ML18094A678
| ML18094A678 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 09/05/1989 |
| From: | Labruna S Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 NLR-N89180, NUDOCS 8909130183 | |
| Download: ML18094A678 (14) | |
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Public Service Electric and Gas Company Stanley LaBruna Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-4800 Vice President - Nuclear Operations Document Control Desk SEP o 5 1989 NLR -
N89180 U.S. Nuclear Regulatory Commission Washington, DC 20555 Gentlemen:.
RESPONSE TO NOTICE OF DEVIATION; REGULATORY GUIDE 1.97 SALEM UNIT NO. 1 DOCKET NO. 50-272 Public Service Electric and Gas Company (PSE&G) acknowledges
.receipt of the Notice of Deviation contained in Appendix A of NRC Combined Inspection Report 50-272/89-13 and 50-311/89-12, dated August 4, 1989.
Our detailed response to the individual deviations is provided as Attachment 1 to this transmittal. addresses Unresolved Item 272/89-13-05 as requested in your cover letter.
If there are any questions regarding the attached information, please feel free to contact us.
Sincerely, Attachments
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Document Control Desk NLR-N89180 c
Mr. J. c. Stone Licensing Project Manager Ms. K. Halvey Gibson Senior Resident Inspector 2
Mr. w. T. Russell, Administrator Region I Mr. Kent Tosch, Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625 SEP o 5 1989
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ATTACHMENT 1 SALEM GENERATING STATION UNIT NOS. 1 AND 2 RESPONSE TO NRC NOTICE OF DEVIATION DATED AUGUST 4, 1989 This response has been prepared for Deviation Nos. 1 through_ 4 of Inspection Report No. 50-272/89-13 as directed in Appendix A, Notice of Deviation, Reference No. 1.
DEVIATION ITEM NO. 50-272/89-13-01 REACTOR COOLANT SYSTEM HOT LEG WATER TEMPERATURE The Licensee determined the reactor coolant system wide range hot leg water temperature to be a type A variable.
One instrument channel was provided for each of the four reactor coolant loops to cover the water temperature range.
Each instrument channel shares a recorder with the associated cold leg loop water temperature.
The four two-pen recorders are located on the main control board in the control room.
All instruments in the four instrument channels are on the environmental and seismic qualification master lists.
The instrument channels were found to be in calibration as evidenced by the plant calibration records inspected.
Since the four instrument channels are supplied by four redundant sources of power, they were believed to be redundant to each other.
Further investigation by the inspector revealed that the four instrument channels do not have a corresponding redundant instrument channel, as specified by R.G. 1.97 for type A variables.
The licensee noted that each of the four reactor coolant loops contains three dual RTDs located downstream of the ones used for post-accident monitoring purposes.
These RTDs are used in the protection circuits and, after isolation, for narrow range indication of loop average and differential temperatures.
since these instrument channels use a method and location for measuring loop temperature which is different from those used for post-accident monitoring they may not provide the redundancy recommended by R.G.
1.97.
This item is a deviation from the guidance in R.G.
1.97 (50-272/89-13-01; 50-311/89-12-01)
- Page 1 of 9
RESPONSE
PSE&G requests that the NRC reevaluate the categorization of this item.
Additional information was identified subsequent to the inspection which leads us to conclude that the present design is adequate.
The basis for this request is discussed in the following paragraphs.
on November 18, 1982 the NRC solicited comments (Reference No. 8) from the Westinghouse Owners Group (WOG) concerning three (3) alternatives which could be considered to achieve reliable temperature indication relative to Reg. Guide 1.97.
Alternative No. 3 discussed the powering of single point hot leg temperature sensors such that all sensors would not be dependent upon a single power source.
The NRC also stated that "redundancy would be met on a system basis rather than on a loop basis" for Alternative No. 3.
Subsequent to the issuance of the NRC letter discussed above, PSE&G conducted a review of the RCS T-Hot instrument arrangement at Salem Units 1 and 2.
on February 16, 1983 PSE&G issued a memorandum (Reference No. 7) which stated that "The design at Salem provides single point hot *** leg temperature measurements but the hot... leg temperature sensors (RTDs).for each loop have a different power supply which satisfies redundancy on a system basis." This memorandum concluded that the Salem design was consistent with Alternative 3.
on June 14, 1983 the WOG provided comments to the NRC via NSID/WOG-108 and OG-94, Revised (Reference Nos. 5 and 6),
relative to the three alternatives for reliable temperature indication identified by the NRC.
