ML20151X348
| ML20151X348 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 03/31/1988 |
| From: | Udy A EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20151X342 | List: |
| References | |
| CON-FIN-A-6483, RTR-REGGD-01.097, RTR-REGGD-1.097 EGG-NTA-7761, EGG-NTA-7761-02, EGG-NTA-7761-2, TAC-51124, TAC-51125, NUDOCS 8808250328 | |
| Download: ML20151X348 (22) | |
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q EGG-NTA-7761 TECHNICAL EVALUATION REPORT CONFORMANCE TO REGULATORY GUIDE 1.97: QUAD CITIES-1 AND -2 Docket Nos 50-254/50-265 Alan C. Ucy Pubitshed March 1988 I
t Idaho National Engineering Laboratory FG&G Idaho, Inc.
I Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Under DOE Consract No. DE-AC07-761001570 FIN No. A6483 34 O
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l ABSTRACT This EG1G Zdaho, Inc., report documents the review of the Regulatory Guide 1.97 submittals for the-Quad Cities Station, Unit Nos. 1 and 2, and identifies areas of nonconformance to the regulatory guide.
Exceptions to Regulatory Guide 1.97 are evaluated and those areas where sufficient basis for acceptability is not provided are identified.
Docket Nns. 50-254 and 50-265 TAC Nos. 51124 and 51125 11
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i FOREWORD This report is supplied as part of the "Program for Evaluating Licensee / Applicant Conformance to RG 1.97" being cenducted for the U.S.
Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Engineering and System Technology, by EG&G Idaho, Inc.,
Electrical, Instrumentation and Control Systems Evaluation Unit.
The U.S. Nuclear Regulatory Commission funded the work under authorization b&R 20-19-10-11-3.
I Docket Nos. 50-254 and 50-265 TAC Nos. 51124 and 51125 l
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1 CONTENTS A85 TRACT,,,....................................,_,,,,,,,,,,,,,,,,,,,,,
it POREWORD,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,
111 1,
INTRODUCTION,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,
1 2.
kEVIEW REQUIREMENT,;,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,
2 3,
EVALUATION,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,
4 3.1 Adherence to Regulatory Guide 1.97,,,,,,,,,,,,,,,.........,
4 3.2 Type A Variables,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,
4 3,3 Excepti on s to Reg ul ato ry Guide 1,97,,,,,,,,,,,,,,,,,,,,,,,,
5 4,
CONCLUSIONS..,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,
17 5,
REFERENCES,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,
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CONFORMANCE TO REGULATORY GtlIDE 1.97: QUA0 CITIES-1 AND -2 1.
INTR 0000T10N l
On December 17, 1982, Generic Letter No. 82-3' (Reference 1) was 4
issued by D. G. Eisenhut, Director of the Division of Licensing, Nuclear Reactor Regulation, to all 11censees of operating reactors, applicantr. for operating licenses, and holders of construction permits. This letter included additional clarification regarding Resulatory Guide 1.97, Revision 2 (Reference 2), relating to the requirements for emergency I
response capability.
These requirements have been published as Supplement No. I to NUREG-0737, "TMI A: tion Plan Requirements" (Reference 3).
1 Commonwealth Edison, the licensee for the Quad Cities Station,
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provided a response to Item 6.2 of the generic letter en August 1,1985 (Reference 4). Schedular information was provided in letters dated January 31,1986(Reference 5), October 6,1986(Reference 6),May28,1987 j
(Re'erencs 7), and May 29,1987(Reference 8). A letter dated l
Notamher 4,1985 (Reference 9). addresses instrumentation readouts for the emergency response facilities. Additional information on the instrumentation provided for Regulatory Guide 1.97 was subr'tted on December 17, 1987 (Reference 10).
This report provides an evaluation of the submitted material.
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2.
REVIEW REQUIREMENTS Item 6.2 of NUREG-0737, Supplement No. 1 sets forth the documentation to be submitted to the NRC describing how the licensee complies with Regulatory Guide 1.97 as applied to emergency response facilities.
The submittal should include documentation that provides the following information for each variable shown in the applicable table of Regulatory Guide 1.97.
1.
Instrument range P
2.
Environmental qualification 3.
Seismic qualification 4.
