ML20213F416

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Conformance to Reg Guide 1.97,Millstone Nuclear Power Station,Unit 3
ML20213F416
Person / Time
Site: Millstone, 05000000
Issue date: 07/31/1985
From: Udy A
EG&G IDAHO, INC.
To:
NRC
Shared Package
ML20213F414 List:
References
CON-FIN-A-6493, FOIA-87-213, RTR-REGGD-01.097, RTR-REGGD-1.097 NUDOCS 8511010320
Download: ML20213F416 (21)
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u-CONFORMANCE TO REGULATORY GUIDE 1.97-MILLSTONE NUCLEAR POWER' STATION, UNIT NO. 3 A. C. Udy j

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EG&G Idaho Inc.

's Idaho Falls, Idaho 83415

~j Prepared for the-U.S. Nuclear Regulatory Comission.

Washington, D.C.- 20555.

Under DOE Contract No. DE-AC07-761001570-FIN' No. A6493.

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- s ABSTRACT This EG&G Idaho, Inc., report reviews the submittals for Regulatory Guide 1.97 for Unit No. 3 of the Millstone Nuclear Power Station and' identifies areas of conconformance to the regulatory guide. Exceptions to j

Regulatory Guide 1.97 are evaluated and those areas where sufficient basis for acceptability is not provided are identified..

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FOREWORD-This report is supplied as part-of the " Program for Evaluating -

j Licensee / Applicant Conformance to RG 1.97," being conducted for the l

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

The U.S. Nuclear Regulatory Comission funded the work under authorization B&R 20-19-40-41-3.

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i ASSTRACT..............................................................

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FOREWORD..............................................................

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INTRODUCTION.....................................................

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REV I E W REQU I RE ME NT S..............................................

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E V AL UAT I ON....................................................... -

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3.1 Adherence to Regul atory Guide 1.97.........................

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3.2 Type A V ar i ab l e s...........................................

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-l 3.3 Exceptions to Regulatory Guide 1.97.......................

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CONCLUSIONS......................................................

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REFERENCES.......................................................

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1 CONFORPANCE TO REGULATORY GUIDE 1.97 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 3

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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 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, f

Revision 2 (Reference 2), relating to the requirements for emergency response capability. These requirements have been published as Supplement No. I to'NUREG-0737, "THI Action Plan Requirements" (Reference 3).'

1 Northeast Utilities, the applicant for. Unit No. 3 of the Millstone Nuclear Power Station, provided a response to the generic letter on 1

April 15,1983 (Reference 4). The letter referred to a previous letter dated February 2,1983 (Reference 5), for a review of the instrumentation provided for Regulatory Guide 1.97. Additional information was provided in letters dated December 16,1983 (Reference 6), January 13, 1984 (Reference 7) and May 28,1985 (Reference 8).

I This report provides an evaluation of these submittals.

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REVIEW REQUIRENENTS -

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 applicant complies with Regulatory Guide l'.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

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Environmental qualification 3.

. Seismic qualification

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Quality assurance 5.

Redundance and sensor location 6.

Power supply.

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Location of display 8.

Schedule of installation or upgrade Furthermore,.the submittal should identify deviations.from the regulatory-guide and provide supporting justification.or alter' natives.

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. Furthermore, where licensees or.

applicants explicitly state that instrument systems: conform or will conform to the' regulatory guide, it was noted that'no further staff review.

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- would be'necessary.. Therefore, this report only addresses exceptions to

.i Regulatory Guide 1.97. The following evaluation is an audit of the i

applicant's submittals based on the review policy described.in the NRC regional meetings.

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3. ' EVALUATION The applicant provided a response to NRC Generic Letter 82-33 on April 15, 1983. This response referred to an earlier submittal of February 2,19'83, which described the applicant's position on post-accident monitoring instrumentation. Additional information was provided on December 16, 1983, January 13, 1984 and May 28, 1985. This evaluation is based on these submittals.

l 3.1 Adherence to Regulatory Guide 1.97 Table 420.6-1 of the applicant's responses dated December 16, 1983, and January 13, 1984, identifies each' variable and shows whether or not the instrumentation provided complies with Regulatory Guide 1.97. Therefore, it is concluded that the applicant has provided an explicit commitment on conformance to Regulatory Guide 1.97.

Exceptions to ano 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 varisbles that provide information required to permit the control room operator to take specific manually controlled safety actions.

