ML20235V481

Forwards Complete Safety Evaluation on TMI Action Item II.F.2.First Page Inadvertently Omitted from 870910 Rept

ML20235V481
Person / Time
Site:	Crystal River
Issue date:	09/30/1987
From:	Silver H Office of Nuclear Reactor Regulation
To:	Wilgus W FLORIDA POWER CORP.
References
TASK-2.F.2, TASK-TM NUDOCS 8710150182
Download: ML20235V481 (7)

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l Septeinber 30, 1987 3,.

' Docket: No. 50-302 :

DISTRIBUTION'

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NRC & Local PDRs G. Schwenk' t

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,Mri W. S. Wilgus-PD22. Reading.

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,7Vice'. President,'NuclearL0perations S. Varga

, Florida Power Corporation.

JG. Lainas=

. ATTN: Manager, Nuclear Licensing D. Miller

, (P.O. Box 219 H. Silver-M Crystal River, Florida' 32629'

' OGC-Bethesda.

LE. Jordan.

Dear Mr.;Wilgus:

J. ' Partlow

' SUBJECTi 'TMI: ACTION ITEM II.F.2 - CRYSTAL RIVER UNITt3

0niSeptember 10, 1987, the'NRC staffLissued a-safety' evaluation'on

the above

~

. subject. Due to an administrative error, the first page of the safety evalu--

.l ation.was; inadvertently omitted~from the reproduced copies transmitted.tofthe service. list.

Enclosed.is'a-copy of the safety evaluation in'its entirety.

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Sincerely, j

p-1 Harley. Silver, Project Marager j

Project Directorate II-?'

Division of Reactor _ Projects-I/II 1

Enclosure:

.As' stated cc w/ enclosure-See next page'-

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Mr.W.lS.Wilgus Crystal River Unit No. 3 Nuclear

Florida Power Corporation '

Generating Plant cci l

Mr. R..W. Neiser-State Planning and Development o

-Senior Vice President Clearinghouse i

,s and General. Counsel Office of Planning and Budget-t

Florida--Power Corporation Executive Office of,the Governor

~P. 0. Box 14042 The Capitol Building 1

St. Petersburg, Florida -33733 Tallahassee, Florida - 323011 Mr. P.~F. McKee Mr. F. Alex Griffin, Chairman Director,~ Nuclear Plant' Operations Board of County Commissioners Florida Power Corporation Citrus County P. O. Box 219 110 North Apopka Avenue l

-Crystal River, Florida 3262.9 Inverness, Florida- '36250 Mr. Robert B. Borsum:

Mr. E.'C. Simpson

-Babcock &'Wilcox Director, Nuclear Site.

Nuclear Power Generation Division Florida Power Corporation Support Suite 220, 7910.Woodmont Avenue P.O.-Box 219-Bethesda,. Maryland 20814 Crystal River, Florida 32629 Resident Inspector-U.S. Nuclear. Regulatory Commission 15760 West Powerline Street Crystal River, Florida 32629 Regiona1' Administrator, Region II U.S. Nuclear. Regulatory Commission j

101 Marietta Street N.W., Suite 2900 i

Atlanta, Georgia 30323 Jacob Daniel Nash Office of Radiation Control

' Department of Health and

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-Rehabilitative _ Services 1317 Winewood Blvd.

Tallahassee, Florida 32399-0700 Administrator Department of Environmental Regulation

. Power Plant Siting Section State of Florida 2600 Blair Stone Road Tallahassee, Florida 32301 Attorney General Department of Legal Affairs i

The Capitol Tallahassee, Florida 32304 i

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. UNITED STATES i

g NUCLEAR REGULATORY COMMISSION J

j WASHINGTON, D. C. 20555

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ENCLOSURE'l

. SAFETY EVALUATION BY THE OFFICE OF NUCLEAR' REACTOR REGULATION.

INADE00 ATE CORE COOLING INSTRUMENTATION SYSTEM FLORIDA POWER CORPORATION CRYSTAL RIVER PLANT DOCKET NO. 50-302

1.0 INTRODUCTION

In response'to the USNRC Order for. Modification of License entitled, " Inadequate Core Cooling Instrumentation Systems" dated December 10, 1989, Florida Power Corporation (FPC) has proposed a system (Ref. 1-1?) for detecting and monitoring inadequate core' cooling (ICC)l conditions, including subcooling margin monitoring' (SMM), core exit thermocouple (CET), a reactor coolant inventory tracking system (RCITS), and a reactor coolant pump void trend monitoring system (RCPVTM) for the Crystal' River Unit 3 Nuclear Power Plant. This staff. review incorporates significant. technical input (Ref.13) prepared by the staff's consultants at.

