ML20112E081

Conformance to Reg Guide 1.97,Arkansas Nuclear One,Unit 1

ML20112E081
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/31/1985
From:	Stoffel J EG&G, INC.
To:	NRC
Shared Package
ML20100F738	List: ML20112E081
References
CON-FIN-A-6483, RTR-REGGD-01.097, RTR-REGGD-1.097 TAC-51069, NUDOCS 8503250088
Download: ML20112E081 (19)

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CONFORMANCE TO REGULATORY GUIDE 1.97 ARKANSAS NUCLEAR ONE UNIT NO. 1 _

I J. W. Stoffel Published March 1985 EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Comission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6483 i

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ABSTRACT This EG&G Idaho, Inc., report provides a review of the Arkansas Nuclear One, Unit No.1, submittal for Regulatory Guide 1.97 and identifies areas of nonconformance to the guide. Any exceptions to the guidelines are evaluated and those areas where sufficient basis for acceptability is not pro-vided are identified.

FOREWORD

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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 Reg-ulatory Commission Office of Nuclear Reactor Regulation Division of Systems Integration by EG&G Idaho, Inc., NRC Licensing Support Section.

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l The U.S. Nuclear Regulatory Commission funded the work under authoriza-tion 20-19-10-11-3.

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Docket No. 50-368 l t

TAC No. 51069 l

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4 CONTENTS ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 FOREWORD . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 1.

INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

2. REVIEW REQUIREMENTS ........................ 2 3.

EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 i

3.1 Adherence to Regulatory Guide 1.97 .............. 4

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3.2 Type A Variables .......................

4 3.3 Exceptions to Regulatory Guide 1.97 . . . . . . . . . . . . . . 5

4. CONCLUSIONS

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5. REFERENCES . . . .

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CONFORMANCE TO REGULATORY GUIDE 1.97 -

ARKANSAS NUCLEAR ONE. UNIT NO. I

1. INTRODUCTION On December 17, 1982, Generic Letter No. 82-33 (Reference 1) was issued by D. G. Eisenhut Director of the Division of Licensing, Nuclear Reactor Reg-ulation 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), re-lating to the requirements for emergency response capability. These require- -

ments have been published as Supplement No. I to NUREG-0737, "TMI Action Plan Requirements" (Reference 3).

Arkansas Power and Light Company, the licensee for Arkansas Nuclear One, Unit No.1, provided a response to the Regulatory Guide 1.97 portion of the generic. letter on June 25, 1984 (Reference 4).

1 This report provides an evaluation of that submittal.

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2. REVIEW REQUIREMENTS E i

Section 6.2 of NUREG-0737, Supplement 1 st:ts forth the documentation to be submitted in a report to the NRC describing how the licensee meets the i guidance of Regulatory Guide 1.97 as applied to emergency response facili- i ties. The submittal should include documentation that provides the following information for each variable shown in the applicable table of Regulatory Guide 1.97.

i 1. Instrument range

2. Environmental qualification

3. Seismic qualification '

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4. Quality assurance

5. Redundance and sensor location

6. Povar supply

7. Location of display

8. Schedule of installation or upgrade Furthermore, the subraittal should identify deviations from the guidance in the regulatory guide and provide supporting justification or alternatives.

l Subsequent to the issuance of the generic letter, the NRC held regional  !

, meetings in February and March 1983, to answer licensee and applicant ques- l tions and concerns regarding the NRC policy on this matter. At these meet-  ;

) ings, it was noted that the NRC review would only address exceptions taken to l

the guidance of Regulatory Guide 1.97. Furthermore, where licensees or ap- -

plicants explicitly state that instrument systems conform to the provisions 2

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of the guide it was noted that no further staff review would be necessary.

Therefore this report only addresses exceptions to the guidance of Regulatory Guide 1.97. The following evaluation is an audit of the licensee's submittal based on the review policy described in the NRC regional meetings.

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3. EVALUATION The licensee provided a response to the NRC generic letter 82-33 on June 25, 1984. This evaluation is based on that submittal.