The applicant classifies the following instrumentation as Type A.

1.

Reactor coolant system (RCS) pressure (wide range) 2.

RCShotlegwatertemperature(widerange) 3.

RCS cold leg water temperature (wide range)~

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Steam generator level (wide range) 5.

Steam generator level (narrow range) 6.

Pressurizer level 4

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Primary reactor, containment pressure

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Steamline pressure 9.

' Refueling water storage tank level 10.. Containment high range internal radiation monitor 11.;. Core exit temperature

.12. Auxiliary feedwater flow 1

13. Containment sump water level j

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14. Fuel. drop monitors (containment radiation)
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15. RCSpressure(extendedrange) 1
16. Containment hydrogen concentration.

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17. ~RCS subcooling monitor This instrumentation meets Category 1 recommendations consistent with' the note n Sect on 3 3 4 4*

-3.3 Exceptions to Regulatory Guide 1.97' 1

. i The applicant identified deviations and exceptions from Regulatory.

- Guide 1.97. These are discussed in the following' paragraphs.

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. 1 3.3.1 Reactor Coolant System Soluable Boron Concentration i.

Regulatory Guide ~ 1.97 recommends Category 3. instrumentation with a range of 0.to 6000 parts per million for this variable.1 The applicant does -

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q not propose to provide instrumentation for this variable, saying that Category-1 n4utron flux monitoring will adequately perform this function.

The applicant 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 i

NUREG-0737. Item II.B.3.

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3.3.2 Reactor Coolant System Cold and Hot Leg Temperature J

Revision 2 of Regulatory Guide 1.97 recommends Category 1 ir.strumentation for'these variables with a range of 50 to 750*F. The applicant has instrumentation with a range of 0 to 700*F, and which is.

i considered non-redundant by the applicant. The applicant justifies the range deviation by stating that it would require a highly improbable event,

- l far beyond the plant design bases, to exceed the saturation pressure corresponding to 700*F (3100 psig).

I We find that the supplied instrumentation range is adequate, as Revision 3 of Regulatory Guide 1.97 (Reference 9). lists the upper limit -

recommendation as 700*F. This is met by the licensee.

The applicant inoicates in Table 7.5-1 of the Final Safety Analysis

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i Report (FSAR, Reference 9) that there is 1 instrument channel per loop for both the cold and hot legs for this 4 loop unit. We consider.this as redundant instrumentation. The power sources are uninterruptable power' supplies (UPS) that are backed by batteries and a. diesel generator. The i

hot leg temperature instrumentation is powered by one UPS, the cold leg temperature instrumentation is powered by a second UPS. Diverse instrumentation (core exit temperature and steamline pressure) are powered by additional UPS power sources. These power sources were previously:

reviewed by the NRC and found acceptable..We find this to be a good faith-j attempt, as defined in NUREG-0737, Supplement No.- 1 Section 3.7 (Reference

.I 3), to meet NRC requirements and is', therefore, acceptable.

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'3.3.3.. Coolant Level in Reactor i

The applicant has committed to supply instrumentation for this variable. The proposed instrumentation is Category 2 with a range from the top of the vessel to the top of the core..This range is consistent with the. recommendation of Revision 3 of the regulatory guide (bottom of hot leg to top of vessel). We find this range acceptable.

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' The applicant takes exception to Regulatory Guide 1.97 with respect to the category of the instrumentation. This exception goes beyond the scope of this review and will'be addressed by the NRC as part of their review of NUREG-0737, Item II.F.2..

3.3.4' RCS Subcooling Monitor

-The applicant has identified this as a Type A variable. As such,.

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Table'2 of Regulatory Guide 1.97.recomends' Category 1 instrumentation.

The applicant is providing Category 2 instrumentation. The~ justification is that the operator would use the monitor to determine subcooling;

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however,'RCS Dressure and. temperature, in conjunction with a steam table, provides the same information. The NRC is reviewing'the acceptability of-this variable as part of their review of.NUREG-0737, Item II.F.2.

3.3.5 Containment Isolation Valve Position a

From the information provided, we find the applicant deviates from a strict interpretation of the Category 1 redundancy recomendation. Only

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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 th'e regulatory:

a guide. Position' indication of check valves is specifically excluded by Table 2 of Regulatory Guide 1.97. Therefore, we find that the instrumentation for this variable is acceptable.

