (exclusion area boundary),.
Use the average annual dilution factor of 1.16 E-06 Sec/(Cubic meter)
(FSAR Section 11.1.1).
The CSF leak rate is continuous over a one year period.
The radionuclides modeled were Tritium, all Krypton isotopes, all Xenon isotopes, and all Iodine isotopes, appearing in FSAR Table 3.7-1.
GA Report No. 909763 dated 10/07/88 documents the calculations supporting the above results.
Therefore, PSC concludes that the 20 lb/hr CSF leak rate assumed i
in the CSF Casing Leak Safety Evaluation (P-88098) would result I
in activity concentrations below the 10CFR20.106 limits, even in the event of a continuous CSF leak at design levels of circulating activity, neither of which has been observed.
No additional actions would be taken unless the CSF continuous leak rate were to exceed 20 lb/hr.
FSAR Section 11.1.3.4 will be updated in Revision 7 to reflect the CSF leak rate that corresponds to the MPC values allowed by 10CFR20.106.
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o P-88312 Page 3 sNRC RF00EST 2 Reference is made to the maximum capacity of the gas waste system of 53 lb/hr. What is the maximum capacity of the core support floor vent line, assuming no other sources of gas into the gas waste system?
PSC RESPONSE 2 Assuming that no other sources of gas are vented into the gas waste system, the core support floor (CSF) vent line capacity is limited to that which would lift its relief valve, V-6389.
Calculations documented in EE-63-0001 Rev A show that over 18 lb/hr of CSF vent line flow is required to reach the current 10 psig set pressure opening of V-6389. This capacity is well above the 14 lb/hr CSF vent flow mentioned in Section 11.1.1.3 of the FSAR and the maximum recently measured CSF venting rate of about 10 lb/hr.
It is noted that, if necessary, the set pressure of V-6389 could be raised to 15 psig, which would raise the maximum capacity of the CSF vent line to about 22 lb/hr, exceeding the 20 lb/hr CSF leak rate discussed in PSC Response 1.
Attachmen2 1 P-88312 Page 4 NRC REQUEST 3 The basis of the operating pressure of 60 psig as stated in the FSAR is not clear.
What was the basis of the 60 psig value and how will operation at 100 psig relate to those factors?
PSC RESPONSE 3 L
No analyses have been found that substantiate the basis for the 60 psig set point of back pressure control valve PV-6364.
Recognizing this situation, PSC prepared the safety evaluation attached to PSC letter, R.0. Williams, Jr.
to Document Control i
Desk, dated March 18, 1988 (P-88098). The purpose of this safety evaluation was to justify the 60 psig and the desired 100 psig set points based upon both past and present operating experience.
The following is an explanation of the rationale for the 60 psig value based upon discussions with General Atomics (GA) personnel and reviews of the FSAR and other related design documents:
The steel liner of the Core Support Floor (CSF) and its support columns are designed to withstand an external pressure of up to the Pressurized Concrete Reactor Vessel (PCRV) reference pressure.
All vent piping extending from the vicinity of the CSF's internal concrete and support columns to the system's remotely opera ted isolation valve, HV-1195, is designed to withstand an internal pressure of up to PCRV reference pressure.
Downstream of HV-1195, the vent system l
piping is designed to PCRV reference pressure while l
vessels such as filters and tanks are designed to the 1
gas waste system's design pressure of 15 psig.
1 Pressurization of the CSF internals is deemed credible 1
from two sources:
- 1) primary coolant the CSF through a breach of the CSF liner; orpassing )into 2 water leaking into the CSF from a ruptored system 46 cooling tube.
The CSF was designed such that vent tubes are spaced within the CSF liner and concrete structure to prevent the internal pressure of the CSF from ever exceeding i
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P-88312 Page 5 cooling water pressure even with the larcest credible leak of primary coolant into the C5F, as ciscussed in FSAR Sections 3.3.2.2 and 5.9.2.4.
Therefore, a
ruptured liner cooling water tube was evaluated as the potential pressure source for the CSF internals and vent system piping. The cooling water pressure is 102 I
psig when it exits the system 46 pumps.
The system flow losses balance out the effect of the increased pressure head due to the change in elevation as this l
water flows down to the CSF.
The water entering the CSF subheaders is 97 psig per system design process flow drawing PF-46-2 From this point the water must i
flow up to the CSF elevation.
Due to this decrease in i
elevation pressure, water exiting the CSF is at 48 l
psig, based on PF-46-2.
Therefore, cooling water within the CSF falls within a pressure range of 48 to 85 psig.
Based on these values, a back pressure control valve with a range of 0 to 100 psig was chosen for the CSF vent system. Because a controller will operate most efficiently and accurately when set near its midpoint, a 60 psig control setting was chosen.
