ML19093A639

Safety Evaluation

ML19093A639
Person / Time
Site:	Surry
Issue date:	05/02/1968
From:	US Atomic Energy Commission (AEC)
To:	Virginia Electric & Power Co (VEPCO)
References
FOIA-2024-000060
Download: ML19093A639 (90)

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SAFETY EVALUATION BY THE DIVISION OF REACTOR LICENSING U. S. ATOMIC ENERGY COMMISSION IN THE MATTER OF VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2

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May 2, 1968

'f TABLE OF CONTENTS *

1.0 INTRODUCTION

2.0 DI.SCUSSION AND DESCRIPTION-*oF PRINCIPAL' PLANT FEATURES 2.1 Nuclear Steam:: Supply System*

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2.-*2 Sharing of Facilities and E9uip1:11ent Betweeii 'Urii~s 'No. i arid 2 2,J. Auxiliary Sys~ems

2. 4*

Containment S'tructural Design

• .. 2. 4.1 Conta:lnment St1:ucttire D~ict:Lption 2.4~2 Design Criteria 2.4.3 Testing 2.5' Instrumentation

• Page 1

3 3

6 7

.. 8.

8 9

12 2.5.1 Nuclear Instrumentation 2.5.2 Process Instrumentation

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2. 5. 3 Instrumentation for Engineered S~f,~ty!.Feathres-' '..

2.5.4 Control Rod Position Indication 2.6* Emergency Power 2.6.1 Off-site 'Power:'.*.

. 2.6.2 On-site Power 2.7 Waste Disposal System

2. 8

• Radiation Protection and Monitoring 3.0 IMPORTANT SAFETY CONSIDERATIONS 3,1 Site 3.1.1 Description 15

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- 18 i9 2-1 21 21

3.1. 2 Meteorology

. C. *.. 3.1. 3 Geology and 3.1. 4. Liquefaction e

-ii-Seismology.

22 23 3.1.5 Hydrology __

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3.2 Stress Analyses 3

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3.2.2 Reactor Coolant System 3.3, Engineered Safety Features

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3. 3.1 Sub atmospheric Conia;Lmne.~~- J:!9~J:~P.~......, ** :*,,,_:.;.:.J'c,:'.;.1-.<_,

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3.3.2 Safety Injection System*

3.4. Thermal Shock on Reactor Vessel 4.0 ACCIDENT ANALYSES 4.1 f Startup of an Inactive Loop.

4. 2., Steam Line Rupture

4. 3 Steam Generator.Tube Rupture(_.;,"'".. -,"

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4. 4. Design Basis Accident 5.0 QUALITY CONTROL 6.0.STAT.ION DESIGN WITH RESPECT TO THE 70 GENERAL*DESIGN*-;CRITERIA':*:*;

7.0 RESEARCH AND DEVELOPMENT

8. 0 REP,ORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEp~~-S.. J,:::.. :'::*~-~

9. 0 TEClINICAL QUALIFICATIONS 10.0 COMMON DEFENSE AND SECURITY

11.0 CONCLUSION

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-iii-Appendix A - Chronology - Regulatory Review Appendix B - Report of Advisory Committee on Reactor Safeguards Appendix C - Report of u. s. Weather Bureau

.Appendix D - Report of u. s. Army Coastal Engineer1ng Research Center Appendix E - Report of u. s. Geological Survey Appendix F - Report of u. s. Coast & Geodetic Survey Appendix G - Report of Nathan M. Newmark Consulting Engineering Services Appendix H - Report of u. s. Fish and Wildlife Service

1.0 INTRODUCTION

The Virginia Electric and Power Company (VEPCO) by application dated March 20, 1967, has requested a license to construct and operate twin nuclear power units designated as Surry Power Station, on a site on the James River in Surry County, Virginia.

Each of the two units will include a three-loop pressurized water reactor nuclear steam supply system and turbine generator suppiied by Westinghouse Electric-Company (Westinghouse).

The remainder of the 1':

unit is designed by VEPCO or the architect-engineer, Stone & Webster Engineering Corporation (Stone & Webster).

The design power rating of each unit is 2441 megawatts thermal [MW(t)]

wi,th an ultimate c~pabili ty. of 2546 MW ( t)

• These are

• equi val~nt to electrical

'power ratings of 815. 5 and 847. 5 megawatts electrical [MW(e)], respectively.

Consequences of accidents and the adequacy of the engineered safety features ha~e-been *analyzed and evaluated by the applicant, the Atomic Energy Commission's regulatory staff, and the Advisory Committee on Reactor Safeguards, assuming a reactor power level of 2546 MW(t); however, the thermal and hydraulic character-.

istics of the reactor were.evaluated at 2441 MW(t).

Before operation at any power level above 2441 MW(t) is ~uthorized, the Commission's regulatory staff must perform a safety evaluation to assure that the reactors can be operated

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' s~fely, at the high~r power level.

The technical safety review of the proposed plant has been based on the applicant's Preliminary Safety Analysis Report (PSAR) and eleven amendments, all of which are contained in the application.

The technical evaluation of the preliminary design of the proposed plant was ~ccomplished hy the Division of.

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[ Reactor Licensing with the assistanc~ of consultants, as requested.

Appendices C through H include the reports of our consultants on meteorology, floodin~,

'f potential, geology and hydrology, seismology, structural design, and radio-logical monito!ing.

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In the course of the review of the material submitted, a number of meetings

. : ~:. >~ ::., '." r~*j were held with representatives of the applicant to discuss the proposed plant.

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As a consequence, additional information was requested of the applicant which

• • i was provided in certain of the amendments.- A chronology of the review process is attached as Appendix A to this report.

The Commission's Advisory Committee on _Reactor Safeguards (ACRS) has met

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'i*.*i-with both the applic~nt and the staff, A copy of its report to the Connnission

• • ... -*::::*:q on the Surry Power Station Units 1 and 2 is included as Appendix B.

l The review and evaluation of the pr~posed design and construction plans of

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i. *.* *,!../'1 the applicant at the construction permit stage of the proposed units is the

-'-~' ::* 't staff of the design, construction, and operating features of Surry Power Station.

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Prior to issuance of operating licenses, we will revie:'7 the final design to

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determine that all of the Commission's safety requirements have been met.

The

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units would then be operated only in accordance with the terms of the ope:ating

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licenses and the Commission's regulations under the continued surveillance of

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The issues to be considered, and on which findings must be made by an

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-, J; Atomic Safety and Licensing Board before the requested co?struction permit may b

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2.0 DISCUSSION AND DESCRIPTION OF PRINCIPAL PLANT FEATURES 2.1 Nuclear Steam Supply System_

The nuclear steam supply system for each reactor consists of a light-water-moderated pressurized water reactor (PWR) which transfers reactor heat to three steam generators~

The basic design is similar to that of other Westinghouse reactors now under construction.

The fuel for the reactor is low enrichment uo2 pellets enclosed within Zircaloy tubes.

Two-hundred-and-four fuel rods are arranged in a square array to form a fuel assembly.

The rods are axially supported by spring clip grid assemblies.

These grid assemblies are welded to control rod guide thimbles.

In conjunction with the nozzles located at the top and bottom of the fuel assembly, they provide the structural support for the fuel.

The reactor core contains 157 fuel assemblies which rest on the lower core plate.

Leaf springs are provided between the top nozzle and upper core support plate to constrain fuel assemblies from moving.

Fuel assemblies with the uranium fuel enriched to 2.94 weight percent U-235 will be located around the periphery of the core.

In the central region of the core, fuel assemblies with uranium fuel enriched to 2.44 and 2.54 weight percent U-235 will be loaded into a "checkerboard" pattern.

The thermal-hydraulic parameters of the proposed VEPCO units are similar to those of other recently licensed Westinghouse plants.

A sum1nary comparison of significant parameters has been included in Table 2.1.

