ML090720825

NUREG-0053, Suppl. 8, Safety Evaluation Report Related to Operation of North Anna Power Station, Unit 1 and 2

ML090720825
Person / Time
Site:	North Anna
Issue date:	12/31/1977
From:	Office of Nuclear Reactor Regulation, Virginia Electric & Power Co (VEPCO)
To:
References
FOIA-2024-00060 NUREG-0053 S8
Download: ML090720825 (64)

Text

SUPPLENENT NO.8 TO THE SAFETY EVALUATION REPORT BY T'clE OFFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULA'I'ORY COMMISSION IN THE MATTER OF VIRGll'ilIA ELECTRIC Al'ID POVJER COMPA.~

NORm ANNA POWER STATION -

UNITS 1 AND 2 DOCKET NOS. 50-338 AND 50-339 NUREG-0053 Supp1emen t No. 8 December 14, 1977

TABLE OF CONTENTS PAGE

1.0 INTRODUCTION

AND GENERAL DISCUSSION *..*.*.*..********.****..*.**. 1-1 1.1 Introduction *.,~ **** e.e ****** Q.e ***** $ *** o ** o ** IiI **** e ** ****** 1-1 2.0 SITE CHARACTERISTICS ***.*.***.***.*.*.***.*.*.*****.******.***.** 2-1 2.6 Foundation Engineering ********..*...****.***.********.****** 2-1 2.6.2 Evalu~tion of Foundation Engineering *****..***.***.** 2-1 3.0 DESIGN CRITERIA - STRUCTURES. SYSTEMS. AND COMPONENTS.*****.**..* 3-1 3.10 Seismic and Environmental Qualification of Seismic Category I Instrumentation and Electrical Equipment *.*.***** 3-1 3.10.1 Seismic Qualification of Seismic Category I Instrumentation and Electrical Equipment ******.****** 3-1 3.10.2 Environmental Qualification of Seismic Category I Instrumentation and Electrical Equipment For Main Steam Line Break Accidents ******** 3-1 3.10.3 Environmental Qualification of Westinghouse and Ba-Iance-of-Plant Seismic Category I Instrumentation and Electrical Equipment ************* 3-4 6.0 ENGINEEREu SAFETY FEATURES *********.*.*******.******************* 6-1 6.2 Containment Systems.******************************.********* 6-1 6.2.1 Containment Functional Design *******************.**** 6-1 6.2.2 Containment Heat Removal System **********.****.***.** 6-2 i

9.0 6.3 PAGE Emergency Core Cooling System ****************************** 6-10 6.3.3 6.3.4 System Performance Evaluation *********************** 6-10 Tests and Inspections.................*............. 6-12 AUXILIARY S,¥STE£VS ******* *****,. ***************** '" 0

• ** 9-1 9.1 Fuel Storage and Handling ********************************** 9-1 9.1.4 Fuel Handling Systems ******************************. 9-1 9.2 water Systerns *********** o ***** e ********** o ********* Go ******* 9-2 9.2.1 Service \\I~ater System ********************* ' *********** 9-2 18.0 REVIEW BY THE ADVISORY COHi'U'rrEE OJ'l REACTOH SAFEGUARDS. * * * * * * * ** 18-1 22.0 COt\\lCLUSIOl\\lS............................ ".......................... 22-1 ii

APPENDIX A APPENDIX B APPENDIX C APPENDIX [)

APPENDIX E APPENDIX F APPENDIX G APPENDICES CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW ERRATA TO SUPPLEMENT NO. 7 TO THE SAFETY EVALUATION REPORT FOR THE NORTH ANNA POWER STATION, UNITS 1 AND 2 "CAVITAnON PROBLEMS FOR NORTH ANNA REACTOR RECIRCULATION SPRAY AND LOW rlEAD SAFETY INJECTION PU~11PS" tjY FREDERICK G. HANMITT, PROFESSOR-IN-CHARGE, CAVITATION AND MULTIPLE FLOW LABORATORY -

DEPARTMENT OF MECHANICAL ENGINEERING UNIVERSITY OF MICHIGAN ADVISORY COMMITTEE ON REACTOR SAFEGUARDS' LETTER DATED JULY 20, 1977 STAFF'S i"lEI'10RANtJLHvJ DATED AUGUST 1, 1977 ADVISORY COjvUvlITTEE ON REACTOR SAFEGUARD'S LETTER DATED NOVElvlBER 10, 1977 STAFF'S LETTER DATED DECEMBER 2, 1977 iii PAGE A-l B-1 C-1 0-1 E-1 F -1 G-l

1.0 IN'I'RODUC'rION AND GENERAL DISCUSSI01~

1.1 Introduction On June 4, 1976, the Nuclear Regulatory corrunission (Commission) issued its Safety Evaluation Report regarding the application by the Virginia Electric and Power Company (applicant) for licenses to operate the North Anna Power Station, units 1 and 2 (North Anna facility).

Supplement No.1 to the Safety Evaluation Report was issued on June 30, 1976; Supplement No.2 was issued on August 2, 1976; Supplement No.3 was issued on September 15, 1976; Supplement No. 4 was issued on December 8, 1976; Supplement No.5 vvas issued on December 29, 1976; Supplement No.6 was issued on February 2,1977, and Supplement No.7 was issued on August 18,1977. Supplement NOs. 1 through 7 to the Safety Evaluation Report documented the resolution of several outstanding items, and summarized the status of the remaining outstanding issues.

The purpose of this supplement is to update our Safety Evaluation Report (and Supplement Nos. 1 through 7) by providing (1) our evaluation of additional information submitted by the applicant since the issuance of Supplement No.7 to the Safety Evaluation Report, and (2) our evaluation of additional information for those sections of the Safety Evaluation Report where further discussion or changes are in order.

1-1

Each section of this supplement is n~~red the same as the section of the Safety Evaluation Report, and is supplementary to and not in lieu of the discussion in the Safety Evaluation Report and the supplements thereto, except where specifically so noted. Appendix A is a contin-uation of the chronology of our principal actions related to the processing of the application, and Appendix B is a listing of errata to Supplement No. 7 to the Safety Evaluation Report. Appendix C is a copy of our consultant's (Professor Frederick G. Hammitt of ~~e University of Michigan) report concerning the "Cavitation Problems For North Anna Reactor Recirculation Spray and Low Head Safety Injection Pumps." Appendix D is a copy of a letter dated July 20, 1977 from the Advisory Committee on Reactor Safeguards concerning reopening the North Anna Power Station units 1 and 2 review.

Appendix E is a copy of a memorandum dated August 1, 1977 to the Advisory Committee on Reactor Safeguards in response to the Con~ittee's suggestion that we affirm that the hydrology of the site is under control. Appendix F is a copy of a letter-dated November 10, 1977 from the Advisory Committee on Reactor Safeguards concerning modi-fication of the recirculation and quench spray systems. Our response to tne C~TIffiittee concerning this matter is presented in a letter dated December 2, 1977 a copy of which is attached as Appendix G.

A summary of the remaining outstanding issue is presented in Section 22.0 of this supplement.

1-2

2.6 2.6.2 2.0 SITE CHARACTERISTICS Foundation_Engineering Evaluation of FOundation Engineering In Section 2.6.2 of Supplement No. 7 to the Safety Evaluation Report, we stated that we intend to condition the operating license of unit 1 to require that the plant be shut down December 31, 1977, if the goundwater control system is not completed and operational.

In a letter dated November 8, 1977, the applicant advised us that the groundwater control system has been completed and the system is opera-tional.