By controlling the CSF internal pressure at 60 psig, either leak path would be directed into the CSF internals.
From there, the leaking fluid pressure can be controlled and discharged through the CSF vent system and either the gas waste or liquid waste systems.
Currently the CSF internal pressure control point is not permitted to exceed 100 psig.
This is based completely upon structural considerations of the CSF and its steel liner as described in the P-88089 safety evaluation.
The increased setpoint serves to reduce l
primary coolant in-leakage to the CSF by decreasing the I
differential pressure across the liner breach.
Primary coolant leakage into the 46 system through any liner cooling water tube leaks internal to the CSF would increase as a result of the 100 psig (vice 60 psig) set pressure.
Primary coolant which enters the 46 system is collected in the liner cooling water surge tanks and is then vented into the gas waste system, r
Attachmon2 1 P-88312 Page 6 NR,C RE0 VEST 4 On July 6, 1988, the licensee reported that Valve V-111063 was found mispositioned. Mispositioning or failure of this valve could lead to failure of the vent system to perform its design function.
Provide a i
safety evaluation addressing the following issues concerning I
potential failure of this valve including:
1 a.
Administrative controls on the valve position b.
Alternative indications of CSF overpressurization i
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Need for a redundant flow path, and d.
Consequences of an overpressurization accident.
PSC RESPONSE 4 PSC has prepared and submitted a safety evaluation on the mispositioned valve V-111063.
The safety evaluation is included in Licensing Event Report (LER)88-011, dated August 5, 1988, which was transmitted to the NRC by PSC letter, C.H. Fuller to Document Control Desk, dated August 5, 1988 (P-88290),
a.
The corrective action in LER 88-011 addressed administrative controls on the position of valve V-111063 as follows:
"Valve V-111063 will be sealed open and added to the Sealed and Critical Valve Check List (SR-0P-12-W).
Valves on this checklist are verified to be sealed in the proper position on a weekly basis."
It is noted that this corrective a t'on has been completed and has also added Yalves V-11726, -6380 -6381. -6382, -63100 and 63202 to SR-0P-12-W as an additional measure of assurance that the CSF vent path will remain open.
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e P-88312 Page 7 b.
The corrective action in LER 88-011 addressed alternative indications of CSF overpressure as follows:
"To provide a.teans of identifyiiig inadvertent isolation of the CSF vent in a timely manner, pressure indicating instrumentation will be added upstream of Y-111063.
This location will indicate as nearly as possible the pressure in the CSF itself.
This corrective action will be completed prior to startup."
c.
Operation of the CSF vent system with V-111063 partially closed was a result of an undetected passive failure of the damaged valve caused by overloading the valve hand-wheel.
The vent system was not designed to withstand a single passive failure.
It was designed to withstand a single active failure. The only applicable active components are back pressure control valve PV-6364 and block valve HV-1195, both of which are designed to fail open on loss of operating power or control signal.
Isolating the CSF cavity by a passive failure of V-111063 or any other manual valve in the CSF vent path will be minimized by the corrective actions discussed in LER 88-011, and in the following paragraph.
Corrective actions have been implemented by CN-2836 to preclude the possibility of inadvertent overpressurization of the CSF.
This CN relocated the pressure tap for CSF pressure indication to upstream of V-111063, replaced V-111063 with a
corrosion retistant diaphragm type manual valve which is not susceptible to binding like the original valve, and added an alternate vent path that leads directly to the gas waste vacuum tank. The alternate vent path ties into the CSF vent header upstream of V-111063.
The manual isolation valve in the alternate vent path will be closed during normal operation.
d.
The consequences of an overpressurization accident were considered in LER 88-01).
During this event, the CSF was not completely isolated since V-111063 was not completely shut, allowing some venting of the CSF. When the CSF liner cooling tube leaks opened up during the depressurization of the PCRV, additional venting of the C3F occurred. Also, a comparison of the PCRV depressurization during this event with the depressurization discussed in PSC's March 18, 1988 (P-88098) subrtittal showed that this event's depressurization was much slower than that assumed for the DBA-1 case.
This slower rate of depressurization allowed more time for the CSF internal pressure
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P-88312 l
Page 8 to equalize with PCRV pressure as the PCRV was depressurized.
The consequences of this event are therefore enveloped by the consequences of the DBA-1 analysis included in P-88098, and it is concluded that at no time during this event did the CSF internal pressure exceed the PCRV pressure by more than 210 psi.
As discussed in P-83098, the differential pressure of 210 psi has been calculated to be that conservative value at which deformation of the CSF top head liner is postulated to occur due to pullout of the top liner anchor studs from the concrete inside the CSF.
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