The minimum DNB ratio at nominal conditions is 1.86, the average linear heat generation at rated power is 6.2 kw/ft, and the maximum linear heat generation rates are Table 2.1 Comparison of Westinghouse PWR's Parameter Total heat output (MW).-

Number of-* 1oops Nominal system pressure (psia)-

Hot channel factors

- Fq F~H DNBR at nominal conditions Average coolant velocity (ft/sec)

Nominal inlet temperature (°F)

Average core 4 T (°F)

• ,*2 Average core heat flux - Btu/hr ft 2

Peak core heat flux - Btu/hr ft

~verage linear heat generation rate

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Maximum linear heat gene-rad:on rate at rated power (kw/ft)

Maximum linear heat generation Fate

• at il-2% overpower** (kw/ft)

Peak fuel" temperature -a:t ov'Efrpower (°F)

Number of* fuel assemblies

• Em:'ichment * (weight percent-_ U-235)

Region Region 2 Region 3 H. B.

Robinson

• 2094 -

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• -2250-

. 3 0 25 1.88 1.85 14.5 546.5 59 164,000

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533,000 5.3 17.3 19.4 4270 2.28 2.43 2.73 Diablo Canyon 3250 4

2250 2.82

1. 70 1.81 15.7 539 63 207,000

~83,000 6.7 18.9

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• 4450 193

• 2.2 2.7 3.3 Surry 244-J: - -

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1. 70 1.86 14.5 542.9 68 191,000 539,000 6.2

• 17.5 19.6 4356

- _:157

- 2.44 2.54 2.94

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e 17.5 and 19.6 kw/ft at 100% and 112% of power, respectively.

To evaluate the ability of the three-loop core to operate safely at these power densities, the applicant has performed parametric studies varying uo2 thermal conductivity as a function of temperature, and hot channel factors, coolant flow rate, coolant inlet temperature, and gap conductance to determine the effect of small errors at both rated power and.overpower conditions.

These studies indicate that small deviations from nominal conditions will not result in fuel damage.

Based on our evaluation of the results of these studies, we have concluded that the thermal and hydraulic design is conservative with adequate margin in DNB ratio (1.86.at 100% power with design peaking factors) and fuel temperature

_(maximum value of 3990°F at 100% power with design peaking factors) provided to prevent the occurrence of significant fuel damage during either normal operation or.anticipated transients.

Accordinglr, we conclude the core can be operated safely at the power density proposed.

Reactivity control is accomplished by full length sliver-indium-cadmium control rod assemplies, part-length silver-indium-cadmium control rod assemblies, fixed burnable poison rods (borosilicate glass in stainless steel tubes) and by liquid poison (boric acid) in the reactor coolant..

Reactor coolant at 2250 psia is circulated through the core by three centri~ugal main coolant pumps.

During rated operation, the coolant enters the reactor at 543°F and exits from the reactor pressure vessel at an average temperature of 606°F, The coolant then passes through three steam generators where steam is formed at a pressure of about 700 psig to drive the turbine.

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The nuclear steam supply system, as designed by Westinghouse, is similar to that of other pressurized water reactor systems that have been licensed.

e However, VEPCO has chosen to install two motor-operated stop valves in each of the three loops of the primary system to enable isolation of the pumps and steamgetierators.

These valves are slow acting and are' supplied with inter-locks* to *prevent *injection of cold water by startup of an inactive loop.*

• The efrect of these valves on plant safety is discussed in.the Accident Analyses section*of this evaluation (Section 4.0).

Our evaluation of the stres~ ahalysis

'ari:d thermai shock considerations of the pressure vesse1: are discussed in the Important Safety Considerations section of this evaluation (Section 3.0):

2.2 Sharing of Facilities and Equipment Between Units No. 1 and 2 The *design of systems which are shared between the two units was reviewed in detail.

The two units have identfca:1 design criteria and their designs

'differ only in those area:s affected by plant layout.

Systems*which are 'shared in the true sense (Le., one system serv-~s both p*iant*s as opposed to" two

. separate full capacity ititerconnecfed systems* which may or may not share spare components) are. of an auxiliary nature.

These include *the boron addi tfon**

portion.of the chemical and volume control system, the*component cooling water

. system, the fuel storage arid handling syst'em~ the servic~ ~ater system, -the condensate storage tank, and'the boron-recovery and' waste disposal systems.

Shared systems are sized to the requirements of *both units; Other systems which are interconnected but* consist of two separate syst'.ems* ate the' on-site power system, switchyard. components' of the off-site power system, and the steam generator emergency feedwatet' system~'** 'Structures which 'are shared.

include the au:x:iliary building' turbine building' control area; and 'the *fuel

• storage building.

The* safety related features* of these shared or interconnected facilities or systems are discussed in Sections 2~3, 2.6, and 2.7 of this ev~iuation.

2.3 Auxiliary Systems The Chemical and Volume Control System is essentially identical to that supplied on other Westinghouse PWR's (for which construction permits have been granted) with respect to the.manner of re-cycling letdown primary system water through the purification subsystem and thence back to the primary system.

However, since the charging pumps also serve as the high-head safety injection pumps, automatic isolation valves which close on a safety injection signal are installed to isolate the charging pumps from the remainder of the charging system.

These valves are discussed in the Section 3.0 of this evaluation.

River water will be used as service water and condenser cooling water.

The four circulating water pumps, located at the east side of the site, take water from the James River and raise it to a paved intake canal with a normal water surface elevation of 23.8 feet.

The water then flows by gravity the length of the canal to the plant area at the west side of the site.

This canal is approximately 1.5 miles in length and contains a minimum of 45,000,000 gallons of water available for cooling.

Flow through the main condensers is by gravity.

In the event of a loss-of-coolant accident coincident with a loss of off-site power, the main condenser inlet and outlet valves are automatically closed, and the valves in the lines leading to the recirculation spray h_eat exchangers are automatically opened.

These valves are powered from the emergency bus.

Service water flow from the canal through the recirculation spray heat exchangers is by gravity.

Sufficient water is stored in the canal to ensure removal of decay heat for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> without makeup.

In addition, a diesel-powered emergency service water pump is located at the intake structure.

r This pump is designed to provide an adequate emergency river water supply to remove decay heat in the post-accident environment even if the canal liner were to crack.

In our opinion, the proposed use of a heat removal system ;hich requires no energy input during the first 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.after anacc:l.dent represents ~n improvement in the reliability of the systems provided 'to reject decay heat

• to the ultimate 'heat sink over th6se provided in other PWR' s* *. We have evalu-ated the consequences of fcii1ures 'in this system and have cansidered the effects of t'arnadoes and hur~icanes,onthe :tntake stru~tu~e. *We have concluded that no single failure of either active or passive components would result in the 'inab'ility to supply required cooling water to the two units.

2. 4 c*ontainmen:t Structural Desig'n 2.4.1 *containment Structure Descr'iption The 'reactor containment: structure. consists af a steel.:..lined reinforced con-crete vessel' with straight 'cylindrical w~lls' a flat base~* and a hemispherical dome.

The liner will be fabricated from carbon steel plates'* canforming to the

~peciffcati~n fa~ ASTM A.:.442 Grade 60 material having a.*guaranteed mi~imu~ yield point of 32,000 psi.

The reinforcing in the cyiind~:i.c.ai portion of the vessei will consist af

, c... h6rizontal and Vet'tiC~l ba~S*, and tangent:f.°:al. diagonai b*ars placed 'at 45~ to the horizon'tal in both cifr'ections in the wall to resist tangential shear.

Radial shear will be r*esisted by stirrups or diagonal bars.

The reinforcement bars will conform to ASTM A-15 or ASTM A-408 specifications.

Fer size 14S and 18S bars, the C~dweld method of splicing will.be used except for a minor number of splices and connections to other structur~l elements, which will be ma.de by welding.

(a e A personnel hatch and an equipment h.atch. are provided for access to the containment vessel.

In addition, there are other penetrations for piping and electrical conduits.

2.4.2 Design Criteria The reactor containment concept is based on an operat_ing pressure of aJ?proximately 10 psia.

The containment will be designed to withstand within the elastic range, the effects of the operational basis earthquake* concurrent with accident loads.