We have confirmed that the groundwater control system, that we previously reviewed and approved, has been installed and is operational and, therefore, have concluded that a condition to the operating license for unit 1 regarding this matter is no longer necessary. Therefore, we consider this matter resolved.

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3.10 3.10.1 3.10.2 3

• 0 DESIGN CRI'rERIA-STRUCTURES, SYSTEMS, Al.\\lD COMPONENTS Seismic and Environmental Qualification of Seismic Category I Instrumentation and Electr ical EquipII1§;nt Seismic Qualification of Seismic Category I Instrumentation and Electrlcal Equi~nt In Section 3.10 of the Safety Evaluation Report we stated that we would provide the results of our evaluation of the seismic qualification program which would include both westinghouse supplied and the balance-of-plant seismic Category I* instrumentation and electrical equipment.

Our review of the applicant's seismic qualification prograrn consisted of reviewing test methods, procedures, documentation of test results, seismic input parameters such as arnplitude, duration, frequency content and directional considerations, inspection of test facilities, and site visits, and included requiring the applicant to retest some equipment and components where our evaluation considered this to be necessary.

On the basis of our review, we have concluded that the seismic quali-fication of seismic Category I instrumentation and electrical equipnent for the North Anna Power Station, units 1 and 2 meets staff requirements and is acceptable.

Environmental Qualification of Seismic Category I Instrumentation and Electrical Equipnent for Main Steam Line Break~Accidents In Sections 3.10 and 6.2.1 of Supplement No. 3 to the Safety Evaluation Report we stated that we were reviewing the effect on seismic Category I instrun~ntation and electrical equipment environment resulting from a postulated main steam line break and would report the results of our review in a supplement to the Safety Evaluation Report.

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Analysis of postulated main steam line break accidents inside contain-ment performed by us and DY several applicants have calculated a contain-ment temperature, as nigh as 400 degrees Fahrenheit, which is higher than was previously used in the environmental qualification testing of safety-related equipment.

As a result, there ;s a concern regarding the capability of safety related equipment to remain operable in the accident environment which would result from a postulated main steam line break inside containment.

However, it has been recognized by us and noted in the report NUREG-v153 ("Staff Discussion of Twelve Additional TeChnical Issues Raised By Responses To November 3, 1976 Memorandum From Director, NRR TO NRR Staff") that the methods of analyses approved today contain significant conservatisms. Specifically, we have required analyses based on an instantaneous double endea steam line rupture assuming dry steam blowdown and using conservative assumptions for minimizing containment heat transfer coefficients with a conservative treatment of the thermodynamics of condensate behavior.

The applicant nas provided reanalyses for a spectrum of main steam line break accidents postulated to occur inside containment which resulted in a peak cal cul ated atmosphere temperature of 430 degrees Fahrenheit. For these reanalyses, the applicant has used a steam generator mass and energy release mOdel corresponding to dry steam blowdown.

In addition.

tne applicant has assumed that all of the liquid tormed on the heat sinks due to condensation is transferred directly to the sump without any revaporization to the containment environment.

These conservative assump-tions maximize the temperature response of the containment atmosphere 3-2

Component heat transfer calculations performed by the applicant were used to justify the adequacy of environmental qualification tempera-tures ranging from approximately 285 degrees Fahrenheit to 350 degrees Fahrenheit. The results of these analyses indicate that in general the component thermal response \\l7ill be less than the peak temperature to which the equipment has been environmentally qualified. However, certain portions of the containment electrical penetrations and the outer surface of some electrical cable were found to exceed the temperature to which they were tested during the environmental qualification tests (i.e., loss-of-coolant accident tests). Therefore, we required the applicant to submit additional data to demonstrate that these c9mponents could maintain their functional capability for the higher temperatures calcu-lated for the postulated main steam line break accident. A further investigation by the applicant into the capability of these components to sustain higher temperatures at external surfaces indicates that these components will retain the required functional capability needed to miti-gate the consequences of a postulated main steam line break accident.

Vve have performed a best estimate evaluation of the containment response to a main stewn line break accident for this facility.

Our analysis inclUded consideration of liquid entrainment in the break effluent and partial revaporization of the condensate formed on the containment heat sinks. The use of these input data result in a calculated peak temperature of about 340 degrees Fahrenheit; exceeding 285 degrees Fahrenheit (original qualification temperature of equipment) for approximately 100 seconds.

~~

have also performed thermal analyses of components needed to mitigate the 3-3

3.10.3 consequences of main steam line break. Asa result of these analyses and the environmental qualification testing performed by the applicant, we concluded that the components will retain the required functional capaoility in the high temperature steam/air environment associated with a postulated main steam line break accident inside containment.

The capability to predict liquid entrainment in the break effluent, the use of component thermal analyses and associated heat transfer coefficients, appropriate containment analytical modeling, and acceptable methods of accident environmental simulation for equipment qualification are under generic review by us.

We expect that within approximately one year our review of these investigations will be complete resulting in the development and implementation of a consistent set of environmental qualification requirements for all plants. However, based on our review of th' applicant's analyses and our best estimate evaluation, we find the North Anna Power Station, Units 1 and 2 to be acceptable for the issuance of operating licenses.

Environmental Qualification of Westinghouse and Balance-of-Plant Seismic Category I Instrumentation and Electrical Equipment In Section 22.U of the Safe~ Evaluation Report we stated that our review of environmental qualifications of Westinghouse and balance-of-plant seismic Category I instrumentation and electrical equipment is not complete.

The applicant has only recently provided us with additional information regarding these matters and therefore, our review is not complete.

Upon completion of our review of this matter, we will report the results of our evaluation in a supplement to the Safety Evaluation Report.

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6.2 6.2.1 6.0 ENGINEERED SAFETY FEATURES Containment Systems Containment Functional Design Our evaluation of the main steam line break analysis was presented in Supplement No.3 to the Safety Evaluation Report.

We concluded that the analysis and results were acceptable.

However, we subsequently determined that the analysis was not based on the most limiting containment initial conditions presented in the plant technical specifications. Therefore, we requested that the applicant perform the main steam line break analysis using the most limiting containment initial conditions.

The applicant reported the results of its reanalysis of the main stearn line break accident in Amendment No. 63 to the Final Safety Analysis Report.

The objective of the analysis was to define the envelope of containment operating conditions such that the containment design pressure of 45 pounds per square inch gauge would not be exceeded. Since the highest containment pressures are calculated if the initial containment air and water vapor masses are maximized, the containment operating modes; i.e.,

initial conditions were selected to maximize the masses and still satisfy the above objective.

A spectrum of main steam line breaks and reactor operating conditions was analyzed, and the break resulting in the highest calculated pressure was identified as the double-ended main steam line break, upstream of the flow restrictor, at zero percent power. This break, then, was used to 6-1

6.2.2 define the envelope of containment operating conditions (e.g., contaiment atmosphere bulk temperature and air partial pressure) that will be incorporated into the plant technical specifications.

The applicant's analysis was conservatively based on the assumptions of no entrainment for the Inass and energy release model, and the assumption of no vaporization of the condensate formed on the heat sinks for the containment model.

The LOCTIC computer code was used to calculate the containment pressure and temperature response.

The peak calculated containment atmosphere temperature of 430 degrees Fahrenheit also occurred for the postulated main steam line break discussed above, but for containment initial conditions that minimize the contairunent air and water vapor masses. Our evaluation of the appli-cant's thermal analysis of safety related equipment that may be exposed to the accident environment is discussed in Section 3.10.2 of this report.

