ML20234D548
| ML20234D548 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/31/1986 |
| From: | Stoffel J EG&G IDAHO, INC. |
| To: | NRC |
| Shared Package | |
| ML20234D495 | List: |
| References | |
| CON-FIN-A-6483, RTR-REGGD-01.097, RTR-REGGD-1.097 TAC-51083, NUDOCS 8707070252 | |
| Download: ML20234D548 (23) | |
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CONF 0RMANCE TO REGULATORY GUIDE 1.97 CRYSTAL RIVER, UNIT NO. 3 1
1 J. W. Stoffel Published May 1986 EG&G Idaho. Inc.
Prepared for the U.S. Nuclear Regulatory Comission Washington, D.C.
20555 Under DOE Contract No. DE-AC07-761001570 FIN No. A6483 9707070252 B70616 PDR ADOCK 0500 2
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I ABSTRACT This EG&G Idaho Inc., report reviews the submittals for Regulatory Guide 1.97 for Unit No. 3 of the Crystal River Station and identifies areas s
I of nonconformance to the regulatory guide.
Exceptions to Regulatory Guide 1.97 are evaluated.
9 Docket No. 50-302 TAC No. 51083 11
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FOREWORD This report is supplied as part of the " Program for Evaluating Licensee / Applicant Conformance to R.G. 1.97," being conducted for the U.S.
Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of PWR Licensing-A, by EG6G Idaho, Inc., NRR and I&E Support Branch.
The U.S. Nuclear Regulatory Commission funded the work under authorization 20-19-10-11-3.
Docket No. 50-302 TAC No. 51083 iii
CONTENTS I
ABSTRACT..............................................................
11 FOREWORD..............................................................
iii 1.
INTRODUCTION.....................................................
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2.
REVIEW REQUIREMENTS..............................................
2 3.
EVALUATION.......................................................
4 3.1 Adherence to Regulatory Guide 1.97.........................
4 3.2 Type A Variables...........................................
4 3.3 Exceptions to Regulatory Guide 1.97........................
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4.
CONCLUSIONS.....................................................
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5.
REFERENCES......................................................
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CONFORMANCE TO REGULATORY GUIDE 1.97 CRYSTAL RIVER. UNIT NO. 3 i
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INTRODUCTION On December 17, 1982, Generic Letter No. 82-33 (Reference 1) was issued by O. G. Eisenhut, Director of the Division of Licensing, Nuclear Reactor Regulation, to all licensees of operating reactors, applicants for operating licenses and holders of construction permits. This letter included additional clarification regarding Regulatory Guide 1.97, Revision 2 (Reference 2), relating to the requirements for emergency response capability.
These requirements have been published as Supplement j
No. 1 to NUREG-0737, "TMI Action Plan Requirements" (Reference 3).
l Florida Power Corporation, the licensee for Crystal River, Unit No. 3, provided a response on August 21, 1984 (Reference 4) containing the information required by Section 6.2 of the generic letter. Additional information was provided on November 15, 1985 (Reference 5) and March 27, 1986 (Reference 6).
This report provides an evaluation of these submittals.
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REVIEW REQUIREMENTS Section 6.2 of NUREG-0737, Supplement No. 1, sets forth the documentation to be submitted in a report 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.
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1.
Instrument range 2.
Environmental qualification
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3.
Seismic qualification 4.
Quality assurance 5.
Redundance and sensor location 6.
Power supply q
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Location of display 8.
Schedule of installation or upgrade The submittal should identify deviations from the regulatory guide and provide supporting justification or alternatives.
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 only address 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 was noted that no further staff review would be necessary.
Therefore,
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this report only addresses exceptions to Regulatory Guide 1.97.
The following evaluation is an audit of the licensee's submittals based on the review policy described in the NRC regional meetings.
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3.
EVALUATION The licensee provided a response to Section 6.2 of the generic letter on August 21, 1984. Additional information was provided on
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November 15, 1985 and March 27, 1986.
This evaluation is based on these submittals.