The containment will also b_e designed to :withstand a design basis earthquake**

concurre_nt with accident loads without loss; of function.

The accelerations used for the operational basis earthquake and the design basis earthquake are in agreement with those given i_n the report of the U. S. Coast and Geodetic Survey, attached as Appendix F.*

To achi_eve a degree of conservatism and to provide a reasonable amplifica-tion factor in the velocitY. ccmtro~ling region of the response spectra, the

. 9.pectra for the. design basis earthquake will be. normalized to a maximum ground motion velocity of 9 in/sec i_n the frequency range between 0.3 and 2 cycles per second.

The* operational b_asis earthquake spec_tra_ will be scaled from the The "Operational Basis Earthquake" for a reactor site causes the vibratory ground motion for which all features of the facility necessary to permit continued operation are designed to remain functional.

The maximum ground acceleration of the Operating Basis Earthquake for Surry Station is 0.07g.

_The "Design Basis Earthquake" for a reactor site causes the vibratory ground motion for which *all features of the facility necessary to protect the health and safety of. the public are designed to remain functional.

The Design Basis Earthquake is.the largest earthquake postulated for the site which for Surry Station is 0.15g.

I

• " des~gn basis earthquake spectra for consist_ency in design.

We and our con-

'. suitants. agree *with the general method* of dyna~ic analy~,i~ *for the containment structure.

The --~pplicant will use a damping factor of 5 per~ent for. the operational basis earthquake.

This will be associated with.stress.levels at or slightly below yield and at least a moderate degree of cracking.

For the design basis'" earthquake, the applicant will use a: damping factor of 10%

which includes both damping in the reinforced concrete st~~cture arid *in th~ soil.

Th*e* applicant will.. use lower damping values for Class I system and components

• other ; than the s'truc'ture~.

The' structur'e is to be designed to meet the stress *and strain liiii.its. of

• Ac! 318-63, Part IV_;B, "Sti~cturai" Ana:iysts arid Proporti~nin.g ef Membe:ri; t** --

. *u1dfo.ate Strength Desig~.",,

We and our* seismic design consultant', Nathi:,.n M. Ne"wmark Co~stild.nf Engineering Services, are in agreement with the.damping factors~ l6ad ~om.15:1.na-tions, and *alloJab.le ~tresses si'elec*ted*by* the*:.applicartt.

In partl~ular, we are ill a:gr'eenient* with the requirement' that th~ maximum allowable stress.... ::-

• resultitig

• frorii 'the cotnb':i.nation of* 1oacis* reiiurt:rn"g from* *the. design bas ii a.tcident

'and 'the design basis earthquake with ncfrmaf operating* i~ads b~- limited to 80 percent.of the ini'rii~~ te1isile ~t_rength of th~ r'einfordng steel. Fot' th~t

. rei_nforcing_ st~el chosen, ~O, per~ent of,th~ Ill~ni~um tensile strength does I_,10t exce¢~i _90 percent of the mini~Ulil. yfeld ~.tr~~ith._- ;*Thus, th~ 'steel remains within the elastic range.

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The spa*cing of anchors_,_for *the liner will.be such that. the critical buckling stre~s ;i;L1' be belew the' all'owable stress.

The"* 1tm:h;ng stresses will f

1, e also b,e in accordance with Sectio_n 3 of the ASME Boiler and Pressure Vessel Code.

The,sttesses and strains will be limited to values below yield for both the operational basis and design basis earthquake conditions.

Stress analysis in the region of the large openings of the containment vessel will'be evaluated considering operating loads, seismic loads, and loads due to pressure and temperature resulting from the loss-of-coolant accident.

The analysis will include calculation of ax~al and torsional loads, bending moments, and shear in the ring beam surrounding the opening.

The compatibility of deformations between the ring ~eam and the concrete shell for the various loading conditions will be checked by the applicant as the design effort progresses.

Based on our evaluation of the design criteria and methods to be used in design, we conclude that the design approach proposed by the applicant will provide an adequate margin of safety.

The report of Nathan M. Newmark Consulting Engineering Services, attached as Appendix G,substantiates this conclusion.

2.4.3 Testing Strength and.leakage tests will be performed after the construction is completed.

These tests will.be s.imilar to the tests previously conducted on containment structures of the same type, such as the Connecticut Yankee facility.

, No special provisions have been made for monitoring this reinforced con~

crete structure during the lifetime of the plant.

However, the containment and the liner will be accessible for visual inspection.

In addition, test channels are provided at all liner seams; and test collars, rings, and taps are provided at all penetrations for leakage testing.

We conclude that these provisions.will allow an acceptable surveillance program to be established.

2.5

• Instrumentation We have evaluated the instrumentation system proposed using*for guidance the General Design *criteria and the proposed IEEE Standard for Nuclear Power Plant Protection Systems (Rev. 8). * *A comparison has been made with the Diablo Cari.yQtl instrumentation in particular.* All.significant matters related to instrumentation system design are discussed below.

2. 5.1. Nuclear Instrumentation

. : Source and Intermediate Range** Channeis':.. The* Surry design provicfes period indication from below criticality 'to heatup** range: in a manil'er similar to* the Di.ablo Canyon design.

Period* information derived. from ali four channels will be continuously displayed when: the reactor is being" operat.ed in. these. ranges.

Power Range Channels:

Signals from the four power range (proie*ction)

• ciiS:n'riels are combirted to form the irtputsig~al for.the reactor control system.

This method of combining fi.uiction does*not conform to the requirements of General Design Criterion No. 22 or of IEEE S~ction: 4. 7 regar"ding the separation of ~ontrol'~ri.d protecticiri functions.

2.5.2 Process Instrumentation Pressurizer Pressure:

The Surry St*ation

• design incorporates complete separatio~ :of control and safety. flincdons ~**

Thiee ind~pendent charinel's (2/3' 'iogf~j provide the high' arid l~w p~e§~~te* prcit*e'ction.

Two additional cha~nels are used -for ~orit~oL The ti'an:sdu.cers 'are fed froni three *separ~te

"{a~s ~t'.. t:h~-. p~e~s~rizer. ~iici(that f~:i.l~re;.of itistruiilent pipin;g, at *a tap will disa'.ble otiiy ;;n~ channel.

This. cons ti tut.es'* an improveinertt' 6ver the Diablo Cany~n desi~n. which utilizes f Jur p*;es~ure sensi~t ~harinels {two-o~t-of-four. [2/ 4 l logic), one of which p~ovide*s both* control a~i ~*~f~ty :fonctlons ~

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• • Overtemperature:

There will be two sets of T-avg, sensors, one for control and one for protection.

The other inputs to the overtemperature protection channels -- primary system pressure, delta-flux (the signal difference between the top and bottom sections' of the long ion chambers) and primary system delta-T -- will each be derived from independent channels which have no control

• -- function.

Thus, there will be three independent overtemperature protection channels (2/3 logic) which are *separate fro~ control.

This constitutes an improvement over the Diablo Canyon design in which the primary l~op average temperature (T-avg) sensors are connected to the control and protection systems, and the primary loop delta-T and the power range nuclear channels, which are a part of the protection system, also provide inputs to the reactor control system.

Overpower:

As in the Diablo Canyon design, each overpower protection channel utilizes two distinct signals:

primary system delta-T and delta-flux.

The Surry design utilizes three independent overpower protection channels (2/3 logic) which are*separate from control.

This constitutes an improvement over the Diablo Canyon design in which these channels also provide control functions.

Low Reactor Coolant Flow:

The instrumentation is essentially identical in both systems.

The trip logic among loops_..:. 1/4 or 2/4 at Diablo Canyon and 1/3 or 2/3 at Surry -- is a natural result of a four-loop vs. a three-loop station.

The difference is not significant in terms of reactor protection logic, Low Steam Generator Level The Surry design differs from the Diablo Canyon design in one respect:

the monitor which senses deviation between the level channel assigned the control function and one of the level protection channels has been deleted.

This modification constitutes an improvement since

I the.re is now a greater degree of independence among the level cha'.fi.nels.

