ML18081B275
| ML18081B275 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/28/1980 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Parr O Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8004010438 | |
| Download: ML18081B275 (11) | |
Text
{{#Wiki_filter:Public Service Electric and Gas Company 60 Park Place Newark, N.J. 07101 Phone 201/430-7000 March 28, 1980 Director of Nuclear Reactor Regulation U. S. NuclearRegulatory Commission Washington, D.C. 20555 Attention: Mr. Olan D. Parr, Chief Light Water Reactors Branch 3 Division of Project Management Gentlemen: TMI-2 LESSONS* LEARNED NO, 2 UNIT .SALEM NUCLEAR GENERATING STATION
- DOCKET NO. 50-311 PSE&G hereby transmits updated information concerning imple-mentation of Lessons Learned requirements.
The information enclosed is in the form of new and replacement pages to be inserted in the enclosure to our submittal of January 4, 1980. Additionally, we wish to aovise you that the On-Site Tech-nical Support Center (TSC} is operational, with all equipment and telephone communications described in our January 4, 1980 submittal to meet short term (1980) requirements installed. The upgrading effort, including design and procurement of equipment, to meet the long term (1981) requirements.for the TSC is proceeding on schedule. Should you have any questions, do not hesitate to contact us. ~~7il' R. L. Mittl General Manager - Licensing and Environment Engineering and Construction ~f <?:,- r/JcP ~~LO~ 5 /I f)-{J. CSOS . I I
~ -*-.. **--**"'..... -...... -** Performance Testing f.BWR and PWR Relief and S~ty Valves (Section 2.1.2) NRC Position Pressurized water reactor and boiling water reactor li-cense~s and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves under ex-pected operating conditions for design basis transients and accidents. The licensees and applicants shall determine the expected valve operating conditions through the use of anal-yses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Revision 2. The single failures applied to these analyses-shall be chosen so that the dynamic ~orces on the safety and relief valves are maxi-mized. Test pressures shall be the h~ghest predicted by conventional safety analysis procedures. Reactor coolant system relief and safety valve qualification shall include qualification of associated control circuitry piping and supports as well as the valves themselves.
Response
By letter dated December 17, 1979, the EPRI Safety and Analysis Task Force submitted its "Program Plan for the Performance VP.rif ication of PWR Safety/Relief Valves and Systems," dated December 13, 1979, to the NRC for review. PSE&G considers the program to be responsive to the NRC's position. The EPRI program plan provides for completion of the essential portions of the test program by July, 1981. PSE&G is actively participating in the EPRI program, which is applicable to the Salem design. M P79 54 01/7 Salem 1 & 2 . -*-*.......... *--~-*-*-----.......... -------r-*--*-..,.. --*--.........._-......
Direct Indication of !,er-Operated Relief Valve.a Safety Valve Position for PWRs and BWRs (Section 2.1.3.a) NRC Position Reaclor system relief and safety valves shall be provided with a positiv~ indication in the control r6om derived from a reliable valve position detection device or a reliable indication of flow in the discharge pipe.
Response
Each PORV has been equipped with a limit switch to provide an alarm in the Control Room if the PORv* is not fully closed. These switches are seismically and environmentally qualified. To provide positive indication of safety valve position, a limit switch will be mounted in each safety valve bonnet which will actuate a Control Room alarm if the valve is not fully closed. This modification has been completed. Although the switches installed on the safety valves are qualified for both seismic and environmental conditions, an improved switch will be installed by June 1, 19~0. The improved switch will be capable of indicating open, clos~d and an intermediate position. Both of the above schemes utilize a single switch on each valve and powered from a vital bus. As discussed in the response to NRC Bulletin 79-06A, several reliable backup methods are available to detect an open valve which are \\ M'P79 54 01/8 Salem l & 2 MAR :~ ~ 1980 I I I
addressed in Emergency Procedure EI 4.24, "Malfunction of Pressurizer Relief Valve."
