ML17254A240

Forwards Reg Guide 1.97,Rev 3 Instrumentation Comparison, Tabulating Info Provided in 840131 Submittal & 841130 Suppl 1 to NUREG-0737 SPDS Safety Analysis

ML17254A240
Person / Time
Site:	Ginna
Issue date:	02/28/1985
From:	Kober R ROCHESTER GAS & ELECTRIC CORP.
To:	Zwolinski J Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, RTR-REGGD-01.097, RTR-REGGD-1.097 NUDOCS 8503050463
Download: ML17254A240 (30)

Text

REGULATORY FORMATION DISTRIBUTION 'SY M (RIDS)

ACCESSION NBR:850305046$

-DOC ~ DATE: 85/02/28 NOTARI'ZED!',NO DOCKET' FACIL'50<<204 Rober t Emmet>> Gonna Nuclear ~,Plant< Unit li Rochester G.

05000204 AUTH',NAME~

AUTHOR AF F ILIAD'ION KOBERrR ~ H ~

Rochester Gas 8 Electric Corp ~

'RECIP ~ NAME.

RECIPIENT'FFILIATION Z~~OLINSKI'r J>>e A-o Operating Reactors-Branch 5

SUBJECTS Forwards Reg Guide 1~97~Rev 3 instr umention= compar$ honi tabul ating, info "-pr ovide'd in" 840131 submittal L 841130 'Suppl 1-to NUREG 0737 'SPDS safety analysis

~

DISTRIBUTION "CODE'! A003D COP/ES '-gECEIVED t LTR,

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-TITLE'! OR/Licensing, Submittal:

Suppl 1 to NUREG~0737(Generic Lkr 82+33)

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ID CODE/NAME~

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ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.K 14649-0001 ROGER Wi KO8ER VICE PRES!DENT ELECTRIC St STEAM PROOVCTION TELEPHONE APE*COOETIS 546-2700 February 28T 1985 Director of Nuclear Reactor Regulation Attention:,

Mr. John A. Zwolinski< Chief Operating Reactors Branch No.

5 U.S. Nuclear Regulatory Commission Washington<

D.C.

20555

Subject:

USNRC Regulatory Guide 1.97 R. E. Ginna Nuclear Power Plant Docket No. 50-244

Dear Mr. Zwolinski:

RGSE's letter dated January 31T 1984< provided a comparison of Regulatory Guide 1.97 requirements and existing instrumentation at the R.E. Ginna Nuclear Power Plant.

RGGE committed to submit a

schedule for meeting Regulatory Guide 1.97 recommendations or provide justification of differences in February 1985.

Attachment 1T "USNRC Regulatory Guide 1.97< Revision 3> Instrumentation Comparision"T tabulates information provided in the January 31/

1984 submittal and the NUREG-0737 Supplement 1 SPDS Safety Analysis dated November 30T 1984T and provides required schedules or justification.

V truly yoursT Roger W. Kober Attachment 85030504b3 850228 PDR ADOCK 05000244 F

PDR

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ATTACHNENT 1 USNRC Regulatory Guide 1.97< Revision 3

Instrumentation Comparison Table Rochester Gas and Electric Corporation R.E. Ginna Nuclear Power Plant Docket No. 50-244 February 1985

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REGULATORY GUIDE 1.97 INSTRUMENTATION COMPARISON Letters dated January 31<

1984 and November 30<

1984 provided< respectively>

a USNRC Regulatory Guide 1.97> Revision 2

comparison report and a plant safety analysis and implementation plan describing the basis on which parameters have been selected for display on the Ginna Safety Parameter Display System (SPDS).

This selection process utilized the emergency response guidelines developed by the Westinghouse Owners Group for writing emergency operating 'procedures,(EOPs) at Ginna Station.

In the attached table<

the required plant variables from Regulatory Guide 1.97'evision 3< are listed.

An identification as to whether or not the particular variable is required'to perform a safety functions as specified -in the EOPs> is noted in,,the>>;fifth column.

The evaluation provided in'he, November 30 letter>

as well as a review of the present Ginna-specific emergency operating proceduresr were used to generate this column. It should be noted that since the EOPs are still under active review and are subject to revision>

changes to the Regulatory Guide 1.97 comparison may occur in the future.

This table will be supplemented as appropriate when this occurs.