Quality assurance 5.
Redundance and sensor location C.
Power supply 7.
Location of display 8.
Schedule of installation or upgrade t
The submittal should identify any deviatic.;< taken from the regulatory guide recommendations and provide supporting justification or alternatives for the deviations identified.
Subsequent to the issuance of the generic letter, the NRC held regional meetings in February and March 1983 to answer licensee and applicant questions and concerns regarding the NRC policy on this subject.
At these meetings, it was noted that the NRC review would address only exceptions taken to Regulatory Guide 1.97. Where licensees or applicants explicitly state that instrument systems conform to the regulatory guide, 2
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it pas noted that no further staff review would be necessary. Therefore, this report addresses only exceptions to Regulatory Guide 1.97.
The following evaluation is an audit of the itcensee's submittals based on the review policy described in the NRC regional meetings.
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EVALUATION The licensee provided a response to Item 6.2 of NRC Generic Letter 82-33 on August 1, 1985. Additional information was provided on December 17, 1987. The response describes the licensee's position on post-accident monitoring instrumentation. This evaluation is based on that material. Other schedular information submitted by the licensee is listed in the Section 5 of this report.
3.1 Adherence to Regulatory Guide 1.97 The licensee provided a review of their post-accident monitoring instrumentation that shows instrumentation that presently complies with the recommendations of Regulatory Guide 1.97, discusses modifications to bring instrumentation into full compliance with the regulatory guide, and discusses deviations that the licensee supports as appropriate to the Quad Cities Station design.
The licensee committed to completing all of the modifications required to bring the identified deviations into compliance with Regulatory Guide 1.97 by the completion of the Unit 2 Spring 1988 outage.
One possible exception is that the rescaling of the drywell pressure recorder cannot be completed unt.1 the corresoonding technical specification change has been approved by the NRC.
The change was scheduled for submittal to the NRC in July 1987.
Therefore, we conclude that the licensee has provided an explicit commitment on conformance to Regulatory Guide 1.97.
Exceptions to and deviations from the regulatory guide are noted in Section 3.3.
3.2 Type A Variables Regulatory Guide 1.97 does not specifically identify Type A variables, i.e., those variables that provide the information required to permit the control room operator to take specific manually controlled safety actions.
The licensee classifies the following instrumentation as Type A.
1.
Coolant level in reactor 4
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2.
Reactor' coolant system pressure 3.
Drywell pres:
e 4.
Suppression chamber pressure 5.
Suppression pool water level 6.
Suppression pool water temperature These variables, with exceptions as noted in Section 3.3, either meet or=
will meet the Category I recommendations, consistent with the requirements for Type A variables.
3.3 Exceptions to Regulatory Guide 1.97 The licensee identified deviations and exceptions from Regulatory Guide 1.97.
These are discussed in the following paragraphs.
3.3.1 Neutron Flux Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable. The licensee states that the instrumentation is Category 1, except for environmental and seismic quali'lication of the cables, detectors and the detector drives that are inside the primary containment. These are not qualified for a loss of coolant accident, but meet the original system design criteria.
The licensee states that there is a known relation between the source range reading, when fully withdrawn, and the actual power level. This statement is based on the attenuation factor of the materials in the vicinity of the detector and on the neutron leakage factor that the licensee states are applicable for this boiling water reactor design.
The source range period meter also shows increases or decreases in power level, even with the detectors withdrawn. There are four source range channels 5
per unit which are powered by two separate power sources.
There are also
-4 eight intermediate range monitors that measure down to 5-x 10 percent of full power when fully inserted. These are powered by the same two power sources as the source range instrumentation. Two separate power sources provide motive power for half of the source range drives and half of the intermediate range drives.
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Additionally, the licensee states that a scram can be verified by the following diverse parameters:
1.
Scram relay position indication 2.
Scram valve position indication 3.
Control rod drive scram accumulator lw pressure indication 4.
Scram discharge volume high level alarm 5.
Indication of responses such as makeup flow, pressure decay, and torus pressure increase During our review of neutron flux instrumentation for boiling water reactors, we noted that the detectors and their cables have not satisfied the environmental qualification ~ requirement of Regulatory Guide 1.97.