- l. Pressurizer pressure
- 2. Valve discharge piping temperature
- 3. PRT level, pressure and temperature
- 4. PORV open/close indica t ior:.. ir;'... 'conj unction with PORV block valve position ind~cation S. Control Room alarms for all of the above indicators.
I M P79 54 01/9 -Sa-Salem 1 & 2 MAR ~ b 1980
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Steam generator pressure constant or decreasing at a rate equivalent to the rate of decrease of RCS tempera-tures whil~ maintaining steam generator level with continuous auxiliary feedwater. A further guide for recognition of inadequate core cooling is the recent addition of a computer/CRT display for sub-cooling. Th~ significant parameters which ~re continuously displayed are reactor coolant differential pressure (P act-ual - P saturated) and differential temperature (T saturated - Tactual). Alarms are set for pressure differential less than 200 psi and temperature differential less than so°F. The computer program is predicated on the hottest in-core thermocouple reading. The CRT matrix of in-core thermo-couples will display the location of the hottest in-core thermo~ouple. A dedicated recorder has been provided at the computer console to record differential pressure and temperature. Additional information is provided in Table 2.1.3.b-l. Emergency Procedures, EI 4.4, "LOCA", and EI 4.6, "loss of Secondary Coolant," have been revised to address the use of this computer program to monitor the margin of subcooling in the Reactor Coolant System. PSE&G is a member of the Westinghouse Operating Plant Owners' Group. Westinghouse, under the direction of the Westinghouse Owners Group, is performing\\further analyses to aid in selection of.more direct indicators of inadequate M P79 54 01/11 Salem 1 & 2 MAR 2 b 1960
core cooling, and to serve as a basis for augmented emer-gency procedures. A preliminary report on inadequate core cooling was sub-mitted to the NRC on October 30, 1979 by the Owners' Group. The Salem 2 Emergency Procedures will be revised*on an interim basis to specify precautions and op~rator actions to recover from a condition in which the core has experienced inadequate cooling. These interim revisions will be avail-able for NRC review prior to power testing. The station procedures will be further updated after.completion of the final Owners' Group report. It is our intent to install a device to indicate reactor vessel water level on Salem 2. This.device will be similar to the proposed Westinghouse design for VEPCO's North Anna plant. Installation of this device will be accomplished during the first refueling outage subject to equipment availability and acceptability of the design to provide an unambiguous indication of inadequate core cooling. M P79 54 01/12 -a-Salem 1 & 2 MAR 2 8 1980
I Auxiliary Feedwater Flow Indication to Steam Generators for PWRs (Section 2.1.7.b) NRC Position Consistent with satisfying the requirements set forth in GDC 13 to provide the capability in the control room to ascertain the actual performance of the AFWS when it is called to perform its intended function, the following requirements shall be implemented:
- 1.
Safety-grade indication of auxiliary feedwater flow to each steam generator shall be provided in the contrql room.
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The auxiliary feedwater flow instrument channels shall be powered from the emergency buses consistent with satisfying the emergency power diversity requirements of the auxiliary feedwater system set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9.
