ML20198R832
| ML20198R832 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 05/30/1986 |
| From: | Capstick R VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | Rooney V Office of Nuclear Reactor Regulation |
| References | |
| FVY-86-51, NUDOCS 8606100230 | |
| Download: ML20198R832 (6) | |
Text
,
i VdRMONT YANKEE NUCLEAR POWER CORPORATION R D 5. Box 169, Ferry Road, Brattleboro, VT 05301 nepty To.
. p ENGINEERING OFFICE May 30, 1986 1671 WORCESTER ROAD FVY 86-51 FRAMINGHAM, MASSACHUSETTS 01701 TELEPHONE 617-872-8100 United States Nuclear Regulatory Commission Washington, DC 20555 Attention:
Office of Nuclear Reactor Regulation Mr. V. L. Rooney, Senior Project Manager BWR Project Directorate No. 2 Division of BWR Licensing
References:
(a) License No. DPR-28 (Docket No. 50-271)
(b) Letter, VYNRC to USNRC, WVY 75-95, dated September 9, 1975 (c) Letter, VYNRC to USNRC, WVY 80-132, dated September 19, 1980 (d) Letter, VYNRC to USNRC, FVY 81-148, dated October 27, 1981 (e) Letter, VYNRC to USNRC, FVY 82-30, dated March 26, 1982 (f) Letter, USNRC to VYNPC, NVY 83-192, dated August 19, 1983 (g) Letter, VYNRC to USNRC, FVY 84-76, dated June 26, 1984 (h) Letter, USNRC to VYNRC, NVY 86-29, dated February 14, 1986
Subject:
Response to NRC Request for Additional Information - Appendix J Technical Specifications
Dear Sir:
By letter dated February 14, 1986 [ Reference (h)], you requested additional information in order to complete your review of Vermont Yankee's proposed Appendix J Technical Specifications [ Reference (g)).
In accordance with your request, enclosed please find responses to the five questions. Due to the impact on resource availability associated with the extended 1985/1986 pipe replacement outage, we are unabic to provide a complete response to Question No. 1 at this time.
In order to allow your review of our proposed Appendix J Technical Specifications to continue, we are providing this partial response to your request. We propose to provide a full response to the remaining infornation requested within 90 days of startup from this outage.
We trust that this information is sufficient to allow your review to continue; however, should you have any questions or problems with the proposed schedule for providing the remaining informatien, please do not hesitate to contact this office.
Very truly yours, 8606100230 860530 DR ADOCK 05000271 VERMONT YANKEE N C EAR POWER CORPORATION PDR f
R. W. Capstick Licencing Engineer hk RWC/no i
ENCLOSURE Response to NRC Request for Additional Information Appendix J Technical Specifications Question No. 1 The following penetrations have been listed in the Vermont Yankee Local Test program as having met the requirements of the water-seal as discussed above and in Appendix J ;and consequently no air-leakage testing is proposed.
For each of these systems state how a water seal is provided under the post accident DBA condition which include loss of off-site power and worst case single active failure. Include or reference drawings or sketches showing system piping and the pumps involved. Note that no credit may be given to water legs provided by water in the reactor vessel. A crossover between redundant trains of an ECCS system may be taken into account if a procedure exists to provide the water leg to the isolation valves in the event it would not otherwise be available:
Penetration No.
System X-12 RHR Shutdown Cooling Supply X-13A/B LPCI Injection X-14 RWCU Suction X-16A/B Core Spray X-17 RHR Head Spray X-42 Standby Liquid Control
Response
Penetration No.
Response
X-12 Later X-14 Later X-16A/B Later X-17 The RHR head spray line is essentially a BLANK FLANGED SPARE j
penetration with additional conservative margin in that the I
outside containment portion of its piping is connected to the RHR and l
condensate make-up system instead of open to the atmosphere.
SPARE PENETRATIONS ARE TESTED as part of the Type "A" test.
l l
X-13A/B The LPCI (RHR) discharge piping discharges into the reactor below the 2/3 core height of water so a seal water supply would be available l
even if the LPCI System was not i
running. However, the RHR (LPCI) pumps have redundant and diverse power supplied by the diesels. The post-LOCA seal water supply is, therefore, supplied by running RHR (LPCI) pumps.
X-42 The Standby Liquid Control (SLC)
System has isolation valves that are completely sealed and can only be opened by an explosive charge. This line terminates below the shroud several feet below 2/3 core water level (FSAR Sections 3.3.4.9 and 3.8.3) and thus a long term water seal is provided. The SLC System does not operate during a LOCA so the explosive valves will not be opened.
We have taken exception to the statement within your letter which states,
" Note that no credit may be given to water legs provided by water in the reactor vessel" because LOCA analysis for BWRs generally shows water at or above the top of active fuel and the VY FSAR conservatively shows at least 2/3 core height of water remaining following a LOCA (reference FSAR Figures 3.3-1, 3.3-2, 3.3-7, and 4.3-5).