Oak Ridge National Laboratory (ORNL).

2.0 EVALUATION 1

'2.I' Subcooling Margin Monitoring (SMMI Syftem i

A saturation margin monitoring system, meeting the requirements of NUREG-0737,

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is installed and operational.

The system. consists of two independent monitoring channels with displays located on the main control board.

The instruments use hot leg (T ) and cold leg (T temperatures and the reactor Tressureateachloopto'cNiculatethecorrefp)ondingsaturationtemperature

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'(Tsat).

Each monitor displays the degrees of subcooling of one loop on

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.'dema nd. A pressure-temperature' display has been added to the safety parameter display system (SPDS) which depicts the saturation curve as 'part of the overall SPDS.

1 In addition to T and T. the hottest CET selected from a group of.six is 1

, connected to eacNSMMcfia,nnel. These I? CETs have been selected to provide j

representative temperatures' from each core quadrant and the central region.

1 The hottest CET temperature over a range of O'F to 1,023'F is displayed on 1

demand. Tsat corresponding to the hottest CET also is displayed on demand.

J The instruments include a low margin-to-saturation alarm indicator and an L

alann signal to the plant annunciator and event recorder.

Verification of S*

operability is performed semiannually by performing a calibration using established procedures.

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I The SMMs were tested for and meet the requirements for seismic and environmental qualification.

The redundant channels are powered from separate Class IE power sources.

Suitable buffering is provided between the plant instrumentation i

sensors and the SMM channels.

Each SMM is capable of being manualTy switched to either coolant loop to improve overall availability.

2.2 Core Exit Thermocouple (CET) System l

The core exit thermocouple system consists of a total of 52 CETs distributed throughout the core. The primary operator display is provided by a core map diagram on the plant computer monitor. The temperature at each of the 51 CET locations is displayed and bad readings are identified.

The highest temperature of all operable CETs is highlighted in color and is updated i

continuously when the core map is displayed.

A hard copy of the core map is printed on demand by the computer line printer.

The operable display range is 0*F to 2,500'F. The computer provides a capability to display or print a temperature-time digital trend history for any CET, singly or in groups, by operator selection.

In addition, four analog strip chart recorders are driven by the computer and each can provide a trend record of any CET. Any CET exceeding 700"F will be alarmed on a dedicated alarm monitor and printer and the alarmed points will be indicated in color on the core map.

The CET backup display consists of three multi-pen analog temperature recorders located on the main control board. A minimum of 16 CETs, 4 from each core quadrant, will be recorded continuously over a range of 0"F to 2,500"F. The recorders have alarm capability, can automatically change chart speed on alarm, and will detect inoperable CETs.

i In addition, each SMM channel can display on demand the hottest CET selected from a group of six CETs. A total of 12 CETs have been selected to provide representative temperatures from each core quadrant and the control region.

These instruments display temperature over a range of 0 F to 1,023'F.

The primary and backup display channels are electrically independent, energized from independent power sources, and physically separated in accordance with Regulatory Guide 1.75 up to and including the isolators.

The primary display and computers are not Class 1E. All CETs connected to both safety and nonsafety systems are isolated prior to the connection with nonsafety systems.

CETs connected only to the plant computer (nonsafety) are not isolated. The incore probe assemblies and all cables and connectors associated with the CET system have been replaced with qualified units.

2.3 Reactor Coolant Inventory Tracking System IRCITS1 The reactor coolant inventory tracking system uses differential pressure (DPi measurements across vertical elevations of the hot leg and the reactor vessel to infer coolant level when the reactor coolant pumps are tripped. The wide range measurements are from the top of each hot leg / candy cane) to the bottom of the hot legs. The narrow range measurements are from the top of the reactor I

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vessel head to the bottom of the hot legs.

The total of four DP transmitters are paired for redundancy with one wide range and one narrow range transmitter in each channel.

The channels are independently powered by Class 1E instruments-tion power.

The transmitters are located inside containment and se~al chambers 1

are located at the high point of each reference leg to keep the legs full of-water. The lower pressure taps are located on a common decay heat suction line.

Physical precautions have been taken to minimize the vulnerability of the single lower connection.

Density compensation is employed to correct for temperature effects on the reference leg and process liquid density. A section of the tubing between the vessel top tap and the refueling cavity wall is i

removable for refueling.