We have performed a confirmatory analysis using the CONTEliIfPT-LT Mode 26 computer code, assuming no vaporization of the condensate, and the applicant's mass and energy release data. Our results confirm the applicant's results for both peak calculated pressure and temf?erature. On this basis, we conclude that the applicant's analysis is acceptable, and that the containment operating (initial) conditions have been properly defined.

Containment Heat Removal System Subsequent to the issuance of the Safety Evaluation Report, the applicant reported that it had reevaluatd the net positive suction head available 6-2

to the recirculation spray pumps and low head safety injection pumps based on a more conservative contai~~ent analysis. Net positive suction head is the head, or potential energy, available or required to force a given flow into the impeller of a pump.

Net positive suction head is affected by containment pressure, sump water vapor pressure, depth of sump water and suction piping resistance to flow.

The revised analysis incorporated analytical techniques and assumptions that were selected to minimize tL1e containment pressure and maximize the containment sump water temperature, thereby minilnizing the calculated net positive suction head available to the pumps.

The other factors, namely, depth of sump water and suction piping resistance to flow, have a lesser effect on the revised analysis. The analysis showed that the available net positive suction head was less than previously calculated and possibly less than required. Therefore, a recirculation spray pwnp and a low head safety injection pump were tested by the applicant to determine the performance characteristics of the pumps under conditions of reduced available net positive suction head.

The test results indicated that system modifications and temporary plant operating restrictions would be necessary to avoid pillrrp cavitation under postulated loss-of-coolant accident conditions. The interim system modifications proposed by the applicant to satisfy the net positive suction head requirements of the recirculation spray system pumps are discussed in the following paragraphs.

6-3

For the recirculation spray pumps located inside containment, the applicant proposes to divert 150 gallons per minute of quench spray system water to the suction side of each pump.

This will reduce the temperature (vapor pressure) of the water at the pump suction and increase the available net positive suction head. Therefore, as an interim solution, a four-inch line leading from each quench spray header will be routed to the suction side of the recirculation spray pump on the same safety train as the quench spray pump supplying the water.

A flow restricting orifice will be installed in each line to limit the flow to 150 gallons per minute.

No active components will be used.

For the recirculation spray pumps located outside containment, the applicant proposes, as an interim solution, to reduce the required net positive suction head to a point where the available net positive suction head is adequate. This will be done by limiting the flow of the pumps to 2000 gallons per minute. The flow reduction will be accomplished by.installing a flow restricting orifice in the discharge line of each pump.

Consequently, no active components are involved.

Since the flow rate will be reduced from 3700 gallons per minute to 2000 gallons per minute, it will also be necessary to reduce the number of nozzles in the spray headers to maintain an adequate differential pressure across the remaining nozzles. This will preserve the spray pattern and droplet size for the spray.

In addition to the system modifications described above, plant operating restrictions will be imposed to assure that the net positive suction head 6-4

analysis remains valid. These restrictions will remain in effect until a final resolution of the net positive suction head problem has been implemented in the plant design. Operating restrictions will be placed on the maximum refueling water storage tank water temperature (40 degrees Fahrenheit). the service water temperature (35 degrees Fahrenheit to 8U degrees Fahrenheit), the containment atmosphere temperature (86 degrees Fahrenheit to 105 degrees Fahrenheit) and the containment air partial pressure (per Figure 5-1 of the applicant's letter dated September 16, 1977).

The license will be conditioned to restrict operation pending final resolution.

The applicant has indicated that alternative system design changes are being considered for both the inside and outside recirculation spray systems, as a final resolution of the net positive suction head matter for the recirculation spray pumps.

It is our recommendation that the applicant should seek to increase the margin between the available net positive suction nead and that required by the pumps.

Tnerefore, because of the modifications discussed above our evaluation of the applicant's new containment pressure and temperature response analyses, as tney affect (l) the calculation of pump net positive suction head, (2) the containment depressurization time and the previously accepted radiological analyses, and (3) the minimum containment pressure analYSis of the emergency core cooling system performance evaluation, was required.

The applicant's calculations of the available net positive suction head for the pumps and the containment depressurization time were based on the interim system modifications and plant operating restrictions proposed by the applicant.

6-5

Our evaluation of the pump tests that were conducted to determine the performance characteristics of the recirculation spray and low head safety injection pumps under conditions of reduced available net positive suction head is presented in our consultant's (Professor l?rederick G. Hammitt of the University of Hichigan) report, "Cavitation Problems for North Anna Reactor Recirculation Spray and I/:>w Head Safety Injection Pumps. II Our consultant concluded that both the recirculation spray pumps and low head safety injection pumps should operate staisfactorily. A copy of this report is attached as Appendix C to this report.

The new containment response analysis submitted by the applicant to deter-mine the containment pressure and sw~p water temperature response was based on the following:

(1)

Thermodynamic State of Liquid and Vapor Phases in containment

'rhe analytical technique used to determine the distribution of mass and energy in the liquid and vapor regions of the contaiThuent following a loss of coolant accident can influence the contairunent pressure/

temperature response.

The pressure flash method and the ten~rature flash method are the two currently used techniques. For the net positive suction head analysis, the applicant used the pressure flash method which asswnes that liquid being expelled from the break flashes at the saturation temperature corresponding to the contain-ment total pressure. This maximizes the temperature of ti1e water entering the sump, and is, therefore, conservative. previously, the containment analytical model assumed that the liquid flashed at the dew point temperature of the contairunent atmosphere (temperature flash method). The temperature flash method is typically used for peak containment pressure calculations.

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(2) pipe Break Effluent The pipe break effluent was assumed to be uniformly mixed with the emergency core cooling system injection water spilling from the break. l'his is an important consideration for postulated cold leg breaks and essentially increases the energy transferred to the sump, with a concomitant increase in the sump water temperature. This assumption does not affect net positive suction head calculations for postulated hot leg breaks since the break effluent is already uniformly mixed.

previously, for cold leg breaks, emergency core cooling system water was assumed to spill directly to the sump wi thout mixing, which resulted in lower calculated sump ~vater temperatures.

(3)

Other Assllinptions Regarding Input Date The applicant conducted a number of sensitivity studies to identify the other assumptions that should be used to minimize the calculated available net positive suction head.

~~e have reviewed the results of these sensitivity studies and concluded that the following assumptions used in the analysis will minimize the calculated available net positive suction head:

(a)

A spray thermal effectiveness of 100 percent was ass~~ed.

(b)

A low initial contairunent pressure and high initial contain-ment temperature were assumed.

(c)

A low service water temperature entering the recirculation spray system heat exchangers was assumed.

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(d)

The containment net free volume was increased by five percent.

(e)

Switchover from the injection to the recirculation phase of emergency core cooling system operation was assumed to occur instantly at the low alarm setpoint.

A sensitivity study was also done to identify the single failure and pipe break location that will give the lowest available net positive suction head for the recirculation spray and low head safety injection pumps.

The results of this study indicate that for the recirculation spray pumps, a postulated hot leg double-ended rupture will result in the lowest available net positive suction head, and the available net positive suction head is some-what insensitive to the single failure assumption.

The available net positive suction head fo~the inside recirculation spray pumps was calculated to be 11.0 feet, and the available net positive suction head for the outside recirculation spray pumps was calculated to be 6.4 feet.

The results of the recirculaton spray pump test indicate that the net positive suction head required by the pump is less than that calculated to be available, and therefore, is acceptable.

We have also done a confirmatory analysis for the single failure and pipe break location that the applicant has identified as giving the lowest available net positive suction head for the recirculation spray pumps.