3.1 Adherence to Regulatory Guide 1.97 The licensee stated that based on the information presented in their sub-mittal, Arkansas Nuclear One, Unit No. 1, will conform with the recommenda- -

tions of Regulatory Guide 1.97, Revision 3 (Reference 5), by the end of the

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next two refueling outages. Therefore, it is concluded that the licensee has '

provided an explicit commitment on conformance to the guidance of Regulatory Guide 1.97. Exceptions to the regulatory guide are noted in Section 3.3.

l 3.2 Type A Variables Regulatory Guide 1.97 does not specifically identify Type A variables, 1.e., those variables that provide information required to permit the control room operator to take specific manually controlled safety actions. The licen-see classifies the following instrumentation as Type A:

1. Reactor coolant system (RCS) hot leg water temperature

2. RCS pressure

3. Containment hydrogen concentration

, 4. Steam generator level

5. Steam generator pressure.

i 6. Condensate storage tank level

7. Borated water storage tank level.

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All of the above variables meet Category 1 requirements consistent with the -

requirements for Type A variables.

3.3 Exceptions to Regulatory Guide 1.97

, 3.3.1 RCS Soluble Boron Concentration The licensee has not provided indication with the recommended range for this variable. The licensee states that the letdown line for the boronometer _

is isolated following an accident, and that the Post-Accident Sampling System #

(PASS) for ANO-1 'was designed to provide boron concentration measurement - -

capability. The range and design criteria for the PASS is consistent with guidance provided for Item II.B.3 of NUREG-0737.

The licensee takes exception to the guidance of Regulatory Guide 1.97 with respect to this variable. The 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 Cold Leo Water Temperature The licensee has taken exception to the Category 1 and the range (50 to 700*F) recommendations of Regulatory Guide 1.97 for this variable. The licensee's instrumentation is Category 3 with a range of 50 to 650*F. The licensee states that there must be RCS flow through the steam generators for this variable to represent core conditions. Also, due to the pro.ximity of the i . cold leg RTD's to the high pressure injection (HPI) nozzles, HPI flow may significantly affect the cold leg temperature indication partiedlarly in the absence of forced RCS flow. Incore temperature monitors are defined by the licensee as the key for this function as it provides a more direct indication of core cooling independent of whether or not there exists coolant flow

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through the loops. The licensee states that RCS cold leg temperature serves J

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as a backup variable and is qualified to Category 3 requirements -

accordingly.

We find that the existing Category 3 RCS cold leg water temperature  ;

indication is adequate. However, the licensee should provide justification for the range deviation.

3.3.3 Degrees of Subcooling Regulatory Guide 1.97 recommends instrumentation with a range of 200*F ~~

subcooling to 35'F superheat for this variable. The licensee has instrumenta - -

tion with a range of 0 to 200*F subcooled for this variable. The licensee states that during situations that may result in superheated conditions in the RCS, incore temperature and hot leg RTD's will be monitored against RCS pres-sure to determine the degree of superheat or subcooling, and that the SPDS computers will plot the core exit temperature versus pre.ssure on a grid with the saturation curve so that the operator can tell at a glance the thermo-dynamic status of the reactor coolant. Therefore, the licensee states that a range of 0*F to 200*F subcooled on the subcooled margin monitor is adequate to meet the intent of Regulatory guide 1.97.

The RRC is reviewing the acceptability of this variable as part of their review of NUREG-0737, Item II.F.2.

l 3.3.4 Radioactivity Concentration or Radiation Level in Circulating Primary Coolant

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) The licensee has chosen the following alternate instrumentation to j monitor the radiation level in the RCS during accident and post-accident conditions:

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l letdown line radiation monitor I

radiochemistry analyses

.l post-accident sampling system l

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Based on the justification supplied by the licensee, we conclude that the -

instrumentation supplied for this variable is adequate and, therefore, accept-able.

3.3.5 Containment Effluent Radioactivity--Noble Gases from Identified Release Points Effluent Radioactivity--Noble Gases from Buildings or Areas Where Penetrations and Hatches are Located Condenser Air Removal System Exhaust Common Plant Vent '

The licensee has not provided high-reliability power sources for these variables as recommended by Regulatory Guide 1.97. The licensee states that -

the Gaseous Effluent Radiation Monitoring System (GERMS) for ANO-1 was de-signed to comply with the requirements of NUREG-0737. Item II.F.1, and that these requirements did not include providing a highly reliable power supply.