Still remaining, however, are several areas where control and prot~ctio~ functions are

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combined.

For example, at each steam generato~ there a;e three very-low-level" channels (2/3logic), one of which provides an input to the three-ele~ent level

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controller.

An unsafe malfunction in the common level channel which resulted in a slow rate of level decrease would remove all redundancy since the t-rip logic of the remaining two cha~nels'would be two.:..out-of-~o.

This portion of

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the design does not conform to the requirements of General Design Criterion No. 22 or of IEEE Section 4.7.

Another example of control and safety interaction. occurs in the *i'ste~:.

flow, feedwater-flow mismatch" channels.

Both channels provide control sign:als, and also act" in coincidence with two of the,ilo~-level". chan~els' to provi~~ ci scram function; trip logic is i/2 iow-ievel in.coincid~nce *with*i/2 ffow-mismatch.

Our analysis shows. that if ~ faulted flow 'sens~r indic~ted a sp-ur:ious high'. -.

feedwater flow or low-steam flow' the wat~~ tev~i' could drop ;api.'dly and the affected mismatch protection channel wo~ld *also* be disable'd by the original" fault.

Trip logic wo~ld then revert to one-of~two "low-level" in** coin.cidence with the single remaining "mismatch" channel.

This is a non-redundant system and :i.s at variance wid1 the provisions of General Design Criterion No *. 22 a'.nd of IEEE S~ct:i.on 4.7.

High Pressurizer Water Level:

The design is identical to that prop~sed'for Diablo Canyon.

There are three channels (2/3 logic), on*e of which" also *prci.::.**

vides the control function.

An unsafe malft.inct:ion*in' the.common channel would

.. rem~ve all redunda~cy since the tr:lp logi~ of the remaining channels would be two-out-of.::two.

This design does not conform to the pr'ovisions of Genera:i Design Criterion No. 22 or of IEEE Section: 4.7~

{,

e For those areas which do not conform to the provisions of General Design Criterion No. 22 and IEEE Section 4.7, the applicant will be required to separate control and protection instrumentation to the fullest extent practicable, In other respects, we conclude the instrumentation systems proposed are in con-formance with the General Design Criteria and the proposed IEEE standard and are acceptable.

2.5.3 Instrumentation for Engineered Safety Features The instrumentation systems which initiate containment spray and containment isolation are similar to those provided at Diablo Canyon except that, (a) at Surry, safety injection signals as well as high containment pressure signals will initiate closure of containment purge lines, and (b) at Surry, containment spray is initiated by one set of three containment pressure signals (2/3 logic) instead of two. sets acting in coincidence as at Diablo Canyon.

We conclude these differences constitute design improvements since (a) additional redundancy is provided in closure signals to the valves in the containment purge lines, and (b) fewer signals are required to initiate containment spray.

Safety injection is initiated by two low pressurizer level (2/3 logic) plus two low pressurizer pressure signals (2/3 logic). It is also initiated by two high containment pressure signals (2/3 logic).

With this design, no single failure can disable the ability of the system to function.

Based on the above, we conclude the initiation system for containment spray and safety injection is acceptable.

2.5.4 Control Rod Position Indication The Diablo Canyon design incorporates a single position readout indicator into which the operator could successively switch the output signals from the

e individual position transmitters for each rod.

At Surry, the positions of all rods will be displayed on three multi-point recorders (a portion of the rods per recorder).

The cycling interval for each recorder is approximately 30 seconds.

In addition, a deviation alarm (similar to the deviation alarm at Diablo Canyon) will warn of the departure of one or more rods from the group "demand" positions.

We conclude the design is acceptable for the following reasons:

a. The positions of all rods are continually presented to the operator at intervals no greater than 30 seconds,

b. No manual switching action is required.

c. The recorder traces serve to indicate any trends, or actual departures, from the norm.

d. Departures from the norm (e.g., a dropped rod) are instantly alarmed.

e. The*"demand" positions of all rod groups are displayed and constitute a diverse presentation of rod position.

2.6 Emergency Power 2.6.1 Off-site Power Normal (off-site) power to Unit No. 1 initially will be supplied from (a) six 230-KV lines, and (b) a single 500-KV line.

Upon completion of Unit No. 2, normal power will be available to both units from an additional 230-KV line and from an additional 500-KV line.

Within each unit there will be two emergency buses.

Normal power to each of the emergency buses of Unit No, 1 will be supplied from separate startup transformers, The buses of Unit No. 2 will be supplied from one of the

-17"'-

aforementioned startup transformers and a third transformer.

Based on the redundancy of incoming lines and the arrangement of the startup transformers and the emergency buses, we conclude that the normal power system design is acceptable.

2.6.2 On-site Power The standby (on-site) power system*consists of three diesel generators, two of which will be installed with Unit No. 1.

The third generator will be installed upon completion of Unit No. 2, They are connected to the emergency buses in a manner similar to the startup transformers -- one bus in each unit with its own diesel generator; the remaining two buses sharing the third generator.

The generators will be physically and electrically isolated from each other, Each diesel generator will have independent fuel systems, air supply starting systems, and control equipment.

They will be located in a Class I structure.

Each unit will be provided with two independent and redundant station batteries.

Our review of the standby power system indicates that the diesel generator units are redundant since any one of the two *available to each unit can supply the required engineered safety fe'ature loads.

The independence of the generator units will be preserved by independenc~ of vital appurtenances, such as starting and control systems, fuel pumps, etc., nor will they be synchronized to one another.

Breaker control circuits (load shedding, load connecting, etc.) will be energized from redundant d.c. systems such that the failure of one of the battery banks can be tolerated,

• ' Essential electrical_ equipment components will be specified to withstand

~

without loss of function the maximum conditions expected to exist subsequent to a *design basis accident.

Certified test_data will be requested from equipment vendors to ~onfirm that the component or system will survive the post-incident environment.

If such data are not available, the applicant (or his contractors) will perfo"rm the necessary qualification tests' or provide a local environment for the equipment.

In view of the redundancy of the on-site system, the physical and electrical

• independence of essential equipment, and environmental testing (or local environ-ments) for vital equipment within containment, we conclude that the proposed on-site power system design is acceptable.

2.7 Waste Disposal System Small quantities of radioactive waste products are generated in the normal

'* (.*

operation of a nuclear power plant.

The liquid waste is collected, stored, treated, and either re-used or discharged.

The liquid waste disposal system is designed to meet 10 CFR Part 20 discharge limits taking credit for dilution by service water alone.

The additional dilution available in the discharge canal afforded by the circulating water system will of course further reduce the con-centrations released to the river at the site boundary.

The effluent will be continuously monitored.

High activity will cause the liquid effluent flow control valves to close, thus terminating release of liquid effluent.

• ~* t"'.'

Gaseous radioactive waste is collected, compressed, and stored until sufficient radioactive decay has occured to permit discharge within the require-ments of 10 CFR Part 20.

Monitors are installed in the process vent system.

,*-t,:.

e e If radioactive releases occur which could result in excessive off-site concen-trations of radioisotopes, discharge of gaseous effluent will be automatically terminated.

The system proposed utilizes a hydrogen-oxygen recombiner to reduce the volume of gases to be stored, and is sized to accommodate the volume of gaseous waste which would need to be processed if the recombiner were to fail.

The *increase in volume of wastes resulting from a load-following mode of operation of both units has been considered in the design of the waste disposal system.

Hydrogen monitors are installed in the inlet lines of the storage tanks and means are available to add a nitrogen cover gas should hydrogen concentra-tions approach explosive limits.

The waste disposal system proposed is acceptable because it is adequately sized and provides automatic means of terminating releases of high levels of radioactivity.

Based on the above, we conclude that off-site doses resulting from normal plant operation will be well below 10 CFR Part 20 limits.

2.8 Radiation Protection and Monitoring Radiation monitors.will be provided to (1) detect gross fission product releases from the core, (2) detect primary system leakage, (3) provide personnel protection, and (4) determine the effect of normal operation and of accidental releases on the environment.