For our confirmatqry analysis, we used the CONTEMPT-LT (Mod 26) computer code.

The code was modified to permit the analysis to be based on the pressure flftsh method.

The results of our analysis, 6-8

(i.e., tne containment pressure and sump water temperature versus time) are in good agre~oent with the applicant's results.

We, therefore, conclude that the ~pplicant's net positive suction head analysis is acceptable.

In view of the system modifications that were found to be necessary to satisfy the net positive suction head requirements of the recirculation spray pumps, the applicant also performed a sensitivity study to determine the impact on the containment depressurization time used in performing the analysis of the radiological consequences fOllowing a postulated loss-of-coolant accident.

The assumption in this analysis that is germaine is that the containment is depressurized to below atmospheric pressure within an nour following the accident.

We have reviewed the input parameters used by ttle applicant to perform tne depressurization analysis and concluded that the analysis would result in a reasonably conservative calculation of the containment depres-surization time.

The limiting case for containment depressurization is a cold leg (pump suction) double~ended rupture with minimum engineered safety feature operation. A depressurization time of 3371 seconds was calculated, which is less tnan the one hour which we have previously found acceptable.

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6.3 6.3.3 Also, in view of tne changes that were made in the net positive suction head analysis, we reevaluated the applicant's minimum containment pressure analysis for the emergency core cooling system performance evaluation, to determine if any changes were necessary.

We compared the containment parameters used in the net positive suction head and minimum containment pressure analyses, and concluded that the parameters are acceptably conservative for the minimum containment pressure analysis. Also, tne analysis was done using tne Westinghouse COCO code which is conserv-atively based on the pressure flash method for determining the distribution of mass and energy in the liquid and vapor regions of the containment following a loss-of-coolant acciaent; this is the same method used in the applicant's net positive suction head analysis. Therefore, we have concluded that there is no impact to the previous minimum containment pressure analysis.

Emergency Core Cooling System System Performance Evaluation As stated in Section 6.2.L of this report, the applicant reported that it had recalculated the net positive suction head available for the recirculation spray pumps and the low head safety injection pumps based on more conservative values for containment pressure and containment sump temperature.

The calculations indicated that the available net positive suction head had been reduced, and perhaps there was less net positive suction head available than was required.

6-10

The applicant performed a series of tests to determine the performance characteristics of both the recirculation spray pumps and the low head safety injection pumps under conditions of reduced net positive suction head in order to demonstrate that the net positive suction head required for proper pump operation was not as large as originally stated by the applicant.

During the testing, the net positive suction head available at the pump suction was varied, and the net positive suction head was plotted versus total dynamic head at the pump discharge in order to determine the pump operating characteristics.

The test apparatus was inspected and the test data reviewed by us and by our consultant (professor Frederick G. Hammitt of the University of Michigan).

Based on our review of (1) the test data as discussed in the Virginia Electric and Power Company report, "Analysis and System Modification for Recirculation Spray Pumps Net Positive Suction Head, North Anna Power Station, unit 1, Docket No. 50-338, Construction Permit No. CPPR-77, September 16, 1977," (2) our consultant's report which is attached as Appendix C to this report, and (3) statements from the pump manufacturer (letter from F. F. Antunes, Ingersoll - Rand Company to S. C. Brown, Virginia Electric and Power Company, dated October 28, 1977, we concur that the net positive suction head required for proper pump operation is as presented in the report cited in (1) above and, therefore, is acceptable.

Section 6.2.2 of this supplement addresses the containment calculations which demonstrate the net positive suction head required to permit proper operation of the low head safety injection pumps.

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6.3.4 Tests and !nspections (1)

Emergency Core Cooling System Performance Tests The applicant was required to update the performance of the emergency core cooling system for the North Anna Power Station, units 1 and 2 in the recirculation mode by showing that vortex control will exist when water is drawn from the containment sump during the recirculation mode.

In lieu of insitu preoperational tests to demonstrate vortex control, the applicant performed a model test program using a one-third scale model of the containment and sump.

The model included one half of the containment building floor with all obstructions. Tests were conducted by the applicant to determine the worst combination of pump operation, source and distribution of approach flow, and submergence, with respect to vortex production. Screen blockage of up to 50 percent was also investigated in the scale tests. The results indicated some tendency to produce air entraining vortexes at high flow rates and water tempera-tures if screen blockage was present. The applicant tested various vortex suppression devices before selecting a final design. Scale tests with the selected vortex suppressor in place demonst~ated that it was effective in vortex control. These tests were witnessed by members of the NRC staff. In a letter from S. Brown (Virginia Electric and Power Company) to E. Case (NRC) dated September 13, 1977, Virginia Electric and Power Company sul::xnitted a report entitled, "Hydraulic Model Studies of the Reactor Containment Building Sump - North Anna Power station, unit 1," that concluded that the results may be projected to the full scale de~ign. we have reviewed this report and 6-12

concur with its conclusion.

On this basis, we have concluded that adequate vortex control has been demonstrated and that the sump test requirement of Regulatory Guide 1.79 "Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors" has been satisfied.

(2)

Low Head Safety Injection Pump Inspections The Unit 2 low head safety injection pump was tested in the test facility to determine the actual net positive suction head requirement, the pump was disassemDled and inspected.

The inspection revealed tnat the clearance for the upper shaft bearing was unacceptable.

One-tenth of an inch of bearing wear had occurred in an approximately twelve hour run period.

A subsequent inspection of an in-place Unit 1 low head safety injection pump has revealed approximately the same bearing wear rate.

We have requested that the applicant submit additional information regarding the unacceptable performance characteristics of the low head safety injection pump shaft bearings and the effects on the long-term reliability of the pumps.

We will require that the applicant will perform the necessary modifications to the pump units to reduce bearing wear, and that they will demonstrate the validity of the modifications through appropriate tests.

We will review the applicant's work and provide our evaluation in a supplement to the Safety Evaluation Report. This matter will be resolved prior to a decision concerning the issuance of an operating license for Unit 1.

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9.1 9.1.4 9.0 AUXILIARY SYST~1S

£.u£l_.stor age and Handl ing Fuel Handl ing Systems In Section 9.1.4 of the Safety Evaluation Report we stated that the applicant presented a conceptual design modification to the spent fuel pit that would provide a separating wall between the fuel cask loading area and the spent fuel storage area of the spent fuel pit. We also stated that (1) the final design will be provided in an amendment to the Final Safety Analysis Report prior to use of the cask, (2) until we had evaluated the final design of the separating wall or other means to prevent spent fuel damage, the technical specifications would prohibit *the movement of heavy loads over the fuel pool when irridiated fuel is stored in locations that could be damaged should the load be dropped and (3) the restriction on the use of the fuel building trolley would be removed when satisfactory design luodifications are provided.

In Amendment NO. 56 to the Final Safety Analysis Report the applicant has provided the final design of the separating wall between the spent fuel area and the cask storage area. This design precludes the cask from falling or tipping into the spent fuel area. The applicant has also completed the construction of the separating wall.

We have reviewed the information presented by the applicant. On the basis of our review of this information, we conclude that the fuel handling system and facilities design is in conformance with Regulatory Guide 1.13, 9-1

Y.2 9.2.1 Fuel Storage Faciity Oesign Basis," and is acceptable. Therefore, the restriction concerning the use of the fuel building trolley will not be included in the technical specifications. Therefore, we consider this matter resolved.

