ML20147C289
| ML20147C289 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 02/22/1988 |
| From: | Bird R BOSTON EDISON CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| BECO-88-025, BECO-88-25, NUDOCS 8803030044 | |
| Download: ML20147C289 (26) | |
Text
.
y SEP Program
+
BOSTON EDISON Executive Offices 800 Boylston Street Boston, Massachusetts 02199 Ralph G. Bird BECo 88- 025 Senior Vice President - Nuclear U. S. Nuclear Regulatory Commission February 22,' 1988 Document Control Desk Hashington, DC 20555 License DPR-35 Docket 50-293 ASSESSMENT OF PILGRIM SAFETY ENHANCEMENT PROGRAM
Reference:
- 1. NRC letter, S', A. Varga to R. G. Bird "Initial Assessment of Pilgrim Safety Enhancement Program",
dated August 21, 1987.
- 2. BECo letter, R. G. Bird to S. A. Varga "Information Regarding Pilgrim Station Safety Enhancement Program,"Letter No.87-111 dated July 8, 1987.
Dear Sir:
The purpose of this letter is to provide additional information in response to the NRC staff's request (Reference 1) regarding the Pilgrim Safety Enhancement Program (SEP), as submitted in Reference 2.
The information contained in the attachment to this letter responds to the staff's requests except for those related to the Direct Torus Vent System.
Based on discussions between Mr. J. E. Howard (Boston Edison) and the NRC staff during the period September 23 and 24, 1987, we are deferring our response to the staff's question regarding the Direct Torus Vent System until such time as we can complete additional modeling and analytical work.
I B
Attachment:
Assessment of Pilgrim Safety Enhancement Program HGL/amm/1549 500\\
cc:
See next page il 1
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A 8803000044 880222 PDR ADOCK 05000293 G
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BOSTON EDIS0N COMPANY L
f Februarv 22' 1988 U. S. Nuclear, Regulatory Commission Page 2 cc:
Mr. D. G. Mcdonald, Project Manager Division of Reactor Projects I/II Office of Nuclear Reactor Regulation Mail Stop:
1401 U. S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike Rockville, MD 20852 U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 NRC Senior Resident Inspector Pilgrim Nuclear Power Station 1
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1 Attachment to BECo Letter No.88-025 ASSESSMENT OF PILGRIM SAFETY ENHANCEMENT PROGRAM Contents:
1.
Text (4 Pages) 2.
Figure 1, Additional Sources of Hater for RPV injection and Containment Spray 3.
Figure 2, Backup Nitrogen System 4.
Teledyne Calculation CP-6695 G-1 l
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4 Assessment of Pilarim Safety Enhancement Program 1.
Section 3.4 - Additional Sources of Hater for RPV Iniection and Containment Sorav NRC Reauest The staff requests clarification regarding the modification to the RHR system to provide additional sources of water for RPV injection and containment spray.
This modification may require a change to the Technical Specifications. As described in the enclosure, the valves to be added to the RHR system become part of the reactor coolant pressure boundary during operatica of the RHR system and, consequently, are subject to surveillance testing.
BECo Resoonse No changes are required to the Technical Specifications due to the addition of gate valve 10-H0-511 and check valve 10-CK-510 to the RHR system.
The reactor coolant pressure boundary is designed, fabricated, constructed, and tested to have an extremely low probability of abnormal leakage, or rapidly propagating failure, or of gross rupture.
The reactor coolant pressure boundary consists of all those pressure-retaining components connected to the reactor coolant system, up to and including the outermost containment isolation valve in system piping which penetrates primary reactor containment per 10CFR50.2.
Primary containment isolation provides protection against the consequences of accidents involving the gross release of radioactive materials from the fuel and/or the gross release of reactor coolant inside or outside the primary containment by closing appropriate isolation valves in a timely manner.
Gate valve 10-H0-511 and check valve 10-CK-510 are connected to the RHR system outside the outermost containment isolation valve; they themselves are not containment isolation valves and during normal plant operation are not a part of the reactor coolant pressure boundary. However, under LOCA conditions with RHR operating in the LPCI or containment cooling modes, gate valve 10-H0-511 will be exposed to the pressure of the RHR system.