A Category I system that meets all the criteria of Regulatory Guide 1.97 has been an industry development item.
Based on our review, we conclude that the existing instrumentation is acceptable for interim operation.
The licensee states that they will consider in-core instrumentation for use when such source (or wider) range instrumentation becomes available.
The licensee should follow industry development of this equipment, evaluate newly developed equipment, and install Category 1 instrumentation when it becomes available.
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3.3.2 Drywell Pressure Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable. Therefore, the information should be continuously recorded. The licensee indicates (Reference 10) that this information is recorded over ranges of -5 to 70 psig and -5 to 250 psig. The extended range recorder is environmentally and seismically qualified. We find that the information provided for this instrumentation shows that the instrumentation for this variable meets the recommendations of Regulatory Guide 1.97.
3.3.3 Suppression Pool Pressure The licensee classifies this as a Type A variable, even though it is not a variable defined in the regulatory guide. The licensee states that the instrumentation for the variable drywell pressure will be used for this variable as well, because there are 12 vacuum breakers that keep the suppression pool (or torus) pressure within 0.5 psi of the drywell.
This is within one-half percent of the instrument range and within the accuracy of the instruments. We find this acceptable.
3.3.4 Drywell Sump Level Drywell Drain Sumps Level Regulatory Guide 1.97 recommends Category 1 instrumentation for these variables. The licensee indicates that leakage rate, not sump level, is the parameter of concern.
The leakage rate is monitored by Category 3 flow rate recorders. The leakage rate is determined every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when the sumps are required to be pumped out. A high sump level alarm is caused if the sumps fill in less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; large leaks isolate the sumps.
1 We conclude that appropriate monitoring of the parameters of concern is provided. This conclusion is based on the following:
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(a) for small leaks, the 1:1strumentation is not expected to experience harsh environments during operation.
(b) for larger leaks, the sumps fill promptly and the sump drain lines isolate due to the increase in drywell pressure, thus negating the drywell sump level and drywell drain sumps level instrumentation (c) 'this instrumentation neither automatically initiates nor alerts the operator to initiate operation of a safety-related system in a post-accident situation.
Therefore, we find the Category 3 instrumentation provided acceptable.
3.3.5 primary Containment Pressure Regulatory Guide 1.97 recommends instrumentation with a range of -5 psig to 4 times the design containment pressure of 63 psig (252 psig) for this variable. The licensee's instrumentation has a range of -5 psig to 250 psig. The licensee has chosen this range because the scale is less awkward than the recommended range and because it meets the intent of the regulatory guide.
l We find this deviation of 2 psig out of 250 psig t] be minor (less than 0.8 percent of the recommended range). Therefore, the range is acceptable.
3.3.6 primary Containment Isolation Valve Position From the information provided, we find that the licensee deviates from a strict interpretation of the Category 1 redundancy recommendation.
Only the active valves have position indication (i.e., check valves have no l
position indication).
Since redundant isolation valves are provided, we find that redundant indication per valve is not intended by the regulatory guide.
Position indication of check valves is specifically excluded by l
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...,. -. -,. - -, - - ~,,,.. - -.
..c Table 1 of Regulatory Guide 1.97. Therefore, we find that the instrumentation for'this variable is acceptable.
3.3.7 Radiation Level in Circulatina Primary Coolant The licensee states that the alternate instrumentation provided is justified because of the critical actions to be taken to prevent and mitigate a gross breach of fuel cladding. These actions are to shut down the reactor and to maintain the water level. The licensee states that the post-accident sampling system provides a means of obtaining samples of reactor coolant and determining the status of fuel cladding.
In addition, the primary containment radiation and hydrogen monitors provide information on the status of fuel cladding.
Based on the alternate instrumentation and on the justification provided by the licensee, we conclude that the instrumentation supplied for this variable is adequate and, therefore, acceptable.
3.3.8 Primary Containment Area Radiation Regulatory Guide 1.97 recommends Category 1 instrumentation with a 7
range of 1 to 10 R/hr for this variable.
In Reference 4, the licensee identified the instrumentation as Category 1, except for seismic 8
qualification, with a range of 10 to 10 R/hr in the drywell. The licensee also identiffed Category 3 instrumentation, with a range of 1 to 6
l 10 R/hr, in the torus.