Response
Safety-grade indication of auxiliary feedwater flow to each steam generator is provided in the Control Room. These indicating channels are designed to the same criteria as the protection system indicators. One testable flow instrument with an accuracy on the order of +/- 2%, is provided for each steam generator. In addition, three level instruments are provided for each steam generatot. The instruments are all powered from the vital buses, seismically qualified with environmental qualification for the level instruments which are located inside the containment. M P79 54 01/28 Salem l & 2 M11R ~ o 1980 ... ~- *.... -*... ~~- -~~~~-
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Response
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The plant vent gaseous monitors have the following detection range capabilities: Unit 1: Sxlo-6 to Sx10-l uCi/cc Xe-133 Unit 2: lx10-6 to lxlo2 uCi/cc Xe-133 Design changes have previously been initiated, and equipment purchased to upgrade the detection range capability of Unit 1 to that of Unit 2. Further design modifications are presently being evaluated to provide the gaseous monitors with a detection range capability of lo-7 to 105 uCi/cc Xe - 133. The modified system will utilize multiple monitors with the required overlap to meet the above criteria. An alternate consideration is the use of a detector with a range of 104 uCi/cc if the containment exhaust is diluted by at least a factor of 10. These modifications will be completed by January 1, 1981. In the interim, a single thin wall GM tube will be positioned on top of the auxiliary building approximately 150 feet away from.each unit's vent. In the event of a major discharge of activity, the count rates displayed by \\ the existing radiation monitors will indicate which unit is releasing tha 0 high activity. If the radiation field M P79 54 01/33 Salem 1 & 2 MAR 2 8 1980
J. I 1 i, . j produced is high enough, then radiation levels at the GM tube will be measurable. Curves will be used to relate the direct radiation dose rates to the activity dis-charged. Dose rates in mR/hr are converted into micro-curies per cc using standard volume source calculations. The properties of the instrument are provided below: Instrument - Halogen q~encb~~ GM tube. Range - 0.1 mR/hr to 104 rnR/hr. Energy Dependency - +/- 20% from 100 keV to 2.5 MeV. Calibration Frequency - Every 18 months. Location - 150' away from each unit vent with the GM tube located in a shielded lead cavei Background Correction - Existing equipment is capable of adequately monitoring low range requirements. Cosmic ray background radiation is small enough compared to the dose rates being measured not to be a significant problem. Two inches of lead shielding around the detector will minimize detector response to other background radiation sources. Access To Readings - Readings will be availoble contin-uously and will be recorded in the Unit 1 Control Room. \\ Power Source - Vital power will be provided. M P79 54 01/33.l -25a-Salem l & 2 HAR ~ 8 1980
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- 1 Calculational Methods and Information Dissemination - The calculational method used for converting instrument r~ad-ings to release rates will be provided in the instruc-tions for interpreting monitor response.
Calculations using the ORIGEN code are used to provide spectrum dis-tribution and average energy values. Analysis indicates that within a few hours after reactor shut-down, Xenon 133 becomes the predominant isotope. The channel read-ings will be recorded and the information will be posted on an emergency status board to be located in the on-site technical support center. Since the channel measures the production of ion pairs produced in the tube, calibration is done by determining the GM tube plateau. This interim modific~tion will b~ operational by April 4, 19_80.
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The Salem design provides for iodine sampling by adsorption on charcoal cartridges, followed by onsite laboratory analysis.
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The containment high range monitors presently have the following maximum detection ranges: Unit 1 : l 0 4 R/h r
- unit 2:
io7 R/hr. M P79 54 01/33.2 . -25b-Salem 1 & 2 MAR 2 8 1980
1 / utility has develope.plant specific proceduresttnd trained their personnel on the new procedures. Revised procedures and training are in place in accordance with the requirement
- in Enclosure 6 to Mr. Eisenhut's letter of September 13, 1979, and Enslosure 2 to Mr. Denton's letter of October 30, 1979.
The work required to address the other two areas--inadequate core cooling and other transient and accident scenarios--has been performe.d in conjunction with schedules and require-ments established by the Bulletins and Orders Task Force. Analysis related to the definition of inadequate core cooling and guidelines for recognizing the symptoms of inadequate core cooling based on existing plant instrumentation and for restoring co~e cooling following a small break LOCA were submitted on October 31, 1979. This analysis is a less detailed analysis than was originally proposed, and will be followed up with a more extensive and detailed analysis which will be available during the' first quarter of 1980. The Salem 2 Emergency Proc~dures will be revised on an interim basis as described in the response to Section 2.1.3.b. With respect to other transient accidents contained in the Salem FSAR, the Westinghouse Owners' Group has performed an e~aluation of the actions which occur during an event by ~ constructing sequence of event trees for each of the non-LOCA and LOCA transients. From these event trees a list MP79 54 01/40 . Jla Salem l & 2 Ml\\R ~ 8 1980}}