Water Systems Servi ce Water System Under the provisions of Section 50.55e of 10 CFR Part 50, the Virginia Electric and Power Company notified Region II of the Office of Inspection and Enforcement that sections of the service water piping between the service building and the Unit 2 main steam valve house were possibly overstressed due to differential settlement.

In a letter dated January 14, lY77, the Virginia Electric and Power Company further advised us that a cneck of existing building deviations versus theoretical elevations at construction time indicated that the service building E-line had settled wnile tne main steam valve house had not.

Studies performed using con-servative assumptions to determine stresses in the pipe indicated that the Unit 2 lines may be stressed above their allowable limits. Therefore.

the Virginia electric and Power Company determined that the portion of the Unit 2 service water lines under the access road should be cut out and be replaced to relieve the settlement stress.

The applicant has modified the Unit 2 service water piping between the service building and tne main steam valve house to account for the differential settlement between the buildings over the facility life.

The portion of the service water piping adjacent to the main steam 9-2

valve house was replaced.

The applicant has calculated the expected future differential settlement affecting the piping at approximately 0.25 inches. This value was increased by the applicant to provide a margin of safety above that determined by stress calculations performed in accordance with the American Society of Nechanical Engineers Boiler and Pressure Vessel Code.

The modified service water piping stresses are expected to be within the allowable code stress limits if the differential settle-ment does not exceed 0.375 inches. The applicant will monitor the settlement for the modified Unit 2 service water system to assure that this value will not be exceeded.

On the basis of our review of the information provided by the applicant, we have concluded that the modified service water piping system for Unit 2 will provide an acceptable rnethod of accommodating the expected differential settlement in the service water piping system.

We further conclude that the monitoring program to detect future building settlement is adequate to assure that code stresses in the Unit 2 piping will not be exceeded during the life of the plant.

9-3

lH.U REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS In a letter dated July 20, lY77, a copy of which is attached as Appendix D to this report, Raymond F. Fraley, Executive Director for the Advisory Committee on Reactor Safeguards, advised the staff regarding the Committee's conclusions arrived at the 207tn Advisory Committee on Reactor Safeguards meeting.

At this meeting the Committee considered whether, on the basis of information presented to tne Committee at this meeting and to the North Anna Subcommittee at a meeting on July 6, lY77, the Advisory Committee on Reactor Safeguards shou"ld reopen its review of tne application of the Virginia Electric and Power Company for licenses to operate the North Anna Power Station, Units 1 and 2.

The Committee concluded that on the oasis of this information there was no reason to alter its report of January 17, 1977 on the North Anna Station.

The letter a"lso suggested that the staff affirm that the hydrology of the site is under control.

In a memorandum dated August 1, 1977, a copy of which is attached as Appendix E to ttlis report, the staff reconfirmed the conclusions contained in tne appropriate sections related to hydrology in the North Anna Power Station, Units 1 and 2 Safety Evaluation Report and its supplements.

18-1

In a 1 etter datea November 10, H77, a copy of whi ch is a ttached as Appendix F to tnis report, f~l. Bender, Chairman of the Advisory Committee on Reactor Safeguards, suggested that modifications of the recirculation and quench spray systems to ameliorate the calculated net positive suction head deficiency may be unnecessary and undersirable. This suggestion was based on the assurance given by the Virginia Electric and Power Company that tne recirculation spray pumps, as installed, will function satis-factorily in the event that tneir net positive suction head is somewhat reauced for a period of time.

In a letter dated Uecember 2, 1977 to M. Bender, Chairman, Advisory Committee on Reactor Safeguards, a copy which is attacl1ed as Appendix G, we stated that we believe that the proposed interim containment spray system modi fications are necessary to assure rel iabl e operation of the recirculation spray pumps within the parameters for \\'1hich they were field tested.

Our basis for our conclusion is presented in Appendix G.

18-2

22.0 CONCLUSION

S Subject to satisfactory resolution of the outstanding matters described in Sections 3.10.3 and 6.3.4 of tnis report, the conclusions as stated in Section 22.U of the North Anna Power Station, Units 1 and 2 Safety Evaluation Report remain unchanged.

22-1

July 22, -1':)77 July 2b, 1':)77 July 26, 1':)77 July 29, 1 ':)77 August 1, 1 Y77 August and 2, 1977 APPENDIX A CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW VEPca letter transmitting revisions to the North Anna Environmental Technical Specifications.

UPM letter transmitting to VEPCO the ACRS letter of July 20, 1977.

Representatives from NRC & VEPCO met in Bethesda, Md.

to discuss the inservice testing of pumps and valves in accordance with ASME Section XI requirements.

Suwnary of July 6, 1977 ACRS Subcommittee Meeting.

Memorandum to Raymond Fraley, ACRS from E. G. Case concerning the Review of the North Anna Power Station, Units 1 and 2.

Representatives from VEPCa and NRC met in Louisa, Virginia to discuss the site fire protection plan.

July 27,1977 VEPCO letter concerning the cleaning of chemical additions to the service water and bearing cooling water systems of North Anna Power Station, Units 1 and 2.

August 3, 1977 Summary of July 26, 1977 meeting to discuss matters related to inservice testing of pumps and valves.

August 19, 1977 Representatives from VEPCO & NRC met in Bethesda, Md.

to discuss problems relatea to the NPSH for recirculation spray pumps during LOCA.

August 22, 1977 DPM letter concerning upgraded STS bases program.

August 24, 1977 DPM letter transmitting Supplement No.7 to the SER to applicant and service list. (Supplement No.7 issued 8/18/77).

July 2~, 1977 VEPCa letter requesting an extension of the construction completion date for North Anna, Unit 1 from September 1, 1977 to December 31, 1977.

August 29, 1977 DPM letter concerning fire protection functional responsibilities.

September 2, 1977 Summary of August 19, 1977 meeting to discuss matters related to the available NPSH for the recirculation spray pumps during a LOCA.

A-l

September 7, 1977 VEPCa letter transmitting an updated version of the fire protection systems review.

September 7, 1977 DPM lette~ requesting additional information on the reactor coolant system.

September 13, 1977 VEPCa letter transmits an Alden Research Laboratories report of the Unit 1 containment sump.

September 13, 1977 VEPGU letter transmitting a report entitled "A Seismic Monitoring Program at the North Anna Site in Central Vi rgi ni a - January 21. 1974 - Through August 1. 1977".

September 16, 1977 Representatives from NRG & VEPCa met in Bethesda, Md.

to discuss problems related to the NPSH for recirculation spray pumps during LOCA September 16, 1977 VEPGU letter concerning the recirculation spray (RS) pumps.

A report entitled "Analysis and System jvlodification for Recirculation Spray Pump Net Positive Suction Head" was enclosed.

September 19. '1977 DPrvl 1 etter to all CP hoi ders concerni ng plant security systems.

September 21, 1977 Summary of September 16, 1977 meeti ng to di scuss matters related to the available NPSH for the recirculation pumps during a LUCA.

September 21, 1977 VEPCU" etter concerni ng the securi ty program for the North Anna Power Station, Units 1 and 2.

September 23, 1977 VEPGa letter transmitting additional information regarding questions from members of the NRC Staff regaraing the recirculation spray pumps.

September 23, 1977 September 2d, 1977 September 2d, 1971 September 28, 1977 September 29, 1977 September 29, 1977 VEPGa letter containing a reevaluated response to NRG Comment SP-4.

DPM letter extending tne construction completion date for Un; t No. 1 from September 1. 1977 to December 31, 1977.

VEPGO transmits Amendment No. 64 to the FSAR.