The licensee's identification of the l
instrumentation (Reference 10) states that seismic qualification is in accordance with IEEE Standard 344-1975, except for the display, which meets the original station seisinic design criteria. We find this acceptable.
The licensee states that the detectors have been calibrated over the 8
range of 1 to 10 R/hr. This exceeds the range recommended by Regulatory Guide 1.97.
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3.3.9 Containment and Drywell Hydrogen Concentration Regula, tory Guide 1.97 recommends instrumentation with a range of 0 to 30 percent for this variable. The range of the licensee's instrumentation is 0 to 10 percent. A remote display that is accessible has a range of 0 to 20 percent; however, this remote display is not Category 1 as recommended by the regulatory guide. The licensee states that the O to 10 percent range monitors the hydrogen concentration well into the explosive range and that there are no additional operator actions required for concentrations greater than 10 percent.
The NRC has reviewed the acceptability of this variable as part of their review of NUREG-0737, Item II.F.1.6.
3.3.10 Radiation Excesure Rate Revision 2 of Regulatory Guide 1.97 recommends Category 2
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instrumentation with a range of 10 to 10 R/hr for this variable.
The licensee's instrumentation is Category 3.
Revision 3 of the regulatory guide (Reference 11) changes the recommendation to Category 3 instrumentation; therefore, we find the category of instrumentation acceptable.
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The licensee states that the range is 10 to 10 mR/hr
~4 (10 to 1 R/hr) and that this is adequate for normal operation, for use in determining local accessibility, and for alarming under abnormai conditions in non-LOCA situations. Entry into an area is controlled by portable meters and by emergency plant procedures.
The licensea states that this variable is a function of primary containment and emergency core cooling system fluid radioactivity.
The licensee also states that the use of effluent radioactivity monitors provides a positive indication of a break or of leakage from these systems. While analysis of the post-accident radiation levels expected for the monitor locations shows that the existing radiation exposure rate 10
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monitors have ranges that can be exceeded in a post-LOCA situation, this does not affect accident mitigation. Any entry into the-reactor building under these conditions is based on portable radiation instruments.
From a radiological standpoint, personnel would not be permitted into these monitored areas without portable monitoring (except for life saving) if the radiation levels reached or exceeded the upper limit of the instrumentation provided. Based on the alternative instrumentation used with this variable, we find that the range for the radiation exposure rate monitors is acceptable.
3.3.11 Suppression Chamber Spray Flow The suppression chamber spray is derived from the residual heat removal (RHR) system, and uses the same flow detector that the variable low pressure coolant injection flow uses. The range of this instrue.entation is 0 to 20,000 gpm; the recommended range for the variable suppression chamber spray flow is 0 to 110 percent of design flow. The licensee identified the flow as 275 gpm.
The licensee acknowledges that the instrumentation accuracy is not adequate for measuring 275 gpm.
The licensee states that the piping is sized to limit the rate of the suppression chamber spray flow. The licensee also indicates that other instrumentation is available. This instrumentation includes system valve position indication, pump running indication, pump discharge pressure, suppression pool pressure and suppression pool temperature. The suppression pool pressure and temperature instrumentation determine the ef'ectiveness of the spray.
We find that the instrumentation will provide adequate indication for this variable.
Therefore, this instrumentation is acceptable.
3.3.12 Standby Liouid Control System (SLCS) Flow The licensee has not elected to implement this variable as recommended by Regulatory Guide 1.97.
The licensee's justification is that the SLCS 11 l
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pump discharge pressure provides indication that the SLCS pump is operating.
In addition, the level indication fcr the SLCS storage tank gives indication that flow is occurring. A flow switch indicates flow and squib valve centinuity lights indicate whether the valves are open or shut. Additionally, neutron flux level and period demonstrate the effectiveness of the SLCS.
We find the above instrumentation valid as an alternate indication of SLCS flow.
3.3.13 Cooling Water Flow to Engineered Safety Feature (ESF) System Components The licensee stated (Reference 4) that the diesel generator cooling water system flow is not monitored. The component cooling service water flow is monitored in accordance with the regulatory guide. The diesel generator cooling water provides cooling water to the pump room coolers and to the diesel generators.