Variables not considered to be presently required at Ginna Station to perform safety functions required by the EOPs are considered acceptable as existing and are designated with an N/A in the last column.

Generally>

RGGE plans to meet the recommenda-tions of Regulatory Guide 1.97 for those items identified in the table on a schedule consistent with the implementation of the Safety Assessment System.

Where RGGE finds that acceptable implementation of the Regulatory Guide 1.97 guidance can be attained without meeting the explicit wording of Regulatory Guide 1.97>

the rationale is provided in the table.

It should be noted that differences from the recommended ranges specified in Regulatory Guide 1.97 are denoted in the fourth column.

These differences have been reviewed on an individual basis and found acceptable for plant-specific application.

Regulatory Guide 1.97 specifies provisions for redundancyi qualification<

QA> surveillance testing>

and human factors>

among others.

For environmental qualification purposes<

RGGE's submittal of October 31<

1980>

and subsequent NRC letters and RG6E responses form the basis for selection of equipment to be environmentally qualified (harsh environment) to meet 10 CFR 50.49.

For QA requirements>

RGGE's Appendix A to the QA manual uses a "graded QA" approach.

The Regulatory Guide 1.97 instrumentation will be reviewed and placed in the proper "grade" by the implementation date.

In terms of identification and human factors>

RG6E is now pursuing a Control Room Design Review for NUREG-0737.

Any enhancements in these areas will be addressed as part of that review.

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February 1985 ROCHESTER GAS 6 ELECTRIC USNRC REG.

GUIDE 1.97 REVISION 3 COMPARISON TABLE Variable Required Ran e

NRC Cate or Present Ginna Status Required for EOPs Schedule for Upgrade or Justification of Existin Confi uration Neutron Flux 10 to 100$ power Existing intermediate and source No range not qualified to Category 1

See Attachment 2, item 1, Control Rod Position RCS Soluble Boron Concent.

RCS Cold Leg Temperature RCS Hot Leg Temperature RCS Pressure Core Exit Temperature Full in or not full in 0-6000 ppm 50-700 F 50 700oF 0-3000 psig 200-2300 F

Existing R'vailable on PASS, System Range:

50-6000 ppm TW-450 and 451 are not qualified to Category I; indication on Main Control Board is by recorder TR-410 (range 0-700 F).

Kw No direct control room readout

. existing.

Have qualified hot leg temperature to subcooling meter.

3 PT-420A feeds PR-429 (0-3000 psig) and is Category 1;

PT-420 feeds PR-420 (0-3000 psig) but is not powered from Category I source.

Existing 39 chromel-alumel thermocouples positioned over outlet nozzles at selected core locations.

Range:

40-700oF at control room display, up to 1200oF at computer for all TCs, and up to 2200oF on computer for 5 TCs.

No No Yes Yes Yes Yes K/A See Attachment 2, Item 1

Category 1 TE-409B-1 and TE-410B-1 to be installed in 1985 (range 0-700oF).

Category 1 TE-409A-1 and TE-410A-1 to be installed in 1985 (range 0-700oF).

PT-420 power sources scheduled for upgrade following instal on of SAS/PPCS in 1986 Category 1 thermocouple system to be installed in 1985 (range 0-2300oF).

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Variable Required Ran e

NRC Cate or Present Ginna Status Required for EOPs February 1985 C

Schedule for Upgrade or Justification of Existin Confi uration Coolant Level in Bottom of hot leg 1

Reactor to top of vessel Not existing No A reactor vessel level monitoring system will be installed as des-cribed in RGB's letter on that subject of August 7, 1984.

Degrees of Sub-cooling Containment Sump Narrow Range Containment Sump Wide Range Containment Pressure Containment Isolation Valve Position Radioactivity Concentration or Radiation I.evel in Circulating Primary Coolant 200~F sub-cooling to 35 F superheat Plant specific Plant specific 10 psia to 3 times design pressure Closed/not closed 1/2 to 100 times Tech.

Spec. limit R/hr

Existing, (range:

0-100 F subcooling) 2

Existing, Sump A:

LT-2039 and LT-2044 (range:

0-30 ft.)

Existing, Sump B:

LC-942(A-E) and I,C-943(A-E) indication of 8, 78,113,

180, 214 inches (214 inches

= approx.