This is insufficient justification for this deviation. The licensee should provide high-reliability power sources for these variables.

3.3.6 Accumulator Tank (Core Flood Tank) Level and Pressure The licensee has taken exception to supplying Category 2 instrumentation for the core flood tank pressure. The licensee states that the Core Flood Tank Pressure is the key variable for preaccident status for assuring that this passive safety system is prepared to service its function. The licensee states that the key variable necessary to determine whether the Core Flood Tanks have fulfilled their safety function is Core Flood Tank Level. There-fore, Core Flood Tank Pressure is a backup variable and has accordingly been classified as a Category 3 variable.

I The accumulators are passive devices. Their discharge into the reactor coolant _ system (RCS) is actuated solely by a decrease in RCS pressure. We find that the instrumentation provided for this variable is adequate to

! determine that the accumulators have discharged. Therefore, the instrumen-tation for this variable is acceptable.

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3.3.7 Boric Acid Charging Flow ~

, The licensee does not have instrumencation for this variable. Their nuclear steam supply system design does not include the charging system as i part of the emergency core cooling system. The high pressure injection and j low pressure injection are the flow paths to the RCS that are monitored and this is done with Category 2 instrumentation. Therefore, we find that this j variable is not applicable at Arkansas Nuclear One, Unit 1.

l 3.3.8 Quench Tank Temperature The licensee has not provided quench tank temperature indication, stating that the quench tank level and pressure are available for monitoring quench tank performance and that the liquid temperature will always be less than or equal to the saturation temperature for the quench tank pressure.

Temperature indication for the quench tank is a necessary parameter.

Monitoring the temperature when slow leakage into the tank exists or after a pressurizer relief valve has lifted provides useful information. The licensee should install temperature indication for this tank.

- 3.3.9 Steam Generator Pressure The licensee has instrumentation for this variable that does not meet the  ;

range recommended by the regulatory guide (20 percent above the lowest safety valvesetting).

ANO-1 has control room indication of main steam pressure at the turbine as well as at the steam generators. The steam generator outlet pressures are measured over the' range from 0-1200 psig. Since the lowest safety valve is 1050 psig, the range should be to 1275 psig.

The main steamlines are provided with relief valves, atmospheric dump .

valves, and condenser dump valves to prevent overpressurization of the lines as well as for pressure control capability. The licensee has approximately '

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40 percent excess steam relief capacity with all safety valves and condenser -

dump valves operable. ANO-1 technical specifications require 14 of the 16 main steam safety valves to be operable during power operation. The licensee states that fourteen safety valves will relieve approximately 110 percent of rated steam flow. Combined with the cond'nser e dump valves, these 14 safety valves will provide a total of approximately 25 percent excess steam relief capacity.

Based on the highest safety valve setting being 1100 psig, and an excess .

relief steam capacity of approximately 25 percent being maintained when as many as two safety valves are inoperable, we find the existing range to be ~ ~

adequate to monitor steam generator pressure during all accident and post-accident conditions. Therefore, this is an acceptable deviation from Regula-

, tory Guide 1.97.

3.3.10 Containment Atmosphere Temperature The licensee has taken exception to the Category 2 and range (40 to 400*F) recomendations of Regulatory Guide 1.97. The instrumentation provided is Category 3 with a range of 0 to 300*F. The licensee states that the re-actor building atmosphere temperature is not a key variable for accident monitoring; that the key variable for reactor building monitoring is reactor

! building pressure which is measured by Category 1 instrumentation; that the f reactor building atmosphere temperature is a backup variable for reactor building accident monitoring and as such is measured by Category 3 instru-mentation with a range of 0 to 300*F. This range is justified by the licensee based on a safety analysis which demonstrates that the worst case peak reactor j buildingtemperaturewouldbe286.5'F(FSARTable6-9).

i The licensee indicates that the maximum containment temperature will be less than 287'F. Therefore, the range of 0 to 300*F is acceptable. We also 1 find that the licensee's application of Category 3 backup instrumentation is in accordance with the regulatory guide.