VEPGa letter transmitting corrected figures 3.2-1 and 3.3-1.

DPM letter concerning VEPGQ's request for issuance of a 40 year operating license.

VEPGa letter concerning ASME Code.

A-2

September 30, 1977 October 5, lY77 October 7, 1917 October 1U, 1977 October 11, 1 "J77 October 12, 1~71 October 12, 1977 Uctober 13, 1977 October 13, 1977 October 14. 1977 October 14, 1977 October 17,1977 October 17, 1977 October 19, 1977 VEPCO letter addressing security provisions for the interface of North Anna Units 1 and 2.

Representatives from VEPCa, NRC and NRC consutlant met at the North Anna Site to discuss tests performed by VEPCO on the North Anna Unit 2 recirculation spray pump.

VEPCO letter transmitting Revision 1 to a report transmitted on June 24, 1977 entitled, "Engineering Report for Seismic Documentation Package, Stone & Webster Specification NA-155, Virginia Electric and Power Company, North Anna Nuclear Power Station 1 and 2u

• VEPCO transmits Certificate of Service on Amendment No. 64 to the FSAR.

VEPCO letter transmitting the response to staff comment 5.ts9.

VEPCO letter transmitting the missing pages to Amendment 64.

AlIlendment 64 was originally transmitted on September 30, 1977.

Summary of October 5, 1977 meet; ng to di scuss matters rel ated to the required NPSH for the recirculation pumps during a LOCA.

. VEPCO letter concerning impact of grid deformation on the acceptability of North Anna 1, initial core fuel assemblies.

VEPCU letter concerning the net positive suction head for the recirculation spray and low head safety injection (LHSI) pumps at North Anna 1 and 2.

VEPCO transmits proposed North Anna Unit 1 technical specification 3/4.7.12 "Settlement of Class I Structures".

VEPCO letter concerning stem mounted limit switches (SMLS).

DPM letter concerning relief from the requirements of the ASME Boiler and Pressure Vessel Code Section XI (North Anna Power Station, Unit 1).

VEPCO letter concerning a proposed technical specification requiring monitoring of groundwater levels in the area of the service water reservoir at North Anna.

DPfVi letter requesting additional information concerning the reactor coolant system.

A-3

october 21,1977 october 25, 1~77 octooer 25, 1~77 OctoDer 26. 1~77 October 27,1977 October 31, 1977 Novemoer 1, 1977 November 1, 1977 November 1, 1~77 November 3, 1977 November ~, 1977 November Y, 1977 November 1 a, 1977 November 10, 1977 VEPca letter concerning an inspection conducted at North Anna Power Station on August 23-27,1977.

VEPGO letter transmitting additional information pertaining to a report titled "A Seismic Monitoring Program at the North Anna Site in Central Virginia - January 21, 1974 Through August 1, 1977."

DPivl letter concerning physical security assessment mode-Is subject to the requirements of 10 GFR 73.55(a).

VEPGa transmits revised figures 1.1-2 and 3.2-2.

VEPGa letter requesting a delay of 30 days in which to submit an amendment to the station security plan.

VEPCO letter advising that Region II was notified on October 26, 1977 that a potential significant deficiency or substantial safety hazard existed at the North Anna site of Units 1 and 2.

VEPGO letter giving the reasons for fuel load delay for Unit 1.

VEPCO letter advising that a final design review of the North Anna Unit 1 emergency diesel loading schedules was performed in conjunction with the preoperational testing of the diesel generator.

VEPCa advises that a safety hazard existed at the North Anna Unit 1 and 2 site in the design of the control circuit for the auxil iary steam generator feed pumps.

DPM letter requesting additional information - fire protection.

VEPCO letter concerning the horizontal drains beneath the service water pump house.

VEPCO letter transmitting revisions to the security program.

Letter from Chairman, AGRS to Lee V. Gossick concerning North Anna Power Station, Unit 1, modification of rec; rcul ati on and quench spray systems.

VEPCO letter transmitting an amendment to response to NRC Comment SP-4 concerning security at the North Anna Power Station, Units 1 and 2.

A-4

November 16, 1977 NovemDer 17, lY77 November 23, 1977 November 26, lY77 VEPca letter concerning the bearing wear on Unit 2 low head safety injection pump.

VEPCU letter concerning four additional analyses using the MULTIFLEX computer code performed by Westinghouse Electric Corporation.

OPt"j letter requesting additional information concerning ECCS.

Issuance of operating license NPF-4 for the North Anna Power Station, Unit No.1. This license authorizes fuel loading only and maintaining it in a cold shutdown condition.

A-5

\\

I i

I I

I I

I

Dr

PAGES C-15 thru C-22 APPENIJIX B ERRATA TO SUPPL£~ENT NO. 7 TO THE SAFETY EVALUATION REPORT FOR THE NORTH ANNA POWER STATION, UNITS 1 AND 2 Delete pages.

B-1

APPENDIX C CAVITATION PROBLEMS FOR NORTH ANNA REACTOR RECIRCULATION SPRAY AND LOW HEAD SAFETY INJECTION PUMPS by Frederick G. Hammitt Profe~sor-in-Charge Cavitati on and 1vJulti phase Flow Laboratory Department of f'tlechani cal Engi neeri ng University of ~lichigan and NRC Consultant November 9, 1977 C-1

tNr~Ul)UCTION

"~",,,,..

On October 4, 1977, I met with members of the NRC staff to discuss their concerns regarding the pump performance tests that were done by the Virginia Electric Power Company (VEPC) using a recirculation spray pump and a low head safety inject; on pump from the North Anna Power Statl_on.

The NRC staff was concerned about the adequacy of the pump tests and the acceptabil ity of snort term and long term pump operat; on under the post-accident conditions described in Reference 1.

On October 5.1977, we visited the North Anna Power Station site to see the pump installations and the test facility; a full comprehensive discussion of the pump tests and pertinent problems ensued.

In addition to the NRC staff members and myself, tne meeting was attended by Virginia Electric and Power Company (VEPC) representatives, the VEPCa pump consultant, Stone &

Webster (S&W) representatives. and a representative of the recirculation pump vendor (Bingham-Willamette). During the meeting questions were raised to which VEPCO responded in References 3 and 4.

Subsequen~ to reviewing the information in References 3 and 4, a telephone conference call was held on October 20, 1977, involving myself, the NRC staff, VEPCa. S&W, the recirculation spray pump vendor (Ingersoll-Rand). Reference 5 ;s a written statement from the pump vendors of what was discussed during the conference call.

My evaluation of the recirculation spray and low head safety injection pump tests, and pump operation under the post-accident conditions calculated for the proposed interim solution to the pump NPSH problem is presented below, and is based on my review of the information presented in Reference 1-5 and conversations that I have had with the NRC staff, VEPCO, including their pump consultant, S&~, and the pump vendors.

EVAL.UATION The VEPCO NPSH analysis indicates that the recirculation spray (RS) pumps and the low nead safety injection (LHSI) pumps would be required to operate under the following conditions after a loss of coolant accident:

(a) The two inside recirculation spray pumps would operate at suction specific speed (SSS) values greater than 10,000 (17,000 maximum) for about 12 minutes, and at SSS values below 10,000 thereafter.

(b) The two outside recirculation spray pumps would operate at SSS values greater than 10,000 (20,000 maximum) for about ten minutes, and at SSS values below 10,000 thereafter.

(c) The low head safety injection pumps would begin operation in the recirculation mode at a SSS value of about 14.000, which would quickly drop to SSS values of 10,000 - 12,000 for long-term operation (on the order of three months).