3.1 Adherence to Regulatory Guide 1.97 The licensee included a schedule in t?cir submittals that indicates that they will conform with the recommendations of Regulatory Guide 1.97, Revision ~3 (Reference 7).
All modifications are scheduled to be completed by December 31, 1987. Therefore, we conclude that the licensee has provided an explicit commitment on conformance to Regulatory Guide 1.97.
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Exceptions to and deviations from the regulatory guide are noted in Section 3.3.
3.2 Tyoe 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.
Reactor coolant system (RCS) hot leg water temperature 2.
RCS pressure 3.
Borated water storage tank level 4.
Steam generator level 5.
Steam generator pressure j
l The above instrumentation meets Category 1 requirements consistent with the requirements for Type A variables.
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3.3 Exceptions to Reculatory Guide 1.97 The licensee identified deviations and exceptions from Regulatory Guide 1.97.
These are discussed in the following paragraphs.
3.3.1 RCS Soluble Boron Concentration The licensee does not provide continuous control room monitoring for this variable. They state _that manual sampling and laboratory analysis is
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sufficient based on the fact that-the loss of negative reactivity due to xenon decay is sufficiently slow that the control room operator need not know the boron concentration instantaneously or continuously.
The licensee takes exception to Regulatory Guide 1.97 with respect _to
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this variable. This exception goes beyond the scope of this review and is being addressed by the NRC as part of their review of NUREG-0737, Item II.B.3.
3,3.2 RCS Hot Leo Water Temperature j
Regulatory Guide 1.97 recomends instrumentation with a range of 0 to 700*F for this variable. The licensee has supplied instrumentation with a range of 120 to 920*F. The licensee submitted the following justification for this deviation; (1) at temperatures less than 280*f, the plant will be in the decay heat removal mode and this temperature is not required (2) cold shutdown is defined in the technical specification as less than 200*F, and (3) RCS cold leg temperature range will indicate down to 50*F.
Based on the licensee's justification we find that the range of 120 to 920*F is acceptable.
4 3.3.3 RCS Cold Leo Water Temperature Regulatory Guide 1.97 recommends Category 1 instrumentation with a range of 50 to 700*f for this variable. The licensee has supplied 1
Category 3 instrumentation with a range of 50 to 650*F. There are, deviations in both range and category.
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1 The licensee's justification for Category 3 instrumentation is that the cold leg temperature may not provide valid information on the status of core cooling.
It is located in the RCS loops and not the reactor vessel, so there must be either forced or natural circulation flow through the steam generators for the indication to be representative of actual core conditions, Also, the licensee states that due to the proximity of the cold leg RTDs to the high pressure injection (HPI) nozzles, HpI flow may significantly affect the cold leg temperature indication particularly in the absence of forced RCS flow.
Incore temperature monitors provide a direct indication of core cooling independent of whether or not there exists coolant flow through the loops. For these reasons, the licensee has designated the cold leg water temperature instrumentation as backup instrumentation.
f The licensee states that the key variables for monitoring the core cooling function are RCS hot leg water temperature, core exit temperature, and steam generator pressure.
As the licensee has supplied Category 1 instrumentation for these variables, we find the Category 3 RCS cold leg water temperature instrumentation acceptable.
The licensee's justification for the range deviation is based on the RCS cold water temperature instrumentation having the capability to measure a value greater than the saturation temperature for the steam generators, which is approxinately 500*F (based on 1050 psig design pressure).
650'F for the high end of the range provides excess measurement capability and is approximately 110 percent of the design temperature of 600*F.
Based on this justification, we find this deviation acceptable.
3.3.4 Containment Isolation Valve Position From the information provided, we find 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 position indication).
Since redundant isolation valves are provided, we find that redundant indication per valve is not intended by the regulatory 6
i guide.
Position indication of the check valves is specifically excluded by j
Table 3 of Regulatory Guide 1.97.
Therefore, we find that the instrumentation provided for this variable is acceptable.
3.3.5 Radiation level in Circulating primary Coolant The licensee states that the letdown line radiation monitors are used 1
to provide indication of fuel failure during normal operation.
The letdown line is isolated during serious accidents requiring containment isolation.