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' 3.3.6: Radiation' Level in Circulating Primary Coolant

' l The applicant states that the post-accident sampling system can

- provide the required information with an itolated nuclear steam supply

' system. Main steamline and letdown'line monitors provide information when the' system is not isolated.

Based on the alternate instrumentation provided by the applicant, we

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1 conclu'de that the instrumentation supplied for this variable is adequate and, therefore,' acceptable.

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' 3.3.7 Containment Hydrogen Concentration j

Regulatory Guide.l.97 recomends instrumentation for this variable with a sensor capable of operating down to 10 psia. The applicant's sensor.

is capable of operating down to 0.8 atmosphere (ll.76 psia). Reference 8.-

providing information that supersedes.this information, states that-this instrumentation will remain functional' down to 8.0 psia. This satisfies Regulatory Guide'l.97 for this variable.

3.3.8 Residual Heat Removal (RHR) Heat Exchanger Outlet Temperature Revision 2 of Regulatory Guide 1.97 recommends a range of. 32.to i.

350'F. Revision 3 recomends a range'of 40 to 350'F. The applicant has supplied instrumentation with a range of 50 to'400*F. The applicant states, in Reference 8, that the RHR water temperature will~not be less I

than 50*F under post-accident conditions. Under normal c'onditions, the.RHR-i heat exchanger outlet temperature is not excepted to be less than 90*F.:

Based on the applicant's justification, we find. this range adequate'.to

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monitor this variable during all accident and post-accident conditions.-

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3.3.9 Accumulator Tank Level Regulatory Guide 1.97 reconinends Category 2 instrumentation for this

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variable with a range of 10 to 90 percent of the-accumulator volume. The j

applicant has Category 3 instrumentation for this variable that is used for

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backup information. The applicant states that adequate indication of,the 4

accumulator status is provided by the accumulator pressure variable and the position indication for the accumulator' discharge and vent valves. All of this indication is Category 2.

Section 6.3.5.4 of the FSAR indicates that each accumulator has two wide range level channels with control room indication and high and low level alarms. The applicant utilizes this as backup instrumentation, and as such, it is acceptable.

3.3.10 Accumulator Tank pressure Regulatory Guide 1.97 recommends instrumentation for this variable with a range of 0 to 750 psig. The applicant is providing instrumentation with a range of 0 to 700 psig. 700 psig is the design pressure of the accumulators and the same pressure at which the accumulator safety relief-valves operate, as identified in Table 6.3-1 of the FSAR. The applicant indicates that the normal operating pressure of the accumulator is manually controlled at 650 psig. Thus, there -is margin between the normal operating pressure and the setpoint of the safety relief valves. Based on.this, we find that the range supplied by the applicant for this variable'is acceptable.

3.3.11 Boric Acid Charging Flow l

Regulatory Guide 1.97 reconmends instrumentation for this variable.-

The applicant does not supply instrumentation for this variable as the j

emergency core cooling system is not supplied water from the boric acid tank. Should the boric acid tank be used for boration following an 9

I accident, the applicant demonstrates boration of the reactor coolant system l

by normal charging flow and reactor coolant system sampling.

1 The applicant does not have instrumentation for this variable and states that the units do not have. boric acid charging flow as.a safety' injection system. Centrifugal charging pump flow, safety injection f. low and residual heat removal flow are the safety injection variables

monitored. Therefore, we find that this variable.is not applicable at

.. Millstone 3.

3.3.12. Reactor Coolant Pump Status Regulatory Guide 1.97 rer~mnends instrumentation to monitor the reactor coolant-pump motor current. The appifcant did not identify instrumentation for this variable in their initial submittals. Reference 8 j

states that this variable is monitored by circuit breaker postion indication and pump motor current. We find this acceptable.

3.3.13 Pressurizer Heater Status' l

. Regulatory Guide 1.97 recommends Category 2 electric current instrumentation for this variable, to determine the operating status of.the heaters. The applicant has supplied circuit breaker position-indication for this variable.