With respect to the RS and LHSI pump designs, the pump impellors are about 12 inches in diameter and run at induction motor speeds of about 17tlu rpm, which results in a blade tip velocity of about 100 fps.

The pump impellors and casings are 18-8 cast stainless steel.

The VEPCO pump tests, which appear to have been carefully done and to have given acceptable results, show that RS and LHSI pump operation under the worst conditions of NPSH is near, but above, the pOint representing a 3% drop in the pump total dynamic head. Therefore, the operating conditions for both the RS and LHSI pumps are always less severe than that defined by this conventional cavitation inception point (see Hydraulic Institute C-3 Stanaards). It is recognized, however, that cavitation; i.e., the presence of collapsing bubbles, very probably occurs at NPSH values 3-4 times that correspond; ng to the 3% drop in tota 1 dynami c head.

(Reference 6).

Based on the foregoing discussion, I can draw the following conclusions:

1)

Cavitation damage of the recirculation spray pumps is not a potential problem because of the extremely short time (about 10-12 minutes) that the pumps would be required to operate at SSS values greater than 10,000 (Reference 7). Continued operation of the RS pumps would not present a problem since the pumps would be operating at SSS values below 10,000.

For the LHSI pumps. cavitation damage is not a potential problem because of the excellent material choice for the pump casing and impeller, from the viewpoint of cavitation resistance, and the low impeller blade tip velocity.

2)

Both the RS and LHSI pumps should operate satisfactorily under the predicted conditions following a loss of coolant accident.

The required suction specific speed values for the pumps, as determined during the cavitation performance tests on the pumps at the North Anna Power Station site, plus the normal vendor laboratory tests of many such pumps, indicate sufficient NPSH margin above the NPSH corresponding to complete pump head break-down.

According to Ingersoll-Rand, the LHSI pumps can operate satisfactorily under conditions of 50% fall-off in head.

It is also the experience of the pump vendors that short pulsations of C-4

.. "'l!I( NPSH (on the order of 10 sec) will not cause loss of suction (Le., "vapor binding"). If "vapor binding" should occur, howevel',

due to the imposition of longer duration (about one minute) zero or negative NPSH (for whatever reason), the pump head-flow could not recover without having to stop and fill. However, there appears to be no plausible mecnanism for long duration negative NPSH transients.

The vendors estimate that possible uncertainties in NPSH due to test inaccuracies and differences between "identical" manu-factured units is no more than 5-10%.

This seems completely realistic to me.

The recirculation spray and LHSI pump operating points, under the worst NPSH conditions, would still be above tne conventional cavitation inception point mentioned above.

REFERENCES

1. "Analysis and System Nodification for Recirculation Spray Pumps Net Positive Suction Head," Sept. 16, 1977.

2.

Letter No. 429 from Sam Brown, Vice President, VEPCO, to E. Case, Acting Director, NRR, Sept. 23, 1977.

3. Letter No. 454 from Sam Brown, Vice President, VEPCO, to E. Case, Acting Director, NRR, Oct. 11, 1Y77.

4. Letter No. 461, from Sam Brown, Vice President, VEPCa, to E. Case, Acting Director, NRR, Oct. 13, 1977.

5. Letter No. 4H7, from Sam Brown, Vice President, VEPCa, to E. Case, Acting Director, NRR, Oct. 28, 1977.

6. F. G. Hammitt, "Detailed Cavitation Flow Regimes for Centrifugal Pumps and Head vs HPSH Curves'" 1975 A5r'IE Cavitati on and ~lu1ti phase Flow Forum,

p. 12-15, with discussion by G. F. Wislicenus.

7.

R.. 1. Knapp. J. W. Daily, F. G. Hammitt, CAVITATION, McGraw-Hill, 1970.'

C-5

.... ~

Lee V. Gossick APPENDIX 0 UNIIED S4::XI ES NUCLEAR REGULATOHV COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON. D. c. 20555 July 20, 1977 Executive Director for Operations REVIEW OF THE NORl'H ANNA FDHER STATION, UNITS 1 Al.'ID 2 During its 207th meeting, July 14-15, 1977, the Advisory Committee on Reactor Safeguards considered whether, on the basis of information presented to the full Committee and to the North Anna Subcommittee at a meeting on July 6, 1977, the ACRS should reopen its review of the application of the Virginia Electric and Power Company for a license to operate the North Anna Power Station, units 1 and 2.

The Committee concluded, as noted in the minutes of the meeting, that, on the basis of this information, there was no reason to alter its report of January 17, 1977 on the North Anna Station.

Nr. Bender informed the NRC Staff of this conclusion and suggested that the Staff affirm that the hydrology of the site is under control.

References

1. Letter dated 1/31/75 from tt. C' Hose1y, NRC, to S. Ragone, VEPCO, reporting the results of a ~mc insepction conducted on January 7-10, 1975 at the Surry Power Station.

2. Letter dated 11/1/76 from S. C. Brmvr1, Jr., VEPCO, to B. C. Rusche, NRC, containing information on the groundwater control beneath the service water p~~phouse.

3. Letter dated 12/4/76 from S. C. Bro~m, Jr., VEPCO, to B. C. Rusche, NRC, re: Responses to Comments 2.19, 2.20 and 2.21 forwarded in the NRC letter of Noveuber 24, 1976.

4.

Letter dated 1/14/77 from S. C. Brmvr1, Jr., VEPCO, to N. C. r.losely, NRC re:

repair of overstressed service water piping at North Anna Unit 2.

5.

Letter dated 3/1/77 from S. C. Brown, Jr., VEPCO, to N. C. r';osely, NRC, reporting the findings of the insepction during the January 11-14, 1977 visit to North Anna Station.

6.

Letter dated 3/4/77 from C. N. Stallings, VEPCO, to D. C. Eusche, NRC, furnishing information requested by NRC relating to specifications for the measurement of the suspended solids and turbidity in the effluent from the horizontal drains beneath the service water pump-house.

D-1

.. ~

lee V. Gossick 2 -

July 20, 1977

7.

letter dated 3/8/77 from F. R. Brown, Army Corps of Engineers, to

\\'1. P. Gammill, NRC, transmitting a report entitled "The Undrained Cyclic Triaxial Response to A Saprolitic Soil" and the subject report.

8.

letter dated 4/20/77 from J. Allen, NAEC, to E. Volgenau, NRC, re:

North Anna Station.

9.

letter dated 5/2/77 from J. Allen, NAEC, to F. Coufal, ASLB, re:

North Anna Station.

10. letter dated 5/5/77 from J. Allen, NAEC, to D. Okrent, ACRS, re:

North Anna Station.

11. letter dated 5/16/77 from O. D. Parr, NRC, to W. L. Proffitt, VEPCO, re:

Request for Additional Information.

12.

Report to the House Committee on Interstate and Foreign Commerce, Allegations of Poor Construction Practices on The North Anna Nuclear Power Plants, dated June 2, 1977.

13. IBtter dated 6/8/77 from J. Allen, NAEC r to D. Okrent, ACRS re:

North Anna Station.

14.

IBtter dated 6/14/77 frora J. Allen, NAEC, to H. Bender, ACRS, re:

North Anna Station.

15.

letter dated 7/12/77 from J *.PJ.len, NP.EC, to N. Bender 7 ACRS, re:

North Anna Station.