Removable seismic supports are provided. The DP transmitters are Class IE and qualified for the containment environment.

1 The narrow range transmitters are calibrated for approximately 12 feet of water which, when compensated for system temperature variations, will be equivalent to the level of the coolant in the reactor vessel, above the bottom of the hot leg, when the RCPs are tripped.

The wide range transmitters are d

calibrated for approximately 50 feet of water, which is equivalent to the level of coolant within the hot leg when the RCPs are tripped.

The DP measurements are not functional when the RCPs are running or during venting operations.

RTDs on the vertical portions of the reference legs and in the hot legs provide the temperature measurements for appropriate compensation of level indications.

Two independently-powered electronic analog equipment racks are used to power the DP transmitters and process the outputs to compute coolant level. The output signals are sent to analog indicators located on a panel in the control

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• ~ room and isolated signals are supplied to the plant computer.

The coolant level indicators will read off-scale high when the RCPs are running and operational procedures will instruct the operators that the level indications k

are invalid under these conditions.

An error analysis of the system predicts measurement uncertainty of +/- 3.26%

for the plant in normal conditions.

For the accident case this uncertainty is

+/- 7.84%, taking into account the system errors under accident conditions.

1 Additional uncertainty may arise during inadequate core cooling conditions due j

to turbulence in the reactor coolant system. This error cannot be calculated and there are no tests to provide validity, but is expected to be in the order of 10 to 20%.

The RCITS is installed and has been functionally tested and calibrated.

2.4 Reactor Coolant Pump Void Trend Monitoring (RCPVTM1 System i

The RCPVTM system provides a monitor of reactor coolant inventory with the RCPs running by measuring RCP motor current to infer the density of the pumped fluid. The system uses pump inlet temperature in an algorithm with the pump current measurements to derive an estimate of the pumped fluid void fraction.

Existing current transformers (non-Class IE) and RTDs (one each per pump 1 are i

, 'used'to provide. pump current and pump. inlet temperature signals to a computer.

The: computer calculates _the corresponding saturated liquid and vapor densities for each temperature input' and combines the densities with the pump current in

accordance with the void fraction algorithm.. Analog. indicators in the control room provide Vindication of the void fraction for any single pump,- or. the average Void fraction for all: pumps running, over a: range of 15 to 40 percent-void. fraction. The indicators are located in close proximity with the RCITS indicators.

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3.0 CONCLUSION

- The staff has reviewed the final design of the Crystal River Unit 3-ICC monitoring system and finds it.to be acceptable. The system design features, including qualification, redundancy,' display,-location, response accuracy, j

' etc;, all satisfy the requirements of the NUREG-0737.

4.0 REFERENCES

.1..

Letter from USNRC to FPC, D. G. Eisenhut.to J. A. Hancock, dated December 10,.198?.

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Letter from FPC to USNRC, 3F-0483-11, G. R. Westafer to H. R. Denton, dated April ".1983.

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Letter from FPC to USNRC, 3F-0483-19, G. R. Westafer to H. R. Denton',

l dated April 25, 1983.

4.

Letter from FPC to USNRC, 3F-0783-15, G. R. Westafer to H. R. Denton, dated' July 18, 1983.

5.

Letter from FPC to USNRC, 3F-1083-09, E. C. Simpson to J. F. Stolz, dated October 7, 1983.

6.

Letter from FPC to USNRC, 3F-0284-07, G. R. Westafer to J. F. Stolz, dated February 15, 1984.

7.

Letter from FPC to USNRC, 3F-0884-16, G. R. Westafer to J. F. Stolz, dated August 31, 1984 8.

Letter from FPC to USNRC, 3F-0285-01, G. R. Westafer to J. F. Stolz, dated February 1, 1985.

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FPC, Specifications for Electric Transmitter SP-5061 R2, dated l

February 13, 1984.

10. Letter from FPC to USNRC, 3F-1085-12, G. R. Westafer to J. F. Stolz, dated October 23, 1985.

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Letter from FPC to llSNRC, 3F-1186-?3, E. C. Simpson to J. F. Stolz, dated

. November 24, 1986.

'12..'FPC Calculations - ICC' 14235.13-K-02 Rev. ?, dated November 28, 1986.-

.13.. Letter.from'J. L. Anderson (ORNL) to G. A. Schwenk (NRC) dated June 15,

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1987.

Dated:; September.10 1987:

1l Principal Contributor: '

.G.'Schwenk-i l

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