These provisions are discussed below.

1.

Monitors are located on the letdown line to the chemical and volume control system.

These consist of a low-level device using a Geiger-Mueller tube.

Monitor readouts are located in the control room and are equipped with a variable alarm setpoint.

The applicant has estimated that gross fuel element failures would be rapidly detected by this device.

r 2.

The applicant has analyzed the response of the containment air particu-late monitors to primary system leakage.

This analysis indicates that even with failed fuel inventories as low as 0.01%, a leakage rate of 10 cc/min would be detected within. one hour. Thus, this analysis indicates that small leakage rates can be detected in a reasonable period of time.

In addition to the particulate monitor, containment humidity indication, and containment sump water levei indication will be provided.

In addition, a leakoff will be provided between the double gaskets of the vessel head-to-vessel closure joint.

3.

Shielding design criteria, plans for personnel monitoring, and the location and ranges of area monitors will be specified in the detailed design of the plant to meet the 10 CFR Part 20 requirements.

4, VEPCO plans to conduct a detailed environmental monitoring program during plant operation.

This will be supplemented by a pre-operational program to determine background levels.

This program will include the collection of samples of rain water, river water, soil, vegetation, airborne particulates, small animals, fish, shellfish, and plankton.

Reconcentration of radionuclides in aquatic biota will be evaluated with particular emphasis on oysters, in view of the nearby commercial shellfishing beds, and on fish caught by sport or commercial fishermen.

The reports of the Fish and Wildlife Service also recom-mend the conduct of environmental surveys.

Copies of these reports are attached as Appendix H,

• e 3.0 IMPORTANT SAFETY CONSIDERATIONS

'3.1 Site 3.1.1 Description

• The proposed site is located in Surry County, Virginia, on Gravel Neck, a small -peninsula in the James River.* Gravel Neck is 4.7 miles northwest of the nearest corporate limit of Newport News, Virginia, and 14 miles northwest of the nearest suburban residential district of Newport News.

Williamsburg, Virginia, is seven miles north of the site.

Jamestown Island, a popular tourist attraction, is three miles northwest of the site.

The site is a tract of 840 acres.

The minimum distance to the site boundary is 1650 feet.

The nearest dwelling is 0.6 miles southwest of the reactor containment structures.

Land usage in the vicinity is primarily agricultural, with the major products being hogs, peanuts, timber, soybeans, and corn.

The nearest dairy farm is 3.5 miles north-northwest of the site on the northern shore of the James River.

Military reservations occupy most of the land on the north side of the James River.

The area immediately adjacent to the proposed site is sparsely populated.

Estimated population is presented below:

Distance from Containment

[miles]

1 2

3 4

5 10

• 20 Cumulative Population 4

(1966) 36 (1966) 121 (1966) 295 (1966) 791 (1966) 39,212 (1960) 178,386 (1960) e Considering the very low potential doses which could result as a consequence of the design basis accident, the low population zone radius as defined in 10 CFR Part 100 is conservatively three miles.

The nearest boundary of Newport News is 4.7 miles from the site, This -is considered the population center distance even though there is low population density for a total distance of 14 miles in Newport News.

In the Accident Analysis section of this report (Section 4.0), we have shown that radiation doses to the population would be within the 10 CFR Part 100 guidelines at both the site boundary and the low population distance, assuming the postulated fission product release were to occur.

_I 3,1,2 Meteorology Meteorological data which are used to evaluate atmospheric diffusion for the site are available from instrumentation located at Ft. Eustis, approximately four miles* east of the site.

These data from Ft. Eustis* will be supplemented by an on-site meteorological program which will gather information on short-term and 'iong-term averages of wind speed and direction, and atmospheric stability and persistence of stability conditions.

At least two years of on-site meteoro-logical data will be a~ailabie prior to issuance of an operating license.

The data from Ft. Eustis give no indication of unexpected meteorological conditions.

The potential off-site radiation doses :res'i11ting from accidents have been analyzed, assuming invariant wind direction, wina' speed of 1-m/sec, and Pasquill Type-F stability for the duration of the accident.

We have evaluated the model proposed and in view_of the data from Ft. Eustis have concluded that it is conservative/ Site meteorology has also been reviewed and the diffusion values r

found conservative by the Environmental Meteorology Branch, Weather Bureau, ESSA.

Its comments are attached as Appendix C to this report.

Structures have been evaluated for a tornado having a tangential wind velocity of 300 mph and a central pressure index of 3.0 psi.

The hurricane considered in the design has a central pressure index of 26.8 inches of mercury and a maximum sustained wind speed 30 feet above the water surface of 130 mph.

Based on our review of the characteristics of severe tornadoes and hurricanes which have occurred, we conclude the parameters chosen are conservative.

Surges resulting from the probable maximum hurricane have been analyzed by the applicant.

Calculations indicate that wave runup on the east shore of the site will reach elevation 27.4 feet.

Waves impinging on the emergency service water pump house are expected to run up to elevation of 24.3 feet.

To accommo-date this runup, the pumps and other equipment are protected to elevation 30 feet.

Several hurricane tracks have been analyzed to assure that the surge is maximized.

The detailed surge calculations, as well as the hurricane parameters, have been reviewed by the U. S. Army Coastal Engineering Research Center (CERC).

The CERC, in their calculation of hurricane surges, use parameters published by ESSA defining the probable maximum hurricane.

These parameters are to be revised by ESSA in the near future.

When the new parameters are available, CERC will recheck the VEPCO analysis of hurricane surges.

This will be performed prior to issuance of an operating license.

A copy of the CERC comments is attached as Appendix D to this report.

3.1.3 Geology and Seismology Investigations at the site indicate that soil deposits with low to moderate shearing strength extend to depths of 50 to 80 feet and consist of layers of

sand, silts, and clays with interspersed thin lenses of iron oxide cemented sands.

The lower portion contains decayed vegetation and shell fragments.

This formation overlies a layer of stiff clays of Miocene age with occasional sand and silt members which are strong and stable with moderate to high shearing strength.

This Miocene layer is estimated to be 240 feet thick.

Underlying the Miocene layer are Eocene, Paleocene, and Cretaceous sediments.

Crystalline bedrock is estimated to be at a depth of approximately 1300 feet below ground surface.

We have been advised by our geological consultant, the U. S. Geological Survey, that the applicant's analysis presents an adequate appraisal of aspects pertinent to site evaluation.

The report of the U. S. Geological Survey is attached as Appendix E to this report.

The seismicity of the area around the proposed site has been investigated by the Seismology Division of the U.S. Coast and Geodetic Survey.

~ cop~ of the Coast and Geodetic Survey's report is attached as Appendix F to this report.

(~

The Survey concluded that an acceleration of 0.07g at ground surface would be adequate for representing earthquake disturbances likely to occur within t!'te

. _.,.,.: ~

lifetime of the facility, and that designing for an acceleration of 0.15g at ground surface would adequately protect the facility from the maximum earthquake likely to affect the site.

3.1.4 Liquefaction I-In the event of earthquake-induced cylcic shearing of a soil mass, particles can become separated by films of liquid. If this occurs, all shearing strength is lost and the soil behaves as a heavy fluid, possibly resulting in the co,llapse or distortion of structures founded above it.

In order for the particles to

" separate and liquefy readily, soil density must be low, and grain size should be uniform.

Because of the nature of the soil deposits at the Surry site, the potential for liquefaction was investigated in depth.

This investigation included an extensive boring program, placement of piezometers to determine ground water table, in-situ density determinations, and several dynamic tri-axial tests of samples of critical sands at their in-situ relative densities.

This examination revealed that there are three sand members present below elevation +4 feet (El. +4).

These are identified as Sand A, a fine uniform sand extending from*El. +8 to El. -15 with relative densities ranging between 60 and 95 percent; Sand B, a silty layer with a discontinuous thin, well-graded clean sand and gravel member near its bottom extending from El. -25 to El. -37 with relative densities generally between 80 and 90 percent with occasional values as low as 65 or 70 percent; and Sand C, two clayey silty fine sand layers separated by about 2 feet of clay found at El. -55 in one boring.