500,000 gal.

which was previously justified and accepted by NRC).

Qualified Category 1.

1

Existing, PT-946 8 948 (10-200 psia)

Existing, status lights on MCB and backlighted push-bottons on CI matrix panels 1

" Available with PASS system.

Yes No Yes Yes No No See Attachment 2, item 2 N/A N/A N/A See Attachment 2, item 3 See Attachment 2, item 1

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February 1985 Variable Required Ran e

NRC Cate or Present Ginna Status Required for EOPs Schedule for Upgrade or Justification of Existin Confi uration Analysis of Primary Coolant Containment Area Radiation 10 to 10 Ci/gm or TID-14844 source term in coolant volume 1 to 10 R/hr 7

Existing capability Existing No Yes N/A N/A Radiation 10 to 10 R/hr 2

4 Exposure Rate (areas where access required to service equipment)

Radiation meters provided (or will be provided) where post-accident operator access required.

Ranges meet-. dose rate calculated using TID 14844 source term. - Based on single failure consideratio'ns no access required to service equipment.

No N/A Effluent Radio-activity-Noble Gas:

-Condenser Air Ejector Exhaust

-Containment Purge Vent Exhaust 10 to '10 uCi/cc 10 to 10 uCi/cc Existing Existing No No N/A N/A

-Plant Bldg.

Exhaust Vent 10 to 10 uCi/cc Existing No N/A

-Vent from S/G Safety Relief 8 Atmospheric Dump Valves 10 to 10 uCi/cc Existing Yes N/A

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February 19.85 Variable Required Ran e

NRC Cate or Present Ginna Status Required

-for EOPs Schedule for Upgrade or Justification of Existin Confi uration Effluent Radio-activity Particulates and Halogens.

Sampling with Onsite Analysis Capability:

-Containment Purge Vent Exhaust 10 to 10 uCi/cc Existing N/A

-Aux. Bldg.

Vent Exhaust 10 to 10 uCi/cc Existing No N/A Containment Hydrogen Concentration 0-10$

Existing Yes N/A Radiation Exposure Rate in Areas Adjacent to Containment 10 to 10 R/hr Area monitors existing in Aux. Bldg. agd Int.1Bldg.

(range:

10 to 10 R/hr)

'-No N/A RHR System Plow RHR Heat Exchanger Out-let Temperature Accumulator Tank Level and Pressure 0-110/ design 40-350~F 10-90/ volume 2

0-750 psig Existing, FT-626 (0-4000 gpm)

(Not redundant.

SI flow or RCS pressure is used as backup indication)

Existing, TE-627 to computer (range:

0-310 7)

Narrow range instrument indicates

+7 inches from normal filllevel for accurate Tech.

Spec.

compliance; 0-800 psig pressure Yes No No N/A N/A N/A

February 1985 Variable Accumulator Isol. Valve Position Required Ran e

Closed or open NRC Cate o

2 Present Ginna Status

Existing, MOV 841 8 865 position indicated on MCB Required for EOPs No Schedule for Upgrade or Justification of Existin Confi uration N/A Charging Flow SI Flow RUST Ievel Reactor Coolant Pump Status Primary System Safety Relief Valve Positions (PORV's 6 Code Safeties) 0-110/ design 0-110'/ design Top to bottom Electric current Closed/not closed 2

1 s4 Existing FT-128 (0-75 gpm, the maximum flow anticipated in normal operations)

Existing FT-924 6 925 (0-1000 gpm)

Existing, LT-920 S 921 (0-100/)

Ammeter existing at 4KV Bus (0-1200A)

Existing No Yes Yes No Yes N/A N/A N/A N/A N/A Pressurizer I,evel Bottom to top 1 -=

Existing Yes N/A Pressurizer Heater Status Quench Tank Level Quench Tank Temp.

Electric current Top to bottom 3

50-750~F For control group of heaters, ammeter existing in Aux. Bldg.

No ammeter for backup group, but have breaker position for both control and backup groups in Control Room.