Reference 10 states that a subsequent review of the diesel generator cooling water system shows that each diesel generator cooling water pump has Category 2 flow indication, with a range of 0 to 2000 gallons per minute. While the indicators are not in the control room, the licensee states that they are accessible to the operator curing the recovery phase following a LOCA.
Because the indicators are accessible to monitor the operation of this auxiliary supporting feature system, we find that the instrumentation provided is acceptable.
3.3.14 Emergency Ventilation Damper Position Regulatory Guide 1.97 recommends Category 2 indication in the controi room for this variable.
The licensee identifies the following deviations:
The diesel generator room ventilation dampers do not have position indication.
Room high temperature alarms are used instead.
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The main control room damper position indicators are located just outside the main control room in an accessible. area. The main control room fan indication is in the control room and, being interlocked with the dampers, provide information on the damper status.
The licensee states (Reference 10) that this instrumentation is located in a mild environment and meets the Category 2 requirements. Based
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on the information provided, we find that the instrumentation provided for l
this variable is acceptable.
l 3.3.15 Secondary Containment Area Radiation Regulatory Guide 1.97 recommends Category 2 instrumentation with a 4
range of 0.1 to 10 R/hr for the Mark I containment for this variable.
3 The licensee's Category 3 instrumentation has a range cf 0.1 to 10 mR/hr. The instrumentation deviates both in category and in the upper four decades of the recommended range.
The licensee states that the range is r.dequate for normal operation, for use in determining local accessibility, and for alarming under abnormal conditions in non-LOCA situations.
Entry into an area is controlled by portable meters and by emergency plant procedures.
The licensee states that this variable is a function of primary containment and emergency core cooling system fluid radioactivity. The licensee also states that the use of effluent radioactivity monitors provides a positive indication of a break or of leakage from these I
systems. Analysis of the post-accident radiation levels expected for the monitor locations shows that the existing radiation exposure rate monitors have ranges that can be exceeded in a post-LOCA situation; however,- this does not affect accident mitigation. Any entry into the reactor building urder these conditions is based on portable radiation instruments.
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From a radiological standpoint, personnel would not be permitted into these monitored areas without portable monitoring (except for life saving) if radiation levels reached or exceeded the upper limit o.f the instrumentation provided. Based on the alternative instrumentation used with this variable, we find that the range for the radiation exposure rate monitors is acceptable.
The licensee concludes that Category 3 instrumentation is acceptable for this variable because using these monitors to detect breach or leakage through primary containment penetration results in ambiguous indications.
We find that Category 3 instrumentation, in conjunction with the noble gas effluent monitors, is acceptable for this variable.
3.3.16 Particulates and Haloge_ns Regulatory Guide 1.97 recommends instrumentation with a range of
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10 to 10 pCi/cc for this variable.
The range of the licensee's grab
-12 sampling instrumentation is stated in Reference 10 to be 10 to 2
10 pC1/cc.
As this exceeds the recommended capacity, we find the provided instrumentation acceptable.
3.3.17 Plant and Environs Radiation
~3 Regulatory Guide 1.97 recommends instrumentation with ranges of 10 4
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to 10 R/hr, photons, and 10 to 10 rads /hr, beta and low energy photons for this variable. The licensee has survey meters with a range of 3
0 to 10 R/hr for this variable.
The licensee states that the reactor building and other plant areas are expected to be inaccessible for post-accident entry.
The licensee 4
states that the expected radiation level for most areas is less than 500 R/hr. As the licensee has determined that the range provided (0 to 3
10 R/hr) is adequate, the supplied instrumentation is acceptable.
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-3.3.18 Plant.nd Environs Radioactivity Regulatory Guide 1.97 recommends portable instrumentation (i.e.,
instrumentation that is not in fixed locations) for this variable. The itcensee is developing procedures that will utilize an analyzer that is in a fixed location. The analyzer will use samples that are taken, as required, in the pisnt and from the environs areas. The licensee states that portable equipment should not be used because of the rough handling it would receive in the field.
The laboratory equipment at this station can provide isotopic analysis and a timely assessment of radioactive releases. Therefore, this is an acceptable deviation from Regulatory Guide 1.97.