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Section II.E.3.1 of NUREG-0737 requires a number of the pressurizer l

heaters to'have the capability of being powered by emergency power.

l Reference 8 notes that these heaters will have Category 2 instrumentation '

j added for these heaters. We find this committment acceptable-in meeting i

Regulatory Guide 1.97..

i 3.3.14 Steam Generator Pressure'

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safety valve setting (1185 psig). The applicant has supplied instrumentation.for this variable with a range of 0 to'1300 ' psia..The highest main steam safety relief valve setpoint is 1225 psig. The licensee-states that the total relief capacity of the safety valves.is greater than

..the normal. loop steam flow. Thus, the steam generator. pressure will not' exceed the range of the pressure instrumentation. Based on this, we. find the provided instrumentation acceptable.

3.3.15 Containment Spray Flow Regulatory Guide 1.97 recommends Category 2 instrumentation for this-variable. Millstone Unit 3 has two separate containment spray systems.

The quench spray system is initiated at' the beginning of an accident by the.

containment depressurization actuation (CDA) signal.. This system takes suction from the refueling water storage tank (RWST) which is also used as thewatersource'foremergencycorecoolingsystem(ECCS) pumps. This system does not have a long term function post-accident, as it does not switchover to the reactor building sumps for a water sourc'e. Category 3' flow transmitters are used for this variable. Backup instrumentation that shows operation of the nuench spray system includes RWST level..which 1s

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Category 1 instrumentation, system valve alignment, which is Category 2 irstrumentation, and containment pressure and temperature (before the recirculation spray system starts), which are Category-1 instrumentation.

Due to the short term operation of this system, and the backup-instrumentation available, we find this combination of instrumentation -

acceptable for the quench spray system.

Long term containment cooling is provided by the recirculation. spray.

system, which is automatically initiated at a fixed time after the CDA s ignal. Flow is from the reactor _ building sump; through the pumps and.

recirculation heat exchangers,. to the spray nozzles.. Category 2.

instrumentation monitors the flow through the~ recirculation heat exchangers.. This instrumentation'is acceptable.

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13.3.161 Heat Removal by the Containment Fan Heat Removal System Regulatory Guide 1.97 recomends ' plant specific instrumentation for i

this variable.. The applicant states that the containment air coolers are not used _in acc'ident and post-accident. conditions. Therefore,

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post-accident monitoring'is not provided. We find that this position.is acceptable for this variable.

3.3.17 Containment Sump Water Temperature a

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j The' applicant has not supplied this instrumentation.' The. applicant justifies this exception by st'ating that containment sump water temperature indication is not'used by the operator.to take'any type of corrective action and.it:is not used in any of the unit's emergency operating; j

procedures. The applicant states that the net position' suction head required for ECCS pump operation is assured regardless of sump temperature. Containment water level is measured along with containment -

temperature and pressure and recirculation spray flow._

This_is insufficient justification for this exception. The applicant-should provide the recommended instrumentation to be able to determine the-1 quantity of heat removed from the containment via the containment sump and-related systems.

3.3.18 Makeup Flow-In Regulatory Guide 1.97' recommends Category 2' instrumentation for this variable. The applicant does not have instrumentation that provides a direct measure of this variable. The makeup flow is' composed of the -

charging flow and the reactor coolant pump seal 11njection flow. The applicant monitors-these flows with Category 2; instrumentation. We find-_

this: acceptable.for this variable.

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3.3.19 ' Volume Control Tank Level Regulatory Guide 1.97 reconsnends instrumentation for this variable with a range from top-to bottom. The applicant has instrumentation for.

this variable that monitors only the cylindrical portion of the tank. The nonilinear portion ~of the tank is not measured. The existing range is adequate to monitor the operation of this variable during all accident and post-accident conditions, therefore, this deviation is acceptable.

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3.3.20 Component Cooling Water (CCW) Temperature to Engineered Safety Feature (ESF) System Components j

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Regulatory Guide 1.97-reconsnends Category 2 instrumentation for this i

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variable.. 'The applicant states that the service water system provioes the.

cooling function for the containment recirculation system heat exchangers, and the charging and safety injection pump lube oil' coolers cooling water systems. These are.the only ESF system components cooled. Category 2 j

instrumentation for the service water pump status, valve status and flow to j

the recirculation system heat exchanger is monitored. The applicant has not provided a readout for this variable in the control room..

The inlet temperature of the service water system is, by' design, based I

on the maximum temperature of the sea water taken from the'Long Island Sound. There is no control over the temperature of the service water.-

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= The justification submitted by the applicant for; this' exception is-adequate. The temperature of the cooling water to the ESF system.

l conponents will always be within the design range. Therefore..this is an

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1 acceptable exception from Regulatory Guide 1.97.