Gate valve 10-HO-511 will be maintained in a locked closed position during all events or conditions analyzed in the FSAR. A normally open drain line installed between gate valve 10-H0-511 and check valve 10-CK-510 will maintain the section of piping between the RHR and fire water systems empty.
In this configuration, only gate valve 10-H0-511 will be a part of the RHR system pressure boundary during RHR system operation within the PNPS design basis.
The valve will not be opened and will perform a passive, pressure boundary function only during events within the PNPS design basis.
Check valve 10-CK-510 will become a part of the RHR system pressure boundary only when the fire water to RHR crosstie is being used.
This configuration will only be used during severe accident scenarios beyond the design basis of PNPS.
1
The original plant design incorporated a similar connection to the RHR system from the salt service water system as shown on Figure 1.
The isolation valve for this connection (i.e. 10-H0-3820) is also maintained in a locked closed position to act as a pressure boundary for the RHR system. Valve 10-H0-3820 is not a containment isolation valve and it is not included in Technical Specifications.
The only valve leak testing contained within the Technical Specifications is that required by 10CFR50, Appendix J: Leak rate testing of containment isolation valves. Check valve 10-CK-510 and gate valve 10-H0-511 are not containment isolation valves and they should not be included in the Technical Specifications.
Figure 1 shows a simplified P&ID of the interconnection between the fire water and RHR systems resulting from this modification.
Figure 1 identifies the containment isolation valves for that portion of the RHR system crosstied to the service water and fire water systems.
(RHR Loop A is shown, Loop B is similar.)
The integrity of the RHR system pressure boundary including 10-H0-511 is verified by hydrostatic pressure testing in accordance with the PNPS ISI program (PNPS Procedure 2.1.8.2, Safety Class 2 & 3 Hydrostatic Test Procedure). Gate valve 10-H0-511 will be checked for leakage during PNPS procedure 8.A.16, RHR System Integrity Surveillance to satisfy the requirements of NUREG-0578, Part 2.1.6.a.
2.
Section 3.7 - Backuo Nitrogen Sucolv System NRC Reauest The staff requests clarification regarding the function of one valve in the backup nitrogen supply system. As described in the enclosure, valve A0-4356 appears to be a containment isolation valve and, consequently, would be appropriate for inclusion in the Technical Specifications.
BECo Resoonse Check valve 31-CK-167 is the primary containment isolation valve in the containment instrument nitrogen supply line.
Check valve 31-CK-167 and its associated containment penetration are functional Class I components.
(Figure 2 illustrates the new Backup Nitrogen Supply System.) As defined in the original PNPS plant design (see FSAR Section 7.3.2), Class C containment isolation valves are in pipelines that penetrate the primary containment, but do not communicate directly with the reactor vessel, the primary containment free space, or the environt.
FSAR Section 5.2.3.5.1 further states that Class C lines require only one valve which closes
(
automatically by process action (i.e., reverse flow) or by remote manual l
operation from the control room. Check valve 31-CK-167 meets the definition for Class C isolation valves. Also, check valve 31-CK-167 is leak tested as a containment isolation valve to the requirements of 10CFR50 Appendix J.
The original design function of A0-4356 which is upstream of check valve 31-CK-167 was to provide remote manual isolation for an instrument air line rupture inside of containment.
This was when instrument air was the only pneumatic source to the containment. Subsequently, the design was modified and nitrogen was made the primary source with instrument air as backup.
The Backup Nitrogen Supply System virtually eliminates the concern for air intrusion since now both the primary and secondary pneumatic sources are nitrogen.
The instrument air line is now isolated by valve 31-H0-162 becoming a normally closed valve.
Inconsistencies between the Compressed Air P&ID M-220 Sh. I and the FSAR were identified and FSAR changes are being made to assure consistency between documents and accuracy of information.
Check valve 31-CK-167 is located outside containment. Check valve 31-CK-167 is a Class C containment isolation valve. A0-4356 is nQi a containment isolation valve.
Changing A0-4356 to a fail open valve is considered an acceptable design.
Check valve 31-CK-167 is leak tested to the requirements of 10CFR50 Appendix J.
The instrument piping is not in direct contact with containment atmosphere or the reactor building atmosphere and the direction of process flow is into the containment.
The containment penetration (X-22), piping and check valve 31-CK-167 within the functional Class I boundaries meet Class I design criteria including seismic qualification. A0-4356 can be closed by a remote manual control switch.