The primary means of monitoring this variable after containment isolation is by the post-accident sampling system which is being reviewed by the NRC
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as part of their review of NUREG-0737, Item II.8.3.
i Based on the alternate instrumentation provided by the licensee, we conclude that the instrumentation supplied for this variable is adequate
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and, therefore, acceptable.
3.3.6 Residual Heat Removal (RHR) Heat Exchanger Outlet Temperature The licensee has not complied with the range recommended by the regulatory guide for this variable (40 to 350*F).
The existing range is 0 to 300*F. The licensee states that the design temperature of the decay heat system and heat exchanger is 300*F.
The licensee states that the RHR design temperature covers the anticipated requirements for normal operation, anticipated operational occurrences and accident conditions.
Based on this, we find that the range is adequate and that this deviation is acceptable.
3.3.7 Accumulator (Core Flood) Tank Level and Pressure The licensee has taken exception to providing Category 2 instrumentation for the accumulator tank pressure instrumentation and states that the key variable to determine whether the core flood tanks have 7
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fulfilled their safety function is core flood tank level.
Therefore, core
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flood tank pressure is designated a backup variable by the licensee and has been classified as Category 3.
The accumulators are passive devices.
Their discharge into the reactor coolant system (RCS) is act9ated solely by a decrease in RCS pressure. We find that the instrun.cntation provided for this variable is adequate to determine that the accumulators have discharged. Therefore,
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this instrumentation is acceptable for this variable.
j 3.3.8 Boric Acid Charaino Flow The licensee does not have instrumentation for this variable. The
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licensee states that the unit does not have boric acid charging flow as a safety injection system. High p^ essure injection flow, low pressure injection flow and core flood tar,k level are the safety injection variables j
monitored. Therefore, we find titat this variable is not applicable at the Crystal River Station.
3.3.9 Refueling Water Storace Tank Level A range of 0 to 600 inches has been provided for this variable instead of the top to bottom range recommended by the regulatory guide. Although 600 inches is not the top of the tank, it is the maximum fill level.
This deviation is minor with respect to the total volume of the tank.
The existing instrumentation is adequate to monitor the operation of the tank during all accident and post-accident conditions. Therefore, this is an acceptable deviation from Regulatory Guide 1.97.
3.3.10 Pressurizer level
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Regulatory Guide 1.97 recommends instrumentation for this variable with a range from the top to the bottom of the pressurizer vessel. The l
licensee states that the reactor coolant system can experience a reactor trip from full power without uncovering the level sensors in the lower shell while maintaining system pressure above the HPI system actuation 8
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setpoint.
The steam volume is such that the reactor coolant system can experience a turbine trip without covering the level sensors in the upper
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shell. The licensee states that the range of 0-320 inches was based on this criteria and setpoints for automatic or manual actions are based on this range.
The pressurizer is 512 inches tall. The zero reference for the range is 96 inches from the bottom, 16 inches below the upper set of heaters, and approximately at the level of the second set of heaters. The upper
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pressurizer level is 92 inches from the top (approximately 37 inches from l
the spray head).
j In Reference 6, the licensee states that the existing range is sufficient to remain on scale for anticipated transients.
For severe transients (accidents) the pressurizer will either void or go solid. This would cause the pressurizer level indication to go off-scale low or high depending on the accident, regardless of the span of the range.
In these cases of off-scale pressurizer instrumentation, action to be taken must be determined by subcooling margin, RCS 3ressure, PORV status and pressurizer safety valve status. These indications are all available in the control room.
Based on the, licensee's justification and the available alternate instrumentation, we conclude that indication of pressurizer level outside of the range provided would provide no significant additional information.
Therefore, we find this to be an acceptable deviation from Regulatory Guide l.97.
/3.3.11 Pressurizer Heater Status Regulatory Guide 1.97 recommends instrumentation to monitor the current drawn by the pressurizer heaters. The licensee's instrumentation consists of on/off indication of the redundant pressurizer heaters. The licensee indicates that the control of these heater banks is either on or off and, therefore, the instrumentation is appropriate.