Existing, LT-442 (0-100/)

Existing, TE-439 (0-300~F)

No No No N/A N/A N/A N

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February 1985 Variable Quench Tank Pressure S/G Ievel S/G Pressure Safety/Relief Valve Positions or Main Steam Flow Required Ran e

0 to design pressure Tubesheet to separators-=

From atmos-pheric press.

to 20/ above lowest safety valve setting (1300 psig)

Closed/not closed NRC Cate or 1

(flow only)

Present Ginna Status Existing, PT-440 (0-150 psig)

Existing, LT-460 6 470 input to LR-460 on MCB (0-518" H 0) 2 Existing, PT-468,

469, 478, 479 (range:

0-1400 psig)

Existing, main steam flow FT-464, 46$, 474, 475 (range:

0-3.8 x 10 pph).

Safety/

relief valve positions only during high radiation in secondary sys.

Required for EOPs No Yes Yes Yes Schedule for Upgrade or Justification of Existin Confi uration N/A N/A N/A N/A Main Feedwater Flow Aux..Feedwater Flow 0-110/ design 3

0-110% design 1

flow Existing, FT-466, 4(7, 476, 477 (range:

0-3.8 x 10 pph)

Existing No Yes N/A N/A Condensate Storage Tank Plant Specific 1

CST Transmitters LT-2022A and LT-2022B are qualified; read 0-24 ft.

Yes N/A Containment Spray 0-110/ design 2

Flow flow Containment spray flow itself is not available,

however, SI, RHR, and total flow are available -

CS flow can be determined.

No N/A

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February 1985 Variable Required Ran e

NRC Cate o

Present Ginna Status Required for EOPs Schedule for Upgrade or Justification of Existin Confi uration CV Fan Heat Removal CV Atmosphere Temp.

CV Sump Mater Temp Plant Specific 2

0 400oF 50-250~F CV fan 1A, 1B, 1C, 1D on/off status at MCB, plenum exhaust temp.

24 CV RTD's go to Leak Rate Test Panel; range 40-130~F Also, RTD's in plenum exhaust read 0-6004F Not existing No No No N/A N/A N/A-Letdown Flow 0-110$ design 2

Volume Control top to bottom 2

Tank Level FT-134 (0-100 gpm)

LT-112 (0-100/)

No No N/A.

N/A Component Cooling 40-200~F Mater Temp. to ESF TE-621 from CCW Hx goes-to computer (50-200~F)

No N/A High Level Radioactive Tank Level top to bottom 3

L1001 (0-100%)

- waste holdup tank No N/A Radioactive Gas 0 to 150/ design 3

Holdup Tank Pressure Emergency Vent Open/close Damper Position status

P1036, 1037,

1038, 1039 (0-150 psig) design pressure 150 psig, normal operation 100-110 psig.

Existing for containment vent on MCB No No N/A N/A

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February 1985 Variable Required Ran e

NRC Cate or Present Ginna Status Required for EOPs Schedule for Upgrade or Justification of Existin Confi uration Status of Standby Power and Other Energy Sources Important to Safety (hydraulic, pneumatic):

-480 V Bus Voltage, current 2

pressure Existing diesel voltmeters S

ammeters on MCB No N/A

-Instrument Bus Existing voltmeters on panels in control room; ammeters on inverters in battery rooms for bus lA 8 1C No N/A

-125 VDC Bus Existing voltmeters and ammeters in Control Room No N/A Radiation Expo-sure Meters (continuous indication at fixed locations)

Range, location, 2

and qualifi-cation criteria to be developed to satisfy NUREG-0654,Section II.H.5b and 6b requirements for emergency radiological monitors Existing procedures and equipment are used to initiate emergency measures in accordance with Appendix I and II.H.5b and 6b of NUREG 0654 and NRC approved plant Technical Specifications for compliance with 10 CFR 50 Appendix I.

No N/A Airborne Radio-halogens and Particulates (portable sampling with onsite analysis capability) 10 to 10 uCi/cc Existing No N/A

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February 1985 Variable Plant and Environs Radi-ation (portable instrumentation)

Plant and Environs Radioactivity (portable instrumentation)

Wind Direction Wind Speed Required Ran e

10 to 10 R/hg, photon 10 to 10 rads/hr, beta radiations and low energy photons Multichannel gamma ray spectrometer 0-360 0-67 mph NRC Cate or

-.Present Qiana Status 3

Existing Existing Existing---

Existing at 33,

150, 250 ft.

elevations (range:

0-100 mph)