1 3'.3.21 High Level Radioactive Liquid Tank Level i

l Regulatory Guide 1.97 recommends instrumentation for this variable with' a range from the top to the bottom. The applicant's instrumentation 13-c._......

indicates from 9 inches above the lower hemispherical head to the upper hemispherical head. This range is stated to be adequate. High and low level alarms are provided.

The exist'ing low limit is adequate to indicate additional capacity to store radioactive liquids. Likewise, the upper limit is adequate to indicate the lack of additional capacity, and to show any inleakage to the tank. Based on this, we find this instrumentation adequate. Therefore, this is an acceptable deviation.

I 3.3.22 Radioactive Gas Holdup Tank Pressure j

Regulatory Guide 1.97 recommends instrumentation for this variable.

Millstone 3 does not have radioactive gas holdup tanks, using a charcoal delay system instead. Thus, instrumentation for this variable is not needed.

3.3.23 Containment or Purge Effluent-Vent Flow Rate Regulatory Guide 1.97 recomends instrumentation for this variable with a range of 0 to 110 percent of design flow. The applicant has instrumentation for this ventilation vent flow rate that is from 10,000 (4.3 percent of design flow) to 260,000 cubic feet per minute (113 percent ofdesignflow). We find the deviation (measuring down to 4.3 percent of design flow instead of 0 percent) insignificant, as the ventilation vent' flow would be higher than this for any one blower in operation. Therefore.

we find this oeviation acceptable.

3.3.24 M1 Other Identified Release Points Hydrogen recombiner cubicle ventilation-noble gas--The applicant has-instrumentation for this variable with a range of 7.1 x 10~4 to 6 vC1/cc. The applicant states that this is not a release point, but is i

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used to actuate closure of the hydrogen recombiner cubicle. Because this is not a' release point, instrumentation in conformance to Regulatory l

Guide 1.97 is not needed.

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gas--Regulatory Guide 1.97 recommends instrumentation for this variable.

l The applicant has grouped it under "all other identified release points,"

l which recommends a range of 10~0 to 10*2 uCi/cc. 'This recommended I

rar.ge. is for measuring releases from containment. For the steam exhaust, the range for vent from steam generator safety relief valves, ~10-I to 10 uCi/cc would be more appropriate. This is the range-supplied by the applicant.

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CONCLUSIONS

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J Based on our review, we find that~ the licensee either conforms to or 1

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is justified in deviating from Regulatory Guide 1.97, with the following j

exception:

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Containment sump water temperature--the applicant should supply the recommended instrumentation (Section 3.3.17).

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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 tmergency Response capability, NUREG-0/37 supplement No.1, NRC, Office of Nuclear Reactor Regulation, January 1983.

4.

Northeast Utilities letter, W. G. Counsil to D. G. Eisenhut, NRC,

" Supplement 1 to NUREG-0737, Requirements for Emergency Response Capability (Generic Letter No. 82-33)", April 15,1983, A02959.

5.

Northeast Utilities letter, W. G. Counsil to n. R. Denton, NRC,

" Application for an Operating License", February 2, 1983, 810671.

6.

Northeast Nuclear Energy Company letter, W. G. Counsil to Director of Nuclear Reactor Regulation, NRC, " Response to Question 420.06 ".,

December 16, 1983, A03541.

7.

Northeast Nuciear Energy Company letter, W. G. Counsil to Director of i

Nuclear Reactor Regulation, NRC, " Response to Question 420.6,"

J anuary 13, 1984, 811002.

8.

Northeast Utilities letter, J. F. Opeka to Director of Nuclear. Reactor Regulation, NRC, "Conformance to Regulatory Guide 1.97, Revision 2 Guidelines," May 28, 1985, A04668.

9.

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, NRC, Office of Nuclear Regulatory Research, May 1983.

10. Millstone Nuclear Power Station, Unit No. 3, _ Final Safety Analysis Report, Revision O.

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1 Division of Systems Integration Office of Nuclear Reactor Regulation Technical Evaluation Report i

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This EG&G Idaho, Inc. report reviews the submittals for the Millstone i

Nuclear Power Station, Unit No. 3, and identifies areas of nonconformance to Regulatory Guide 1.97. Any exceptions to the regulatory guide are evaluated and those areas where sufficient basis for acceptability is not provided are identified.

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