It was concluded that "The present reactor coolant system (RCS) temperature indication is reliable and adequate for any accident condition." This conclusion was based on the RTDs being powered from a single bus with three diverse power sources: offsite A.C., onsite A.C. (emergency diesel generators) and onsite safety grade o.c. power.
It is PSE&G's position that system basis redundancy is sufficient to meet the requirements of Reg. Guide 1.97.
At Salem Generating Station Units 1 and 2, each hot leg temperature instrument loop is powered from a separate vital instrument bus.
Each instrument bus is provided with three diverse power sources: offsite A.C., onsite A.C. (emergency diesel generators) and onsite safety grade D.C. power.
This configuration is consistent with Alternative 3 of Reference 8 and exceeds the basis for the WOG conclusion.
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DEVIATION ITEM NOS. 50-272/89-13-02 EFFLIJEHT RADIOACTIVITY -
NOBLE GAS EXHAUST The licensee classifies the effluent radioactivity monitor for noble gases effluents from the condenser air removal system exhaust as a Type A variable.
For this variable the licensee provided instrumentation which did not meet the R.G. 1.97 recommendations.
Therefore, in accordance with the position taken in the SER, the licensee assigned to this function the R41C instrument loop which was classified as a Type C variable.
This instrument loop is category 1, environmentally and seismically qualified, and provides post-accident monitoring via a digital indicator in the control room.
The calibration records were evaluated and found to be acceptable.
The licensee, however did not provide appropriate redundant instrumentation, as recommended by R.G. 1.97 for Type A variables.
The licensee indicated that other safety related instrumentation existed to provide the operator with a redundant instrument loop in the event of failure of the R41C loop.
However, the instruments suggested by the licensee monitors radioactive noble gases in a range not monitored by the R41C loop.
These instruments, therefore, are not acceptable redundant instruments.
This issue is a deviation from the R.G. 1.97 recommendations on redundancy.
RESPONSE
PSE&G requests that the NRC reevaluate the categorization of this item.
Additional information was identified subsequent to the inspection which leads us to conclude that the present design is adequate.
The basis for this request is discussed in the following paragraphs.
A review of the original Type A Variable analysis documentation indicates that the Rl5 monitor (Condenser Air Removal Monitor) was added to the Type A list for the purpose of providing a redundant indication of Steam Generator Tube Rupture (SGTR) and not for monitoring plant radioactive effluents.
The monitoring of plant effluents is classified as Type C or E and Category 2 or 3 in accordance with R.G. 1.97.
Additionally, R.G. 1.97 allows the monitoring of a common release point in lieu of individual streams.
The condenser air removal system discharges into the common plant vent.
The plant vent is monitored by the R41 (low range) and R45 (high range) monitors.
This variable (No. 18) is classified Type c, Category 2 and as such, does not require redundant indication.
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Based on the above, it is PSE&G's position that the compliance exists for the Noble Gas Effluent Monitoring variable.
The final disposition of the R15 Condenser Air Removal monitors is discussed in the response to Deviation 272/89-13-04.
DEVIATION ITEM NO. 50-272/89-13-03 AUXILIARY FEEDWATER FLOW The licensee considers the auxiliary feedwater flow to be a type A variable.
As such, the instruments involved were classified as Class lE and environmentally and seismically qualified, as indicated by the licensee's associated master lists.
The variable is monitored by four instrument loops, one for each feedwater train and includes indicating meters on the benchboard in the control room and recording via the safety-related SPDS computer.
However, contrary to R.G. 1.97, the variable is not monitored by redundant post-accident instrumentation.
This item is a deviation from the Regulatory Guide (50-272/89-13-03; 50-311/89-12-03).
The existing instrument loops are properly calibrated as evidenced by the records provided by the licensee.
RESPONSE
PSE&G requests that the NRC reevaluate the categorization of this item.
Additional information was identified subsequent to the inspection which leads us to conclude that the present design is adequate.
The basis for this request is discussed in the following paragraphs.
Auxiliary feedwater flow indication for SGS Unit Nos. 1 and 2 is provided by a single flow element located in the individual feed line to each of the four steam generators.
The four AFW flow channels satisfy Reg. Guide 1.97 Category 1 design requirements with the exception that there is no redundant AFW flow indication provided.
However, it is PSE&G's position that redundancy for each of the four AFW flow channels is provided by the Type B, Category 1 wide range level indication channel provided for each of the four steam generators.