Required for EOPs No No No No Schedule for Upgrade or Justification of Existin Confi uration N/A N/A N/A N/A Estimation of Atmospheric Stability Based on vertical temperatures differences "Existing, 2 RTD's at 33, 150, 250 ft. elevations; delta T

- between each elevation No N/A

'Primary Coolant and Sump:

-Gross Activity

-Gamma Spectrum

-Boron Content

-Chloride Content Grab Sample 10 to 10 Ci/cc Isotopic Analysis 0-6000 ppm 0-20 ppm Available with PASS system Existing (50-6000 ppm) with PASS 5 ppb - 100 ppm lab analysis No No No No N/A N/A N/A N/A

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February 1&85 Variable

-Dissolved Hydrogen

-Dissolved Oxygen Required Ran e 0-2000 cc(STP)/Kg 0-20 ppm NRC Cate o

Present Ginna Status (10-2000 cc/Kg) with PASS (0.1 20 ppm) with PASS Required for EOPs No No Schedule for Upgrade or Justification of Existin Confi uration N/A N/A

-pH Containment Air:

-Hydrogen Content 1-13 Grab Sample 0-10/

1-13 with PASS Available with PASS No No Available with H2 monitors and PASS No N/A N/A N/A

-Oxygen Content 0-30$

-Gamma Spectrum Isotopic Analysis 0-30'/, with PASS Existing No No N/A N/A I

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Attachment 2:

Explanatory Information Item 1:

Source Range Flux In response to the issue of inadvertent boron dilution~

as addressed in SEP Topic XV-10 (NRC SER dated September 4<

1981)<

RG6E stated that operator actions would be used to terminate the dilution event.

Note that dilution accidents are addressed by "abnormal"<

rather than "emergency" operating procedures.

If all source range flux instrumentation becomes inoperable<

RGGE plans to implement a procedural step which would direct the operator to immediately prevent the possibility of a dilution event (e.g.>

secure the volume control tank suction and align the charging pump to the RWST).

Plant staff would also be directed to take samples of the RCS to determine the boric acid concentration.

The details of the procedure changes are not yet complete; however<

the following is anticipated:

When on RHR< sampling will be done at 1/2 hour intervals until the reactor is 5% delta rho subcritical<

and at 2

hour intervals thereafter until the source range monitor is restored.

Under other applicable circumstances (refueling> startup>

shutdown) the sampling interval will be one hour until the reactor is 5% delta rho subcritical and two hours thereafter.

With such a procedure in place<

RGGE has concluded that the source range flux monitor will not be needed to fulfillthe guidance of Regulatory Guide 1.97.

Boric acid "(and other par'ameters<

such as coolant activity) can be de'termined using the Ginna Post Accident Sampling System (PASS).

The Post Accident Sampling System has been designed and installed as discussed in RGSE's resolution of NUREG 0737 Item II.B.3.

Item 2:

Degree of Subcooling The Ginna Emergency Procedures direct the operator to determine subcooling margin by calculating the degree of subcooling using available instrumentation (RC wide-range pressure

[0-3000 psig] and core exit thermocouple temperature

[0-2300 F]).

Thus<

there is no practical instrumentation limit on the available range of sub-cooling which can be determined.

The actual subcooling meter installed separately at Ginna uses RC wide-range pressure and hot leg temperature to calculate the degree 0

of subcooling.

This meter has a range of 0-100 F

subcooling<

compared to the Regulatory Guide 1.97 recommendation of 35 superheat to 200 F subcooling.

0 0

Since this is a backup instrument<

RGGE does not consider this difference to be significant.

Item 3:

Containment Isolation Valve Position Indication As previously discussed in RGGE's submittals regarding "Environmental Qualification of Electrical Equipment 10 CFR 50.49<"

RGSE has determined that position indication is not a parameter required for the performance of a safety function.

As discussed in detail in RGSE's responses to IE Bulletin 78-04 (Narch 3> 1978) and Inspection No. 78-27 (January 16>

1979)<

the valve safety functions are accomplished using control circuits not affected by any postulated failure of the limit switches which provide the position indication function.

Further>

because of equipment redundancy<

failure of limit switches would not negate the accomplishment of any safety function.

Thus>

RG6E has designated these indications as not required by the Emergency Operating Procedures>

but only provide useful backup information.

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