Because PNPS has an operating license application docket date of June 30, 1967, it is not subject to the requirements of 10CFR50 Appendix A, Criterion 57. However, a comparison of the PNPS design bases with the proposed General Design Criteria published by the AEC in the Federal Register dated July 11, 1967 was performed and documented in FSAR Appendix F.
This comparison concluded that PNPS met the proposed criteria at the time.
Therefore, a single Class I containment isolation check valve in a line that does not communicate directly with the reactor vessel or the containment free space meets the PNPS design bases and licensing commitments.
3.
Section 3.12 - Modification to Reactor Core Isolation Coolina System IV.Chinst NRC Reaugli The staff still has questions regarding the proposed modification to the reactor core isolation cooling (RCIC) system.
Prior to implementing this modification the staff requests that BECo conduct an assessment of hydrodynamic loads on the RCIC piping and supports, based on the proposed exhaust pressure of 46 psig, and make the results of that assessment available to the staff.
BECo Response RCIC operation at higher exhaust back pressure up to 46 psig is acceptable for the following reasons:
The starting transients and air clearing loads are low. The RCIC turbine has approximately a 10 second start up time.
A gradual start up over such a long time does not produce high air clearing loads or dynamic effects and will not change significantly with higher back pressure as concluded in the attached Teledyne calculation.
Flow rates through the RCIC exhault line are low.
Steam flow at 25 psig back pressure is 12.3 lbm/ftz/sec (i.e.16250 to 15,350 lbm/hr in an 8" line) and would not change appreciably as back pressure is increased to 46 psig.
Data from SRV testing indicated that flow rates in this range do not result in significant containment loads.
The pipe stresses and support / penetration loads for the Reactor Core Isolation Cooling (RCIC) exhaust piping have previously been evaluated in the Mark I Containment Long Term Program for the combined loads that occurred during the simultaneous application of Condensation Oscillation (CO) shell loading, C0 drag loads applied to the submerged piping, SSE loads, thermal loads, and weight loads (Teledyne Report TR-5310 Rev. 2).
This load combination controlled the pipe stresses in the Mark I Containment Long Term Program because the sinusoidal C0 forces occurred in a frequency range where the piping has high dynamic amplification.
This is a severe condition that bounds any forces related to continuous steam condensation at the low flow rates associated with RCIC operation.
The proposed back pressure setpoint of 46 psig will not result in excessive pressure stress on the RCIC turbine exhaust piping.
Original design conditions for the RCIC exhaust piping are 100 psig, 325'F.
Dynamic loading from stop/ start operation of the RCIC is not a concern because the RCIC will only cycle automatically between high and low RPV level and this is not a rapid transient. Calculations show that for the nominal RCIC makeup capacity of 400 gpm it would take about 36 minutes to go from the low water level for RCIC start to the reactor high water level for RCIC stop. Water clearing times from the exhaust line have been calculated to be approximately 4.4 minutes with the existing 1" vacuum breaker and 1" drain line.
The water clearing time is based on the entire volume of the exhaust line from the turbine to the torus.
Reflood/ restart water hammer potential does not change significantly with the higher back pressure of 46 psig vs the previous 25 psig because reflood to the first check valve off the torus occurs in both cases.
Since there have not been any historical problems with RCIC exhaust piping and supports, reflood/ restart water hammer is not a problem for the RCIC exhaust piping.
B ACKUP NITROGEN SUPPLY SY, STEM
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- 1. F. Kreith, PRINCIPLCS (5 HEAT TRANSFER, Scranton Pa., International Textbook Co., 1965, 2nd ed.
- 2. Gordon J Van Wylen, THERMODYNAMICS, New York, John Wiley and Sons Inc.,
1959.
- 3. F.W. Sears and M.W.Zemansk, UNIVERSITY PHYSICS, Reading Ma., Addison-Wesley Publishing Co.,1955, 2nd ed., 4th printing.
- 4. J.E. Torbeck, ELIMINATION OF LIMIT ON BWR SUPPRESSION POOL TEMPERATURE FOR SRV DISCHARGE WITH QUENCHERS, General Electric, NEDO 30832 class 1, December 1984.
- 5. Boston Edisen's Pilgrim Nuclear Power Station, Final Safety Analysis Report.
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