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Section II.E.3.1 of NUREG-0737 requires a number of the pressurizer heaters to have the capability of being powered by the emergency power sources.
Instrumentation is to be provided to prevent overloading a diesel generator.
In Reference 6, the licensee has maintained the position that an on-off mode of indication is adequate to monitor this variable.
The licensee bases this on the fact that the two emergency heater banks are
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either "on" or "off" and are not controlled by modulating the current to them. The licensee further states that the heater current can be monitored with the diesel ammeters when the heaters are loaded onto the diesels.
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We find the justification provided by the licensee unacceptable. A means of monitoring pressurizer heater current in the control room should
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be provided.
3.3.12 Ouench Tank Temperature Regulatory Guide 1.97 recomends instrumentation for this variable with a range from 50 to 750*F. The instrumentation will have a range of 0 to 400*F. The licensee states that the quench tank is equipped with a rupture disc that relieves pressures at 110 psig. The corresponding saturation temperature is 345'F.
Since the rupture disc will preclude the tank temperature from exceeding 345'F, we find the licensee's range uf 0 to 400'F acceptable.
3.3.13 Steam Generator pressure Regulatory Guide 1.97 recomends instrumentation for this variable with a range from 0 to 20 percent above the lowest safety valve setting.
The lowest safety valve setting is 1050 psig; therefore, the range should be from 0 to 1260 psig. The instrumentation for this variable has a range of 0 to 1200 psig, 9 percent above the highest safety valve setting of 1100 psig. The licensee states that the steam relief capacity is 20 to.25 10
percent above the expected steam flow rate and that excess relief capacity is required to be maintained when safety valves are inoperable, and that the FSAR analysis indicates a maximum steam pressure of about 1100 psig.
Based on the licensee's justification, the range cf 0 to 1200 psig is acceptable for this variable.
'3.3.14 Safety / Relief Valve positions or Main Steam Flow Regulatory Guide 1.97 recommends Category 2 instrumentation to monitor this variable. The licensee's submittal indicates that there is no instrumentation installed to read either valve position or main steam flow.
In Reference 6 the licensee committed to install instrumentation to monitor this variable prior to the completion of refueling outage 7.
We find this committment accaptable.
3.3.15 Containment Sump Water Temperature Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable with a range of 50 to 250'f. The licensee does not have instrumentation for this variable. Their justification is that monitoring the sump temperature is not needed to assure that net positive suction head (NPSH) exists for the decay heat removal pumps and no automatic or manual actions are required to protect safety equipment, based on this temperature. Further, the licensee states that containment atmosphere temperature instrumentation provides the most direct indication of containment cooling system operation. In addition, other instrumentation is provided by the licensee to monitor the reactor building spray system and reactor building emergency cooling system.
The sump water temperature can be determined by the use of the low pressure injection system inlet temper 6ture while in the recirculation mode.
We find this alternative instrumentation acceptable to monitor the containment cooling system operation.
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3.3.16 Makeup Flow-In Letdown Flow-Out Regulatory Guide 1.97 recommends Category 2 instrumentation for these variables. The licensee has classified this instrumentation as Category 3 for the following reasons:
1.
During design basis events, the makeup and purification system (MU&PS) is isolated.
2.
During normal operation and certain design basis accidents where l
the MU&PS is still operable, the Category 2 makeup tank level is the key variable used to provide indication of proper MU&PS operation.
3.
Since the makeup tank is a surge volume for the RCS, makeup tank level and pressurizer level indications can be used to qualitatively assess makeup flow into the RCS and letdown flow from the RCS.
Since pressurizer level instrumentation is Category 1 and makeup tank level instrumentation is Category 2, high quality instrumentation is available to provide information on the status and operation of the MU&PS.
The Category 3 makeup flow-in and letdown flow-out are considered as backup instrumentation. Therefore, this is an acceptable deviation from Regulatory Guide 1.97.
3.3.17 Component Coolina Water Flow to ESF System The licensee has not provided Category 2 flow instrumentation in the control room for this variable. Local flow indication for these systems is available.