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l 3.3.11 Containment Sump Water Temperature

• The licensee has not provided instrumentation for this variable. The licensee states that the FSAR accident analysis assumes saturated conditions for sump water during sump recirculation, and with this conservative assump-tion, adequate NPSH exists for reactor building spray and safety injection pumps at all feasible sump water temperatures.

Pump NPSH is not the only consideration. The licensee should address the -

purpose of this instrumentation as stated in the regulatory guide. The 11cen-

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complished, or provide the recommended instrumentation for this variable.

3.3.12 Makeup Flow-in Letdown Flow-out Volume Control Tank Level i

The licensee takes exception to the Category 2 reconnendations of the

regulatory guide for these variables. Category 3 instrumentation is provided by the licensee. The licensee states that for accidents in which harsh en-vironments a,re a result, the system containing this instrumentation (letdown j and makeup portion of the Makeup and Purification System) is not required.

Letdown is automatically isolated upon an ESF actuation.

As these variables are not utilized at ANO-1 in conjunction with a safety j system, we find that the instrumentation provided is acceptable.

3.3.13 Component Cooling Water Temperature to ESF System The licensee has not provided a readout in the control room for this variable. The licensee states that the service water system is used for this variable.

The inlet temperature of the service water by design is based on a maximum temperature of 129.5* from the emergency cooling pond and the average i

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temperature of the pond (June through September) is 85'F. Furthermore, there .

is no control over the temperature of the service water.

The justification submitted by the licensee for this deviation is ade-quate. Therefore, this is an acceptable deviation from Regulatory 4

Guide 1.97.

4 3.3.14 Component Cooling Water Flow to ESF S_ystem I

Regulatory Guide 1.97 recommends a range of 0 to 110 percent design flow I for this variable. The licensee uses pump pressure and valve position to _ -

monitor the operation of this system. The licensee states that the design flow to various ESF components varies from 6 gpm for the reactor building

spray pump bearing coolers to 3000 gpa for the decay heat removal coolers.

The licensee states that due to this wide range of design flows to ESF components, total loop flows would not be indicative of overall system performance. The licensee states proper system operation is shown by the i

correct service water header pressure and by knowing that the remote actuated i

valves supplying service water to ESF components are in their proper posi-tions. Service water header pressure and remote actuated valve position indications are available in the control room and meet Category 2 qualificatio'n~ requirements.

We find this instrumentation acceptable.

4 i .3.3.15 Radioactive Gas Holdup Tank Pressure

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l The licensee has not provided control room instrumentation for this vari- l able. The licensee states that in the event of an accident which results in 1

significant failed fuel or significant radioactive gas release, the manual transfer of radioactive gases to the radioactive gas holdup tanks would not be attempted. There are no automatic transfer operations involving the radio-l active gas holdup tanks. The licensee states that the monitoring of the '

radioactive gas holdup tanks during post-accident conditions is not necessary since these tanks are not utilized for accident mitigation.

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I Since these tanks are not used for accident mitigation at this station,

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this is an acc-,., table deviation from Regulatory Guide 1.97. l

, 3.3.16 Radiation Exposure Rate The licensee takes exception to the instrument range recomended by Reg-ulatory Guide 1.97(10-1 to 104 R/hr) for all of their radiation monitors.

l ANO-1 currently has an Area Radiation Monitoring System consisting of 20 Area Monitors: four with a range of 10-2 to 103 R/hr and 16 with a range ',,

of 10-4 to 101 R/hr. The licensee states that 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, the licensee states that a health physics escort would accompany personnel into these areas using portable instrumentation to assess radiation levels. The high range for portable instrumentation at ANO

. is 103 R/hr. The licensee does not anticipate, even under emergency I

conditions, sending personnel into radiation fields of this magnitude.

From a radiological standpoint, if the radiation levels reach or exceed the upper limit of the range (103 R/hr), personnel would not be permitted to the areas except for life saving. Therefore, we find the ranges for the radi- l ation exposure rate monitors acceptable.  !

t f 3.3.17 Plant and Environ Radiation Regulatory Guide 1.97 recommends portable instrumentation with a range of 10-3 to 10-4 R/hr, photons; and 10-3 to 10 4 rads /hr beta radiation and low-energy photons. The licensee does not comply with the recommended range for 2

this instrumentation'and stated that the existing portable instrumentation can i detect gama dose rates from 10-3 to 103 R/hr and beta dose rates from 10-3 to 50 rad /hr. The licensee does not anticipate encountering radiation fields

greater than those which can be measured by their current equipment except

, under severe accident conditions. Even under accident conditions. -

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the licensee does not anticipate sending individuals into greater than -l 103 R/hr fields.