3.3.19 Estimation of Atmospheric' Stability Regulatory Guide 1.97 recommends instrumentation for this variable with a range of -9'F to +18*F or an analogous range for alternate stability analysis. The licensee has supplied instrumentation with a range of -10*F to +10*F, based on a elevation differantfal of either 115 or 265 feet.
Table 1 of Regulatory Guide 1.23 (Reference 12) provides seven atmospheric stability classifications based on the difference in temperature per 100 meters elevation change. These classifications range from extremely unstable to extremely stable. Any temperature difference greater than +4*C or le.ss than -2*C does nothing to the stability classification. The licensee's instrumentation includes this range.
Therefore, we find that this instrumentation is acceptable to determine the atmospheric stability.
3.3.20 Accident Samoling (primary coolant, containment air and sumo)
The licensee's sampling system obtains samples and provides the analyses within the ranges recommended for this variable with the following exceptions:
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Containment air hydrogen content - the range is 0 to 10 percent o
Containment air oxygen content - the range is 0 to 10 percent o
The licensee considers these ranges adequate because the maximum range covers potential explosive mixturos.
The licensee deviates from Regulatory Guide 1.97 with respect to post-accident sampling cap:>ility. This deviation goes beyond the scope of this review and has been addressed by the NRC as part of the review of NUREG-0737, Item II.D.3.
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CONCLUSIONS Based on our review, we find that the licensee either conforms to or is justified in deviating from Regulatory Guide 1.97, with the following exception:
o Neutron flux--the existing instrumer:tation is acceptable 'until Category 1 instrumentation is developed and installed (Section 3.3.1).
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REFERENCES 1.
NRC letter, D. G. Eisenhut to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Permits,' "Supplement No. I to NUREG-0737--Requirements for Emergency Response Capability (Generic Letter No. 82-33)," December 17, 1982.
2.
Instrumentation for Light-Water-Cooled Nuclear power Plants to Assess Plant and Environs Conditions During and Following an Accident, Regulatory Guide 1.97, Revision 2, NRC, Office of Standards Development, December 1980.
3.
Clarification of TMI Action Plan Requirements, Requirements for Emergency Response Capability, NUREG-0737, Supplement No. 1, NRC, Office of Nuclear Reactor Regulation, January 1983.
4.
Letter, Commonwealth Edison Company (J. R. Wojnarowski) to NRC (D. B. Vassallo), "Compliance with Regulatory Guide 1.97,"
August 1, 1985.
5.
Letter, Commonwealth Edison Company (J. R. Wojnarowski) to NRC (H. R. Denton), "Implementation Schedule for Regulatory Guide 1.97 Modifications," January 31, 1986.
6.
Letter, Commonwealth Edison Company (J. R. Wojnarowski) to NRC (H. R. Denton), "Regulatory Guide 1.97 Modification Schedule,"
October 6, 1986.
7.
Letter, Commonwealth Edison Company (I. M. Johnson) to NRC (T. E. Murley), "Drywell Pressure Instrumentation Regulatory Guide 1.97 Commitment," May 28, 1987.
8.
Letter, Commonwealth Edison Company (I. M. Johnson) to NRC (T. E. Murley), "Reg. Guide 1.97 Commitments Regarding Acoustic Monitoring and Containment Hydrogen Analyzers," May 29, 1987.
9.
Letter, Commonwealth Edison Company (J. R. Wojnarowski) to NRC (H. R. Denton), "Emergency Response Facility Regulatory Guide 1.97 Review," November 4, 1985.
10.
Letter, Commonwealth Edison Company (I. M. Johnson) to NRC (T. E. Murley), "Conformance to NRC Regulatory Guide 1.97, Rev. 2,"
December 17, 1987.
11.
Instrumentation for Licht-Water-Cooled Nuclear Power Plants to Assess plant and Environs Conditions During and Following an Accident, Regulatory Guide 1.97, Revision 3, NRC, Office of Nuclear Regulatory Research, May 1983.
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- 12. Onsite Meteorological Programs, Regulatory Guide 1.23, NRC February 17, 1972 or Metec'ological Programs in Succort of Nuclear power plants, Proposed Re81sion 1 to Regulatory Guiae 1.23, NRC, Office of Standards DevMupment, Septemoer,1980, 18 J