Indicated flow measurements in the control room are not deemed necessary by the licensee because the Decay Heat Closed Cycle Cooling (DC) and Nuclear Services Closed Cycle Cooling (SW) systems surge tank levels provide system information to the operator.
The wide range of design flows 12 i
to various ESF components would not necessarily be representative of overall syster., performance according to the licensee.
Service water header pressures and remote actuated valve positions are available to the operator along with the '-"als of the surge tanks.
The licensee indicates that this provides bett:
'all indication of system status.
Since the licensee has committed, in Reference 5, to upgrade the alternate instrumentation to Category 2 (during Refuel VI in 1987), we find the alternate instrumentation acceptable, 3.3.18 Radioactive Gas Holdup Tank pressure Regulatory Guide 1.97 recommends instrumentation for this variable with a range of 0 to 150 percent of design pressure. The tank design pressure is 150 psig and the instrumentation range is 150 psig. The licensee states that the radioactive gas hold-up tanks are equipped with relief valves that are set at 125 psig, and that the range of the pressure indication is 20 percent above the relief valve setting.
The justification for the range of the instrumentation for this variable is acceptable, because the pressure will not exceed the provided range.
3.3.19 Emergency Ventilation Damper position Regulatory Guide 1.97 recommends monitoring the open-closed status of these dampers. The licensee states that the emergency ventilation system dampers are controlled from the fan start circuitry and that they do not have individual control switches. Redundant systems are provided so that a single failure will not defeat their safety function.
Control panel lights show when the fans are operating. fan operation causes the dampers to align their position to their emergency position.
Back-up operational data is provided to the operators by Category 3 low flow and high temperature alarms.
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Based on the licensee's altfrnate instrumentation and justification, we find that the provided instrumentation is adequate to determine that the ventilation system is operational. Therefore, this is an acceptable deviation from Regulatory Guide 1.97.
3.3.20 Radiation Exoosure Rate The licensee takes exception to the instrument range recommended by Regulatory Guide 1.97 (10 to 10 R/hr). Currently installed area radiation monitors cover the range of 0.01 to 10 R/hr.
The licensee states that detection of significant releases by area radiation exposure rate is secondary to that provided by the release path monitoring. Area radiation levels inside the plant are monitored to verify compliance with 10 CFR 20,
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by instruments that are considerably more sensitive (1000x) than required by R.G. 1.97 and are sufficient for supporting, as backup ' instrumentation, the detection of significant releases.
Determination of accessibility of equipment for service or long term surveillance is the function of health physics personnel, costly using portable instrumentation.
Exposure rate instrumentation in areas outside containment have an upper range of 10 R/hr, which the licensee states is adequate for initial assessments of accessibility.
These ranges are based on background readings in the areas in which they are located. Should personnel entry be required in areas where these monitors have gone off scale or indicate a high radiation area, a health I
physics escort would accompany personnel into these areas using portable instrumentation to assess radiation levels.
The licensee does not anticipate, even under emergency conditions, sending personnel into radiation fields beyond the limits of this portable instrumentation.
From a radiological standpoint, if the radiation levels reath or exceed the upper limits of the range (10 R/hr), personhel would not be 14 l
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permitted access to the areas except for life saving. We therefore find the range (0.01 to 10 R/hr) of the radiation exposure rate monitors acceptable.
3.3.21 Condenser Air Removal System Exhaust
-6 Regulatory Guide 1.97 recommends a range of 10 to 10 pC1/cc for this variable. The licensee has provided instrumentation with a range of 2
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x 10 to 10 t.C i/c c.
In Reference 5, the licensee states that the condenser air removal system exhaust is routed into the auxiliary building system exhaust system, where the activity is monitored again before release. We find this arrangernent acceptable to meet the recommendations of Regulatory Guide 1.97.
3.3.22 Noble Gas Vent from Steam Generator Safety Relief Valves or Atmoscheric Oump Valves Regulatory Guide 1.97 recammends Category 2 instrumentation for this variable. The recommended parameters to be monitored for this variable are noble gas, duration of release in seconds and mass of steam per unit time.