Therefore, we reviewed the questions using the FSAR minimum water level. Lines terminating below this water level are considered to have a 30-day water supply since the purposes of the Emergency Core Cooling Systems (ECCS) is to maintain the core in a flooded condition well above this level.
The VY Low Pressure Core Injection System [ Residual Heat Removal (RHR)] has redundant pumps in each of two trains and the redundant pumps have diverse power supplied from each diesel. Therefore, both trains of LPCI (RHR) will be pressurized and supplying water even if a diesel falls to start.
Question No. 2 The following piping lines appear to terminate below tig minimum drawdown level of the suppression pool. As mentioned above, staff interpretation of GDC 54 requires a water test on these valves.
Indicate whether the valves will be water tested and if not justify why a water test should not be
.. =, -~
performed. Note that an exemption from the regulations must be requested and justified to eliminate leakage testing of these valves.
Penetration No.
System X-210A/B RHR, CS Return X-212 RCIC Turbine Exhaust X-221 HPCI Turbine Exhaust X-222 HPCI Drain Pot X-223 RCIC Drain Pot X-224 RHR Suction X-225 HPCI Suction X-226A/B Core Spray Suction X-227 RCIC Suction
Response
Because Appendix A General Design Criteria (GDC) was issued in 1971 well after the VY 1968 construction permit, VY does not necessarily meet the staff's present interpretation of GDC-54.
We have tried to find a basis for the staff's interpretation requiring an individual water test of these valves, but cannot. We do believe VY complies with GDC-54 by using the acceptance limits of the periodic containment pressurization test to determine if overall leakage is within acceptable limits. Because of the originally accepted design features for VY, specific valve testing to determine quantitative leakage rates cannot be done without extensive plant modifications such as additional manual isolation valves, vent valves, and test valves. However, these same original design features can provide qualitative evidence as to which valve or valves are providing unacceptable containment pressurization leakage rates. We believe this meets CDC-54 and that no exemption is required since quantitative setpoints for isolation valve testing (water leakage test) are not specifically required.
Also, as stated by your letter, a water leakage test is not required by Appendix J for these penetrations.
Question No. 3 Discuss the test provisions for the TIP explosive shear valve and the ball valves, Penetration X-35C, D, E.
Also, indicate the proposed testing provisions on the check and solenoid valve of the TIP Air Purge System, Penetration X-35A.
Response
The TIP explosive valves are verified operable by electrical checks and periodic full explosive tests, leak testing in accordance with Appendix "J"
requirements cannot be done.
Penetrations X-35 A, C, D, and E listed isolation valves are tested by disconnecting the purge supply ball valve at each indexer within containment and pressurizing against the purge and TIP line isolation valves in the expected isolation direction.
Question No. 4 For the CAD System indicate the valves that will be Type C air tested in Penetrations X-50A, B, C-205.
Response
Penetration No.
System X-50A, B, C These three penetration lines have a common connection prior to the isolation valves. These valves, VG-26, and VG-23, are Type C air tested in the direction required to isolate.
X-205 The line connected to this penetration has many branches with the primary isolation valves on the branches. All these valves, as listed in the Technical Specification submittal [see Reference (g)], are Type C tested.
Valves 16-19-11A and B should be listed as " tested in the reverse direction."
Question No. 5 The following system isolation valves appear to meet the definition of Section II.A of Appendix J requiring Type C leak testing.
Indicate the extent of testing proposed for the isolation valves in these lines and provide your basis for any valves for which test Type C testing is not proposed.
Penetration No.
System X-23, 24 Inlet RBCCW X-39A/B Drywell spray l
Response
Penetration No.
System X-39A/B The RHR drywell spray line isolation valves are supplied with a water seal from the RHR (LPCI) pumps when drywell spray is not used. The RHR pumps are redundant and have diverse power supplies.
X-23, 24 Type "C" testing is not proposed and our exception is based on the following.
Should the piping outside the containment isolation boundaries fail in a seismic event, the closed piping loop inside the drywell is seismically qualified and designed to remain intact. This piping loop does not communicate with the reactor coolant pressure boundary or the containment atmosphere. The leaktight closed piping loop inside containment provides an equivalent barrier to the release of radioactive material as a water seal penetration which has previously been accepted by the NRC as an acceptable basis for not performing Type C testing and still meeting the intent of Appendix J.
In summary, the RBCCW System is Safety Class 3 and designed to operate post-LOCA provided a coincident seismic event is not also assumed. The pumps and other essential electrical equipment are powered from emergency power sources and environmentally qualified. For the LOCA without scismic cvent, the RBCCW System will be operating and pressurized post-LOCA which is consistent with the present Leak Rate Testing program basis (Note 3) for the RBCCW penetrations. For the seismic event, althcugh the RBCCW System may not be operable (without rapairs), the portions of RBCCW piping essential to containment isolation are designed to remain intact thereby insuring containment integrity.
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