In SER Supplement No. 4, April 1980 (Reference No. 12),
the NRC stated that "Safety grade indication of auxiliary f eedwater flow to each steam generator shall be provided in the control room *..* the flow indication channels should by themselves satisfy the single failure criterion for each steam generator." " As a fall-back position, one auxiliary feedwater flow channel may be backed up by a steam generator level channel **** The applicants have noted that the direct flow indication arrangement provided is backed by safety grade steam generator water level indication.
Taken together then, the combined (direct and Page 4 of 9
indirect) AFW flow indication capability does satisfy the single failure criterion."
Additionally, in SERs provided for the AFW System at SGS Unit Nos. 1 and 2, dated June 16 and June 22, 1981, the NRC determined that, 11 *** steam generator level indication and flow measurement were to be used to assist the operator in maintaining the required steam generator level during AFW system operation."
11 *** in the long term, the overall flowrate system for Westinghouse plants should include at least one auxiliary feedwater flowrate indicator for each steam generator."
11 *** The operator relies on steam generator level instrumentation, in addition to auxiliary feedwater flow indication, to determine AFW system performance."
11 *** In order to adequately determine from the control room the performance of the AFWS, steam generator level instrumentation is used, in addition to flow indication."
It is PSE&Gs position that the existing indication of AFW flow at SGS Unit Nos. 1 and 2 satisfies the redundancy requirement of Reg. Guide 1.97 as applied to Category 1 instrumentation.
DEVIATION ITEM NOS. 50-272/89-13-04 STEAM GENERATOR RADIATION (50-272/89-13-04)
The licensee classifies the steam generator radiation to be a Type A variable.
The post-accident monitoring of this variable was originally provided by non safety-related instrumentation.
In accordance with the position taken in the SER, the licensee added four safety-related, environmentally and seismically qualified instruments loops with indication on a back panel and recording on a front panel in the control room.
To comply with the redundancy recommendations of R.G. 1.97, a second recorder was provided.
This recorder is supplied with a redundant source of power and receives electrically isolated signals from the same four instrument loops.
In evaluating the design used to measure this variable, the inspector identified several potential single failure modes which could impair both redundant channels and, thus, prevent the post-accident availability of this variable to the operator.
An example of failures affecting both channels is the loss of control power to the sample inlet solenoid valves.
In this case, the valves fail close with consequent loss of indication at both recorders.
The existing instrument loops were calibrated in accordance with the applicable technical specifications.
This item is considered a deviation from R.G. 1.97.
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RESPONSE
PSE&G does not dispute this deviation.
The following paragraphs discuss the licensing basis and proposed resolution for the deviation.
The requirement for the selection of Type A variables is driven by the need for the operator to manually respond to plant centered events in the absence of any other automatic function.
In the case of a Steam Generator Tube Rupture (SGTR) accident, the Emergency Operating Procedures require the operator to evaluate the condition of the plant and ~ake specific manual actions to isolate the affected steam generator and place the plant in a safe condition.
In its original analysis for R.G. 1.97, PSE&G identified the Steam Generator Blowdown Monitors (Rl9A through D) as the principle indication for this event.
Because of the lack of redundancy for the Rl9 monitors, PSE&G identified the Condenser Air Removal Monitor (Rl5) as providing a redundant indication.
Both of these monitors were classified as Type A, Category 1 in accordance with R.G. 1.97.
on September 21, 1983, PSE&G responded (Reference No. 4) to the NRC's request for additional information (Reference No.
- 13) relative to implementation of R.G. 1.97 requirements at Salem Units 1 and 2. to that response igentified deviations from the requirements of R.G. 1.97 relative to the seismic qualification of these monitors. also referenced the use of the High Range Main Steam Line Monitors (R46A through E) as providing a backup indication to the Rl5 and Rl9 monitors.
At that time, the R46 monitors were identified as meeting the requirements of R.G. 1.97.
As noted in the inspection report, the present design of the R46 monitors is subject to various single failures which is itself a deviation from the requirements of R.G. 1.97.
The root cause of this deviation was a failure to properly implement the design requirements of R.G. 1.97.
CORRECTIVE ACTIONS PSE&G proposes to substitute the Steam Line Radiation Monitoring Variable in the place of the Condenser Air Removal and Steam Generator Blowdown Monitoring variables on the Type A variable list.
The R46 monitors are seismically qualified, located in a mild environment, and are included in the plant Technical Specifications.