Location of these sands relative to the structures is shown on Figures S9.12B-1 and 2, Supplement Volume 1.

Sand A underlies the foundation mat of the auxiliary building, the fuel building, and the control area.

To eliminate any potential for liquefaction

. under these structures, Sand A will be removed and replaced by dense graded.

. compacted granular fill.

To illustrate the inherent protection against liquefaction,.the applicant has computed safety factors for the sand layers.

These safety factors, defined as the ratio of the shear stress required to cause liquefaction to the average peak shear stress during an earthquake, are summarized in Table 3.1.4 for the 0.15g earthquake.

These factors assume that the pumps provided at the exterior

e.of the containment to draw down the piezometric level do not operate.

The relative density of Sand B ranges between 80 and 90 percent with only occasional values of 65 and 70 percent.

Nevertheless, the safety factor for initial lique-faction for 60 percent relative density is tabulated for information *

/

Sand Layer Structure Sand A Yard Area Sand B Control Area Auxiliary Building Yard Area Sand C Containment

..

• Table 3.1.4 Relative Density 60

80.

100 60 80 100 60

80.

100 60

.80*

100 60 80 100

$afety Factor for.

Initial Liquefaction 2.5 3.2 3.6 2.0 2 *. 6 *.*

2.9 1.9

,- 2-.A-2.8 2.3

. 2.9 3.4 2.4

• 3*.l C,.' :*,*,, e*

3.5 The ~valuation of the factor of* safety against liquefaction has been'**

re\\T:l.~wed by our coiisuitant, Nathan* M~ Newma~k

• Consulting Engineering* Service~f, and it concludes that the applicant's evaluation is accep.table. -A copy of 'i*fs

., report on the adequacy of *the. foundations' -*as well as. the seismic design of containment and items important to' safety,* is attached as Appendix* G~ *

'* 3.1.5 Hydrology The facility will be constructed with service water intake on the downstream side of the peninsula and discharge on the upstream side to reduce the thermal.

effects downstream from the site.

The potential for concentration of radio-activity in the river due to recirculation of discharge water through the plant has been analyzed, using experimental measurements obtained from James River hydraulic model studies.

These studies have shown that there is no potential for concentration with the system designed as proposed.

There are no public water supplies that could be affected by release of liquid radioactive materials from the Surry units.

There are 29 wells which are used for potable water within five miles of the site with the nearest on the State Waterfowl Refuge.

We conclude that none of these would be affected by radioactive liquids released from the Surry units since the coefficient of permeability of the soil is such that it would take decades for water thus released to reach the aquifer.

The low concentrations of radionuclides anticipated in the plant discharge and the time available for radioactive decay further support this opinion.

In summary, we conclude there is no potential for significant contamination of public or private water supplies in the vicinity of the Surry site.

The U.S. Geological Survey has also considered this matter in its report, included as Appendix E.

3.2 Stress Analyses 3.2.1 Reactor Internals The reactor internals and core will be designed to meet normal design loads of mechanical, hydraulic, and thermal origin plus operational basis earthquake

--- ---------------------------------------~ laads, Stesss limit criteria used are as established in Section III of the ASME Boiler and Pressure Vessel Code with the exception of fuel rod cladding whi,ch is not covered by the *code.

C~adding is designed ip such a manner that the total stress during anticipated transients is less than yield and t:he total strain is less than 1%.

Seismic stresses will be cansidered as_ primary stresses and com-bined with nonnal operating stresses_by vectorial addition.

The reactor internals will also be designed to _withstand the concurrent blowdown and design basis earthquake loads_.

Primary tensile _stresses under such

~-...

load combination will not exceed sfre!=J.s_e~,.ca.rre~ponding ):a 20% _ of: the uniform strain at temperature, while the allowable _deflection limit_s will be about 50%

of the loss-af-function. deflecti9ns fo~ _the specific.co_mponep.ts,. ){e conclude

. -~-

that these_ s~ress a~d deformation limits. provide ~dequ~te,marg~.ns __ af safety.

_During reactor blowdown resul_ting ~rom a lo_ss-of-caolant accident, the sta_ted criterion is that the._ reactor inte~~ls must maintain a cooli:iple heat transfer geometry and must p_ermit insertiop. oJ __ the rod cluster control assemblies to shut down the reactor~,, Tpe,.appl_ic.an~ has calc,ulated deflections expected in key portions of. the reactor internals. an4. compared them t_o estimates of maximum i

• , :~

deflections possible without l;.oss. af functi~~* -. In __ 9tl_l J!ases, the pr~dicte_d deflections are smaller. than t_he values needed to.. assure.performance.

Plese data were calculated assuming a 50-msec break:-time.,

As *r_e':oillIIlended by the _ACRS, we intend to continue to, review_ the __ applicant '.s..calculations for the. effect of blowdown forces on reactor internals.

3.2.2 Reactor Coolant System Section III of the ASME (American Society of Mechanical 1!:p.gineers)_ Pressu7;e Vessel Code will be used to design the, reactor vessel_, pressurizer, coolant pump

~.... '.

' casings, and the steam generator.

The vessel and its internals will be con-structed so as to permit removal of the internals during plant life.

The reactor coolant piping design will be analyzed in accordance with the requirements of USASI (USA Standards Institute) B31.l, Code for Pressure Piping.

The applicant stated that the analysis will take into account the inter-relation of the primary system components, piping, and supports.

A complete stress analysis which reflect~ consideration of all design loadings detailed in the design specification will be prepared by the manufac-tµrer to assure: compliance with the stress limits of Section III of the ASME Code for the reactor vessel, steam generator, pressurizer, and pump casing.

Westinghouse independently will review these stress analyses.

A similar analysis of.the piping will be prepared for Westinghouse by a qualified. piping analysis contractor.

The reactor coolant system, and all other Class I (seismic) mechanical systems, will be designed to withstand normal design loads of mechanical, hydraulic, and thermal origin plus operational basis earthquake loads within normal code allowable stresses.

In addition, Class I systems and componen~s will be designed to withstand the concurr~nt blowdown and design basis earthquake loads.

Primary membrane stresses under such load combination will not exceed stresses corresponding to 20% of the uniform strain at temperature.

We con-clude that these design criteria.to provide adequate margins of safety.

3.3 Engineered Safety Features

3. 3.1 Subatmospheri c Containment Concept

• The applicant proposes to operate the containment normally at a pressure below atmospheric pressure at 10 psia (-4,7 psig).

Following a loss-of-coolant accident, the containment pressure would increase and subsequently decay in a manner similar to that* of' other dry ;containtnent *system's~

However, containment sp~ay systems are provided to rapidiy depressurize the containment structure to below atmospheric pr~SSUrE( following a* loss-~f-cdblant accident.

The significance of. s*ubatmospheric containment is that in the analysis of off-site radiological*consequences of accidents, out:..a.leakage from the containment vessel ceases once containment pressure is.below atmospheric pressure.

Thus, the pressure response of the containment as a function 'of tinie is an.important function in dete!'rmining off;:_site doses~

Details of our evaluation *of the consequences*of a design basis accident*applying this assumption are included iri. the Accident Analyses: section (Section 4~0) of this ~valuation.

A descrip.,..

. tion of the containment vessel' heat removal equipment and out analysls of the methods used' to' determine.the ptes~ure 'arid time respons'e -'.are presented below.

Prior to startup, containment vacuum is established through'use of a.steam air 'ejector;' Two 10.:.scfin niechari.icai' vacuumiptimps' a:ie' provided tb maintain this pressure during normal. operation~ ** *They' discharge; fo the atmosphere through

• • two charcoal' filters connected in parallel.:*, Isolation val~es* in series are provided in the lines ' ftom :. the containment to the suction of the vacuum pumps and 'the air e}ector.

These '1alves clos*e upon reception ~-£ a containment

'isolation signal.

• • Containment *sprays. are provided. to *cool the containment.

Two subsystems are provided.