In Reference 6 the licensee committed to modify the existing j
'nstrumentation for this variable by June 1, 1986. We find that these j
modifications are adequate to meet the recommendations of Regulatory
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Guide 1.97.
3.3.23 Plant and Environs Radiation (Portable Instrumentation)
The licensee takes exception to the range reconsnended by Regulatory
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Guide 1.97 for this variable (10 to 10 R/hr, photons). The range of the
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provided instrumentation is 10 to 10 R/hr. The licensee does not anticipate encountering radiation fields greater than those which can be measured by their equipment except under severe accident conditions.
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under accident conditions they do not anticipate sending individuals into greater than 10 R/hr fields.
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This instrumentation is portable < and would not be used to assess levels of radiation greates than the range provided by the licensee.
Therefore, this is an acceptable deviation from Regulatory Guide 1.97.
3.3.24 Estimation of Atmospheric Stability Regulatory Guide 1.97 recommends instrumentation for this variable with a range of -5 to +10*C or an analogous range for alternative stability analysis. The licensee has supplied instrumentation with a range of -5 to
+5'C.
The licensee states that this is acceptable based on its meeting the recommendations of Regulatory Guide 1.23.
Table 1 of Regulatory Guide 1.23 (Reference 8) provides seven
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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 or less than -2*C does nothing to the stability classification. The range supplied by the licensee encompasses this range. Therefore, we find that this instrumentation is acceptable to determine attpospheric stability.
3.3.25 Accident Samplina (Primary Coolant. Containment Air and Sumo)
The licensee's post-accident sampling system provides sampling and i
analysis as recommended by the regulatory guide, except that:
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Regulatory Guide 1.97 reconnends a range of 1 pC1/ml to 10 Ci/ml for gross activity. The licensee has provided a range of 10 pCi/ml to 10 C1/ml.
2.
It does not have the capability to analyze for dissolved oxygen.
3.
It does not have containment air oxygen content analysis on-site.
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The licensee takes exception to Regulatory Guide 1.97 with respect to post-accident sampling capability, This exception goes beyond the scope of this review and is being addressed by the NRC as part of their review of NUREG-0737, Item II.B.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.
1.
Pressurizer heater status--the licensee should provide the instrumentation recomended Dy Regulatory Guide 1.97.
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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. 1 to NUREG-0737--Requirements for Emergency Response Capability (Generic Letter No. 82-33)," December 17, 1983.
2.
Instrumentation for Licht-Water-Cooled Nuclear Power Plants to Assess j
Plant and Environs Conditions Durina and followina an Accident, Regulatory Guide 1.97, Revision 2, NRC, Office of Standards Development, December 1980.
3.
Clarification of TMI Action Pla,n Recuirements. Requirements for Emergency Response Capability, NUREG-0737 Supplement No. 1, NRC, Of fice of Nuclear Reactor Regviation, January 1983.
j 4.
Florida Power Corporation letter, P. Y. Baynard to Director Office of Nuclear Reactor Regulation, NRC, August 21, 1984, 3F0884-07.
5.
Florida Power Corporation letter, G. R. Westafer to Director of i
Nuclear Reactor Regulation, NRC, November 15, 1985, 3F1185-17.
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6.
Florida Power Corporation letter, G. R. Westafbr to, Director of Nuclear Reactor Regulation, NRC, March 27, 1985, 3F0386-11.
7.
Instrumentation for Licht-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions Durina and Followino an Accident, Regulatory Guide 1.97, Revision 3, NRC, Office of Nuclear Regulatory Research, May 1983.
8.
Regulatory Guide 1.23 (Safety Guide 23), Onsite Meteorological Proarams, NRC, February 17, 1972, or Proposed Revision 1 to Regulatory Guide 1.23, Meteorolooical Procrams in Support of Nuclear Power Plants, NRC, Office of Standards Development, September 1980.
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This EG&G Idaho, Inc. report reviews the submittals for Crystal River, Unit No. 3, and identifies areas of nonconformance to Regulatory
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Guide 1.97.
Exceptions to these guidelines are evaluated,and those areas j
wheresufficientbasisforacceptabilityisnotprovidedareident(fied.
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