The range of these monitors is sufficient to allow the operator to distinguish between a tube leak and a SGTR event.
The Rl5 and Rl9 monitors will be reclassified as Type E, Category 2 in accordance with R.G.
1.97.
A complete evaluation of this change will be documented as required by 10 CFR 50.59.
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In order to meet the design requirements for Type A, Category 1 variables, modifications will be required.
These modifications consist of providing additional hardware to satisfy single failure criteria.
PSE&G is evaluating the extent of these modifications and will identify the full scope and schedule by December 31, 1989. It is expected that required modifications would be accomplished as soon as possible, however; in no case will the work be deferred beyond the tenth and sixth refueling outages for Salem Units 1 and 2, respectively.
These outages are presently scheduled for the Fall of 1991 and Spring of 1992.
To address the root cause identified above, PSE&G has established a project team tasked with evaluation and documentation of the as-built compliance level of all R.G. 1.97 instrumentation loops.
The initial phase of the project is underway and consists of a review and consolidation of licensing/design basis information and the preparation of detailed technical standards to be used during the review and for future maintenance/modification activities.
The present schedule for the project calls for completion of the compliance review by November 1990.
The project plan is available at the site for NRC review
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REFERENCES:
- 1)
Correspondence, NRC to PSE&G, Notice of Deviation*and Combined Inspection Report Nos.
50-272/89-13 and 50-311/89-12, August 4, 1989.
- 2)
Correspondence, NRC to PSE&G, Safety Evaluation Report, Salem Nuclear Generating station, Unit Nos. 1 and 2, Docket Nos. 50-272/311, Conformance to Regulatory Guide-
- 1. 97, June 17, 1985.
- 3)
Correspondence, NRC to PSE&G, INEL Interim Report, Conformance to Regulatory Guide 1.97, Salem Nuclear Generating Station, Unit Nos. 1 and 2 (published March 1983), May 4, 1984.
- 4)
Correspondence, PSE&G to NRC, Compliance with Regulatory Guide 1.97, NRC Request for Additional Information, No. 1 and 2 Units, Salem Generating Station, Docket Nos. 50-272 and 50-311, September 21, 1983.
- 5)
Correspondence, Westinghouse Electric to NRC, NSID/WOG-108, Regulatory Guide 1.97, Rev. 2, Requirements for Reactor Coolant Temperature Indication, June 14, 1983.
- 6)
Correspondence, WOG to NRC, OG-94 (Revised),
Regulatory Guide 1.97. Rev. 2, Requirements for Reactor Coolant Temperature Indication, June 14, 1983.
- 7)
PSE&G Memorandum, Regulatory Guide 1.97, Rev.
2, Requirements for Reactor Coolant Temperature Indication, February 16, 1983.,
- 8)
Correspondence, NRC to WOG, LSOS-82-11-072, Regulatory Guide 1.97, Rev. 2, Requirements for Reactor Coolant Temperature Indication, November 18, 1982.
- 9)
Regulatory Guide 1.97, Revision 3, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, May 1983.
Page 8 of 9
- 10)
- 11)
- 12)
, 13)
Correspondence, NRC to PSE&G, Safety Evaluation, Salem Unit 1 - Auxiliary Feedwater Automatic Initiation and Flow Indication, Action Plan Item II.E.1.2, June 16, 1981.
Correspondence, NRC to PSE&G, Safety Evaluation, Salem Unit 2, Automatic Initiation and Flow and Steam Generator Level Indication for the Auxiliary (Emergency)
Feedwater System, June 22, 1981.
Safety Evaluation Report, Supplement No. 4*,
II.E, Auxiliary Feedwater Indication (2.1.7.b
- NUREG-0578), April 1980.
correspondence, NRC to PSE&G, Information Request, Enclosure 2, Regulatory Guide 1.97, Revision 2 -- Additional information needed to complete compliance review, July 26, 1983.
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ATTACHMENT 2 SALEM GENERATING STATION UNIT NOS. 1 AND 2 RESPONSE TO NRC NOTICE OF DEVIATION DATED AUGUST 4, 1989 This response addresses Unresolved Item 272/89-13-05 of Inspection Report No. 50-272/89-13 as directed in Appendix A, Notice of Deviation, Reference No. 1.
R.G. 1.97 UNRESOLVED ITEM NO. 50-272/89-13-05 COMPUTER ISOLATION A review of the wiring diagrams associated with the type A variables discussed in the above paragraphs revealed that the method used to derive a signal to the process computer is by routing the instrument loop current signals through a 250 ohm resistor.