The'* cont'ainment s~ray' s-yst'em drat.rs' water' froni the refueling' water storage tank and sprays it into the containment. :* The reciirculation.'.:

spray system draws water from the containment sump and.again sprays it to

.. the containment.

These two systems, in combination, will decrease containment pressure to below atmospheric pressure and will maintain this condition for the course of the accident.

The containment spray system consists of two 100% capacity pumps taking suction on the refueling water storage tank and discharging to separate 180° spray ring headers located at the bend line. inside containment.

Each pump will have the capacity of pumping 4000 gpm from the storage tank.

The water in the storage tank will be borated. -In addition, in order to increase the heat

.removal capability of the containment spray system, the refueling water storage tank contents will be maintained at a.temperature no greater than* 45°F by circulation of the borated water through mechanical refrigeration units.

Tank insulation is designed to limit the temperature rise of the bulk contents of the tank to l/2°F per day if the refrigeration unit should fail, assuming a 95°F ambient temperatur.e *.

The applicant wil~ provide a chemical additive to the spray in order to improve its effectiveness in removing iodine.

The chemical additive has not been identified and the choice of the additive is considered part of the research and development program discussed in Section 8.0.

A chemical addition tank is to be lo~ated adjacent to the refueling water storage tank. It is the same height as the refueling water storage tank and shares a common vent with the storage tank.

Piping is arranged so that upon receipt of a spray initiation signal, valves will open which permit the additive to flow by gravity to a mixing chamber located at the bottom of the refueling water storage tank and thence to the suction of the_pumps.

The additive will be stored at a concentration which will guarantee that the spray solution at the spray nozzles will have a minimum pH of 11 to adjust the degree of hydrolysis of iodine in the water phase.

Details of.the calculation of spray effectiveness are discussed in Section 5.5.

Such a solution, is highly corrosive, and materials of construction will be**

chosen to assure that the effectiveness of the spray system will not be impaired.

Our evaluati.on of the iodine removal capability of the spray is presented in the Accident Analyses section (Section 4.0) of this report.

Th~ two containment spray trains will share only the' storage and chemical additive tanks.

In addition, each pump will have dual power sources.

Steam turbines*, powered by steam supplied from the steam *generators, will normally drive these pumps.

If necessary~ however, electric drive motors will be used to operate these pumps.

These electric motors will be powered from the emergency diesel generators should off...;site power be lost *.

The recirculation spray system consists of four-50% capacity independent sets of equipment.

Each set consists of a pump, heat exchanger, and a 180° spray header.

No failure of either active or passive components in one set can disable the other* sets*.

• Two* of* the sets* are located compteteiy within the con-tainment *. These start automatically upon receipt of a high containment pressure signal. If either or botl:i of the inside pumps fail to operate within two minutes, the two outside pumps will be automatically started: 'Th.ese outside pumps are located immediately adjacent' to the outer wall of the containment and are supported by an extension of the bottom mat* of the* containment, thu*s minimizing the length.of the pipe run exte:rior'to containment.

Each pump has a capacity of 3500 gpm.

The*four heat exchangers are*located inside containment.

They are cooled by service.water which flows by gravity at approximately 8500 gpm through each heat exchanger.

These heat.exchangers constitute the only method available to transfer decay heat from the containment structure to the ultimate heat sink following a large loss-of-coolant accident.

We have reviewed calculations of the containment pressure response to the loss-of-coolant accident as a function of time.

These calculations assume function of one of two containment spray systems using water chilled to 45°F, a river water temperature of 95°F, start of one recirculation spray pump three minutes after the sta,rt of the incident, and. starting of the second spray pump five minutes after the start of the incident.

The applicant's assumptions regarding condensation heat transfer coefficients, thermal conductivities, methods of calculation, and transfer of heat from the core ar,e conservative.

Primary system blowdown is calculated using a homogeneous flow model with conservative flow c~effictents applied.

These coefficients cqnservatively do not include the effect of directiQnal changes or piping pressure drop.

Transient heat conduction in the static heat sinks in the containment vessel is calculated using a n~merical method which assumes slab geometry.

.. Internal static sinks are assumed to have b*oth faces exposed to the containment atmosphere.

We have evaluated both the calculational technique and the assump-tions made and conclude they are conservative.

Heat transfer from the core is calculated assuming a realistic power decay with time following initiation of a loss-of-coolant accident and conservative heat transfer coefficients.

We conclude that the model used to determine the amount of core latent heat transferred to the containment is conservative.

Further, we have evaluated the effect of in-leakage of air through the contain-ment str.ucture or penetrations *or by' sneak leakag~ paths through lines penetrating

• containment, In view of the measures taken to prevent in-leakage * (spring-loaded check valves, water seals, etc.) and the large additional volume of gas required to raise the containment *pressure to* atmospheric, we conclude that the contain-ment vacuum will be maintained.

In addition, we have** evaluated the containment design' pressure using a*

model which has been applied to several. other' containment' structures in our ev~luation of.. their adequacy.

We *have concluded as a *result of these analyses that the margin provided b*etween peak press'ure and design pressure compares favorably with'that ~f other facilities.

In our view, the assumptions applied in the prediction of.depressuriza-tion time of the containinent should' remain consist'ar1t with the conservatism :.

which"we presently apply in determining the fission p'rodi:ict source.in the con-

'tainnrent atlll'osphere (fiss1on product releasEf:equiva.ieh.t to that resulting. *from a 100% core meltdown) and in evalu'ating the ~dequ~cy of' the containment struc-tural* design with regard *to maximum temperature' and pressure conditions (failure of the emer:gency core cooling system to function).

• Therefore, the pressure decay characteristics. of the containment have been evalu,a'ted on th~; assumption that the emergency core cooling' sys*tem,. including accumulat:o~s' does not function.

The applicant predicts that tinder the assumption of minimum contain-ment spray and recirculation spray systems'opet'atirig with"no'einergency* core c~oling provided, containment pressure will be reduced below the lowest atmos-pheric p*ressure anticipated at* the site :i.n 3090 seconds, thus terminating* leakage~

Off-site radiological consequences of* accidents using this assumption are dis~

cussed in the Accident Analyses section (Section 4,0) of this evaluation.

Based on our evaluation of the pressure-time history of the containment after accidents, and in view of the conservatism.of the various assumptions, we have concluded that the prediction of the time history of the* containment pressure and of the time interval over which the containment pressure would be greater than atmospheric pressure have been conservatively calculated.

3.3.2 Safety Injection System The safety injection system consists of separate high-head and.low-head subsystems with recirculation capability.

The applicant's design 'criteria 'for emergency core cooling are (1) maximum calculated zircaloy clad temperature**

for the entire core will not exceed the zircaloy melting temperab.ire with 'the core in its original heat transfer geometry, and (2) zircaloy-water reactions will be limited to an insignificant amount. It is our position that the emergency core cooling system (ECCS) should provide all of the following functions:

(a) limit the peak clad temperature to well below the clad melting temperature, (b) terminate the temperature transient before the core geometry necessary for core cooling is lost, (c) limit the fuel clad-water reaction to less than one percent of the total fuel clad mass, and (d) reduce the core temperature and remove core heat until the core will remain covered without recirculation and replenishment of &oolant.

To establish that the proposed system can provide functions (a), (b), and (c), the applicant presented performance analyses of the safety injection system.

In determining the performance capability of the Surry ECCS, the applicant has made conservative assumptions with regard to the more significant parameters, e

as follows:

break opening time (all breaks assumed to occur instantaneously),

reactor coalant pumps trip (lass of n_ormal a. c. power coincident with reactor scram)_, reactor shutdown (minimum void formation model for the void shutdown calcul.ation), blowdown heat transfer (no credit taken for transition boiling),

ves~el ~ater level (no credit taken for bailing froth-height), and core heat tr~~.f.er dut:ing r~flooding (uniform coefficient of 25 Btu/hr-ft2-°F) *

...,.. Based on our l)resent understanding of _the blowdown 1:1nd core heatup

\\

phenomena, we conclude that the codes used-by the ap~licant conservatively predict the course of a loss-of-coolant.accident.