0-2

Docket t10s.

and 50-13:1 SO-333 I\\*j u 1 is, ','

APPENDIX E r~UlORANDur'l FOR:

Raymond F. Fraley, (xecutivl; f11rcctol' For,Ijvisorj Commit tee On Reactor Sa fegua rds FROM:

Edson G. Case. Acting Director, Office of Nuclear Reactor Regulation SU3JECT:

REVIEW OF mE NORTH MlNl\\

PO~'IER STATIOI; \\J:,ITS 1 /\\;:~ 2 In your lGtter dated July 20. 1977 to Lee V. Gossick, Executive Director of OpercJtions, ii.~C, you acvised us tllat Hr. ScnJer sUOJcsted that the staff affinn that the hydrology of the site is under contro1.

In vievi of this suggestion tve have reexamin~d the site h'yrJrolo~y, specifically the Jroundwatcr varicJbil1ty in the sitccJreQ cJnd concluJcd that t~e aprlicant's prOjralil rrClilrding qround\\*:;:tter levels is 0CC~rt,1~le.

'.. 12 nJve also reexdlnined th~ (ioprOpriilte sections r(>1cJtcd to h,!drolo()~' in the r:orth Anna PO'r/or Station Units 1 and 2 Safety [valuatiol1 KCPOI't iH1d its supplements and reconfinn the conclusions contained in those documents.

Edson G.

Cas~, Acting Oirector Office of ~uclear ncactor Regulation E-1

APPENDIX F UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON. O. C. 20555 Nove~er 10, 1977 Mr. Lee V. Gossick Executive Director for Operations

u. S. Nuclear Regulatory Comnission Washington, DC 20555

Subject:

NORI'H ANNA PQv"ER STATICN, UNIT 1, MJDIFICATICN OF RECIRQJLATICN h'ID QUEl.~CH SP PM SYST~iS

Reference:

(a)

Vepco Report, "Analysis and System ~lodification for Recirculation Spray Pumps Net Positive Suction Head,"

North Anna Pewer Station Unit 1, SepteJl1Cr 16, 1977 (b)

Vepco letter Serial No. 487, datPd Octooer 28, 1977 froIil Sa.n C. Brown, Jr. to [1r. Edson G. Case, Oi:1R:=\\,

(Control No. 773040067) forNarding Bin~'&u-will&~~te Camp&~y letter of Octooer. 27, 1977 and Ingersoll-Rand CQupany letter of October 28, 1977 In Reference (a) (Section 1.2) the Applicant has L~dicated that tne Recirculation and Quencn Spray systems would be JIDdified to a.'"Ueliorate the c?lculated Net positive Suction Head (NPSH) deficiency by tn.e in-stallation of flow restricting orifices on the disu,arge sides of w~e outside Recirculation Spray pumps and by diverting part of the Quen~~

Spray flow to cool t..'1e containment sump.

The Comwdttee suggests that such modifications ;nay c unnecessary &:G undesiraDle in Vl2W of the assurance given in Reference (b) w~at L,e recirculation spray plliilps, as installed, '.vill fcr.cticn sat':'.sfac<::.cr il-".

in the event t..\\.-;at tneir NJ?SH is sane\\vnat reduced for a per icxi of ti7.

F-l Sincerely, 9n.~

1>l. l:',;::nc.e (

Cha.ir.nan

APPENDIX G UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DEC 2 1977 Mr. Myer Bender, Chairman Advisory Committee on Reactor Safeguards U. S. Nuclear Regulatory Commission Hashington, D. C.

20555

Subject:

North Anna Power Station, Unit 1, Modification of Recirculation and Quench Spray Systems I am referring to your letter to Mr. Lee V. Gossick dated November 10, 1977 and a supplemental letter dated November 15, 1977, in which you have suggested that modifications to the recirculation and quench spray systems for the North Anna Power Station, Unit 1, may be unnecessary and in fact, undesirable.

He are in the process of completing our review of the NPSH analysis submitted by VEPCO, including the proposed containment spray system modifications and the cont3.inment analysis.

We believe that the proposed interim containment spray system modifications are necessary to assure reliable operation of the recirculation spray pumps within the parameters for which they were field tested. The results of the staff evaluation are given in Table 1. The following describes the basis for installing the modifications.

1. Outside Recirculation Spray Pumps The available net positive suction head (NPSH) for the outside recirculation spray pumps at the deSign flow rate of 3700 gpm is calculated to be 2.6 feet. Based on the field tests, the required NPSH at this flow rate is greater than 11 feet (see Table 1). Therefore, without modifications the outside recirculation spray pumps would operate in a cavitating mode and deliver much less than the design flow rate of 3700 gpm.

To increase the available NPSH to the pumps and provide greater assurance of reliable pump operation, the applicant proposes to install a flow restricting orifice in the discharge line of each pump.

The orifices in the discharge lines will reduce the capacity of each pump to 2,000 gpm but will increase the available NPSH to 6.4 feet to the pumps.

The flow reduction will reduce the pressure drop in the suction piping. A depressurization analysis was done with this reduced flow rate and the results indicate that the containment is still depressurized within an hour and is acceptable.

G-l

Mr. Myel'" Bender DEC 2 1977

2. Inside Recirculation Spray Pumps The available NPSH for the inside recirculation spray pumps at the design flow rate of 3,300 gpm is calculated to be 7.2 feet. Based on the field tests, the required NPSH at this flow rate is 9.3 feet (see Table 1). Therefore, these pumps would also operate at a reduced flow rate. To increase the available NPSH to the inside spray pumps and maintain design flow rate, the applicant proposes to divert 150 gpm of quench spray system water to the suction side of each pump; this will reduce the temperature (vapor pressure) of the water at the pump suction.

In discussions with your staff, we understand that you questioned the need for the applicant's proposed modifications on the basis of a statement by the pump manufacturer that no pump damage would be expected even though the available NPSH was less than 7 feet. TI1is may be true; however, the pump flow requirements would not be met o An available NPSH of about 2.. 6 feet was calculated for the outside pump.

Although the field tests did not test the flow capability of the p~~ps at this low NPSH, the flow would be significantly reduced to the point that its contribution to depressurizing the containment could not be relied on.

The inside recirculation pumps would then have to be relied on to accomplish the depressurization function..

Some flow would be realized from the inside pumps, which would also be cavitating.

However, the flow would be insufficient if single failure assumptions were made i 1. e., the failure of one of the two inside recirculation pumps was assumed.

This would result in only one recirculating pump being available for depressurization, and one pump is not capable of depressurizing the containment within one hour.,

The radiological analysis for a subatmospheric containment is predicated on the assumption that the containment can be returned to a subatmospheric condition in less than one hour following a LOCA.

In summary, the proposed recirculation spray system modifications provide further assurance that the pumps will continue to function reliably throughout the course of a LOCA, including the assumption of a single active failure.

Enclosure:

Table 1 Sincerely, Edson G. Case, Acting Director Office of Nuclear Reactor Regulation G-2

~

I W

Before Spray System Modifications After Spray System Modifications Table 1 Comparison of Pump NPSH and Capacity Data for the Recirculation Spray System/North Anna Power Station, Unit 1 Inside Recirculation Spray Pumps Outside Recirculation Spray Pumps Design Flow Available Required Design Flow Available Required Rate (gpm)

NPSH (ft)

NPSH (ft)*

Rate (gpm)

NPSH (ft)

NPSH (ft)*

<3300 7.2 9.3

<< 3700 2.6 11 3300 11**

9.3 2000 6.4***

5.5

• Based on pump tests

• • Quench spray flow diverted to reduce vapor pressure of water at pump suction

• • • Flow reduced to reduce pressure drop in suction piping