The voltage drop across this resistor provides the signal to the computer through two 15000 ohm resistors, one in each of the two lines.
This resistor network is identified on the diagrams as "computer signal conditioner" and provides the isolation between the Class 1 instrument signal and the class 2 plant computer.
The inspector informed the licensee that resistors are generally not considered acceptable isolation devices, even if they are high impedance devices (as defined by the licensee) and part of the original design by the NSSS manufacturer.
The reason for the inspector's concern is that the use of this type of isolation could result in the inadvertent violation of the si~gle failure criterion.
IEEE standard 279-1971, section 4o7.2, "Isolation Devices," states in part that "No credible failure at the output of an isolation device shall prevent the associated protection system channel from meeting the minimum performance requirements specified in the Design Basis.
Examples of credible failures include short circuits, open circuits, and the application of the maximum AC or DC potential." An evaluation of the drawings relating to the pressurizer pressure circuit also shows that computer signal co.ndi tioners were used to isolate the computer input signals from the signals which ultimately control the pressurizer's spray valves, heaters and PORVs.
The drawing do not show whether these devices are Class lE.
Page 1 of 3
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The licensee should evaluate its design and assure that nowhere in its instrument circuits was the single failure criterion violated as a result of its using signal conditioners as electrical isolators between Class 1E and non Class 1E devices.
The licensee should show that its commitment to IEEE 279 with regards to isolation devices has been satisfied.
This item is unresolved pending the licensee's review and resolution of this issue.
{50-272/89-13-05; 50-311/89-12-05)
RESPONSE
The Plant Process Computer (P250) receives input from the analog instrument loops via computer signal conditioners located in the computer interface racks.
Each signal conditioner is comprised of one (1) 250 ohm dropping resistor and two (2) 15 Kohm output resistors. The signal conditioner drops the 4 to 20 ma analog loop signal across the 250 ohm resistor to establish a 1 to 5 volt potential.
This potential is coupled to the P250 computer via the two 15 Kohm resistors (30 Kohms total resistivity).
The signal conditioner is coupled to the P250 computer via a bus & guard relay circuit and an input relay circuit.
This coupling provides contact isolation for non-selected analog instrument loop points.
Points are selected one at a time.
If a point is faulted, selection will halt at the faulted point.
Due to the contact isolation of all non-selected points and the halting of selection at a faulted point, a postulated failure in the P250 computer could be reflected to only one analog instrument loop.
Selected analog instrument loop points are isolated by a coupling transformer which provides frequency to digital conversion and data input to the CPU from the multiplexer.
All RTD bridges have previously been removed from the SGS Unit Nos. 1 and 2 P250 Computers.
The P250 Computers do not impress voltage to any selected points.
Due to the transformer coupling of the selected point and the absence of any impressed voltage from the P250, the 30 Kohms total resistivity protects the analog instrument loop from credible short circuit or unintentional grounding faults internal to the computer.
Page 2 of 3
The resistors of the computer signal conditioners are passive components in that they provide no dynamic response to input.
As passive components, these resistors maintain their function if they maintain their structural/material integrity.
With a loss of structural/material integrity, industry experience has shown that the credible failure mode of a resistor is an increase in the resistivity or an open circuit.
A short circuit is not considered a credible failure mode for resistors.
An open circuit or increased resistivity failure of a
computer signal isolator 15 Kohm resistor would have no effect on the analog instrument loop.
However, an open circuit or increased resistivity failure of a computer isolator 250 ohm resistor would result in the loss of the analog loop indication.
CORRECTIVE ACTION PSE&G is presently evaluating the as-built seismic capability of the computer interface racks.
This evaluation will document the seismic capability of the computer interface racks and the integral components (i.e. computer signal isolators, etc.).
Documentation of the seismic capability of the computer interface racks will ensure that the requirements of IEEE 279-1971, Paragraph 4.7.2 are satisfied relative to postulated single failures due to short circuits, open circuits, and the application of the maximum credible AC or DC potential.
This evaluation is being performed under the previously described (Deviation 272/89-13-04) program to document the as-built compliance level of all R.G. 1.97 equipment.
That program is presently scheduled for completion by November 1990.
The computer interface racks are located in a mild environment and, as such, do not require environmental qualification.
No other failure modes are postulated to prevent the associated analog channels from meeting the performance requirements specified in the design basis.
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