This instrumentation is portable and would not be used to assess levels of radiation greater than the range provided by the licensee. Therefore, this is an acceptable deviation from Regulatory Guide 1.97.

3.3.18 Plant and Environs Radioactivity The licensee has not provided portable instrumentation for isotopic

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analysis as recommended by Regulatory Guide 1.97 and stated that gamma spec- - '

troscopy can be performed using equipment in the health physics department and the radiochemistry department at ANO, and in the Technical Analysis Laboratory in Little Rock. In addition, the licensee has an ND-60 spectrometer in the ANO Emergency Offsite Facility which can be used for less defined analysis.

The licensee states that it is not appropriate for this instrumentation be portable due to rough handling it would encounter in the field and the limited amount of time field teams have to assess the release.

The existing laboratory equipment available at this station is adequate I to provide.1.sotopic 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 -5 to 10*C or an analogous range for alternative stability estimates. The licensee has instrumentation with a range of -3 to 5'C temperature differences and 0 to 40' wind direction sigma. The licensee states that atmospheric stability is derived from the temperature differential indicated between 34 and 180 feet; that the -3 to 5'C temperature range covers

the seven Pasquill stability classes vs. Delta-T as derived from Regulatory _

l Guide 1.23 as specified in the ANO-1 FSAR. In addition to temperature

! differential, atmospheric stability can also be calculated for all seven classes using wind direction sigma.

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Table 1 of Regulatory Guide 1.23 (Reference 6) provides seven atmospheric .

stability classifications based on the difference in temperature per 100 meters elevation change. These classifications range from extremely un-stable to extremely stable. Any temperature difference beyond -2* or +4*C does nothing to the stability classification. Therefore, we find this an ac-ceptable deviation from Regulatory Guide 1.97.

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4. CONCLUSIONS

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Based on our review, we find that the licensce conforms to, or is justi-fled in, deviating from the guidance of Regulatory Guide 1.97, with the fol-lowing exceptions:

1. RCS cold leg water temperature--the licensee should provide justift-cation for deviating from the recommended range for this instrumenta-tion (Section3.3.2). -

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2. Containment effluent radioactivity--the licensee should provide a ~ ~

highly reliable power source for this variable (Section 3.3.5).

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3. Effluent radioactivity--the licensee should provide a highly reliable power source for this variable (Section 3.3.5).

4. Condenser air removal system exhaust--the licensee should provide a highly reliable power source for this variable (Section 3.3.5).

5. Common plant vent--the licensee should provide a highly reliable power source for this variable (Section 3.3.5).

6. Quench tank temperature--the licensee should install temperature indication for this tank (Section 3.3.8).

7. Containment sump water temperature--the licensee should provide Cate-gory 2 instrumentation with the recommended range or show why this cannot be done (Section 3.3.11).

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5. REFERENCES ,

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1. NRC letter, D. G. Eisenhut to all licensees of operating reactors, appli-cants for operating licenses, and holders of construction permits, "Sup- i plement No. I to NUREG-0737--Requirements for Emergency Response <

Capability (Generic Letter No. 82-33)," December 17, 1982.

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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. U.S. Nuclear Regulatory Commission (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. Arkansas Power and Light Company letter, John R. Marshall to Mr. Darrel G. Eisenhut, NRC, "NUREG-0737 Supplement 1. Regulatory Guide 1.97," June 25, 1984.

5. Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, Regulatory Guide 1.97, Revision 3. U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, May 1983.

6. Onsite Meteorological Programs, Regulatory Guide 1.23 (Safety Guide 23),

NRC, February 17, 1972, or Meteorological Programs in Support of Nuclear Power Plants, Proposed Revision 1 to Regulatory Guide 1.23, NRC, Office of Standards Development September 1980.

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