.,.. ~e. applic~t

• has presented results of. blawdown and core heatup analyses 2

. 2 2

~er.,the double-ended (9. 2 ft*), 3 ft., and 0. 5 ft breaks in the cold leg of_

o~~' __ p.f the. reactor coolant loops.

The-cald leg breaks result in higher peak clad ):emperatures than the corresponding hot leg breaks because of core flow "reyers_a_ls d~ring blowdown and steam binding above_ the core during accumulator injection;.. l>oth of these effects were considered in the applicant's analyses.

The.cere*thermal transient has been calculated assuming that two of the

_ tllree.. accumu+at;ors discharge to the core, Blld that minimum EC~S. functions (one of

_the three. h_igh-head pumps and one of. two low-head safety injection pumps).

Peak cl~d.temperatures of 1800°,. 1760°., and 2280°F have been predicted for the 2

. 2 0.5 ft, 3 ft, l:µld.double-ended cold leg.. breaks, resp~ctively.

We will*.review the analysis of ECCS performance for small. breaks (< 0.5 ft 2) including calculations of fuel perforations.and the ability to maintain.suit~ble core geometry for a spectrum of ~reak sizes at the operating. license stage of our review.

Preliminary data indicate that the cladtemp~ratui;e1;1-for_small breaks-will be no higher than those noted above.

Requirement (d) for long-term cooling is satisfied by the redundancy of low-head pumps and ECC recir'culation subsystems as described below.

Also, there is sufficient.water in the refueling water storage tanks to ensure long-term coolant availability for the recirculation cooling mode.

The safety injection. system proposed differs from that proposed for other

.recent licensed PWRs supplied by Westinghouse in the following areas:

(1) the cQ.arging pumps are also used as high-head.safety injection pumps, (2) the residual heat removal system is n~t used in the recirculation system and no direct means

.. qf transfer of heat from the core recirculation system to a heat sink is incorporated in the design, (3) the portions of the recirculation system exterior to. containment are all supported by an extension of the containment mat and are located immed,iately ~djacent to the exterior wall of the containment, minimizing the length of pipe runs, and (4) all piping from the sumps to recircu-lation pumps enters a valve pit which is designed in such a manner that piping failures will not seriously impair system performance.

These differences are discussed in_ detail below.

- 1, Upon initiation of a.safety injection signal by coincident low pressurizer level and low pressurizer pressure*signals or 2/3-containment high pressure signals, the suctions of the charging pumps are automatically switched from-the volume control tank to the boric acid*tanks and the discharge is aligned to the safety injection.lines rather than the normal charging lines.

Upon reception of a low boric acid tank signal, the boric acid tanks are auto-matically isolated and suction :f,s transferred to the :refueling water storage tank.

Valving is provided in series or parallel, as required, to prevent any single failure of a valve to operate from impairing the capability of the safety injection system.

2.

There are no heat exchangers in the core recirculation,system.

Decay heat generated in the core is t_ransferred to the s*afety injection water which spills through the pipe break. to.the c*ontainment floor.

Sump w*ater is cooled

• by _the recirculation spray heat exchangers and. is sprayed throughc: the. -contain-ment atmosphere.

In the recirculat_ion* mode of operation; sump water* is *drawn..

• outside containment and injected into. the co+d.. legs J>f the,_primary i;ystem with,

no further transfe:r o*f he~.t.

The recircul_ation: spray pumps* must operate for,.;__:

long periods of time following an accident to maintain t;he containment atmosphere at a negative differential pressure.

Utilizing_ the heat_ transfer-,capability.ofJ:

the recirc;ulation spray system. imposes ;no significant addi.t.ic;,,nal.duty, on... the...

system.

Further, use of the spray, system for* heat rejection wil_l not.create

-~

.. ***~..

any major difficulties in cleanup operations,following. an accide_nt sinc;e the,,... '.::

system can be :lsola~ed for_ a significant* period of time after. several, months... of..

• fiss;lon. product decay since. the heat, capacity ot. the large: yolume. of: water):,e~ng circulated can be used to store core decay heat.

In add;l_tion,

• dep_E[md:f,.ng o~..

the location and nature of the pipebr_eak, it,may_.also;be_ p9~siple to __ utilize the normal residual heat_ removal equip_men~ for diac::ay hea.~~

• For th.e~e reasons,...

the system as propos~d is adequate *.

,.* "}

3 *. The. recirculatiqn pumps-exteric;,r to qont~inmeJ:?.t are-all supported on an extension of the., containment mat.... They are.. loc~t~d. _ipnne_dia~ely.)ldj acent to th!!

outer wall of the containment in a concrete structure between the containment

• i wall and the permanent-cofferdam.* Rup:t:ur_e. o-f any lin~; lea~ip.g "f;om_ the sump to the recirculation valves or of :that portion of.tl}e line;'from th~.refueling

t water storage ta..~k to the suction of the low-head safety injection pumps within the valve pit will fill the lower portion of the sealed valve pit.

The water level which would result from such a rupture provides sufficient NPSH for the pumps to remain in operation.

Thus, the valve pit would function as a surge tank.'and* system operation would not be impaired.

Failure in the ECCS downstream of the.isolation valves would be detected and*isolated.

The dose consequences of a failure in the exterior recirculation.system are discussed in the Accident Analyses section (Section 4.0) of this evaluation.

3. 4 Thermal Shock on Reactor Vess.el The applicant has presented some results of calculations on the effect of the.hot pressure vessel being deluged with cold water after actuation of the core cooling system.

Ductile failure, brittle failure and fatigue failure were the modes considered in* the analyses.,Brittle failure analyses have been con-ducted using both the transition temperature approach and the fracture mechanics I

approach.

The preliminary results of these analyses indicate that no loss of vessel integrity would result from the transient.

The ductile failure and fatigue failure analyses have not yet been completed.

As has been recommended by the ACRS, we are continuing our evaluation of the calculational models to ensure. that sufficiently conservative assumptions have -*been used in the analyses by reviewing periodic progress reports *.of. the work in this area. If

• the com-

. pleted calculations indicate that vessel integrity may be jeopardized as a resuit of the injection of cold water, we will require the applicant to provide means of reducing the thermal stresses by adding protective features to the vessel and its internals.

e 4.0 ACCIDENT ANALYSES

• ~.,<

• , )

The applicant has analyzed the. consequences of sev~r~*~.a:~c.idents.whic.h.

might result. from ina.dverte11t react,iyitr additicms, o.r eq.µipmet1,t malfunctions and failures.

These analyses cov;er.. the,range of typ:i,cal pptential accide.nt,:s

~

which could occur at the fac:t,lity. *.*. :T~e>se acc~.det').ts wh~c~. ~re.,of part:icula!':**

iQ.terest from, the standpoit1.t of off-.si.t~;,,consequen.G_,~s !.,Pr.. ~hi.ch,.ar_e. cr~4ib}~

because of unique features. of the stat;ton design ar~,Aiscu~sed* b~low.<Y 4.1 Startup of an Inactive Loop

~:.: ' ' '{...

,* '<~..,

As previously stated, the loops of the p~}.m~ry !,lys;em are _equ~PP:.~iW:tf~.

isolation valves.

Since a main coQ_lant pump can.. pe secm~ed.and s.i,nce1 reverse flow would be prevented by the stop valves, the poten,tia:J._cons,eque~ces... ~i,A*-

.J,*

cold water injection accident.are more. s~v~r;e tllaµ.: in*,. a PWR s:r1;1t~m in. whfch~,

no isolation valves are provided.

To preyent,c:,old water... :i.njectiot1,_ valye., and pump interlocks will be. provided whi,ch.will (1) prevent. _sta_rting a.. pump._un.less

~

i

~ * '

the loop cold leg va~ye ii;, closed., and (2). pr,ev:~nt,. o,pen,ing., of.the colf.. leg,...

valve unless the loop temperature. is., su:ffi.c:,ieµ.tly high tp prevent. e~~essive reactivi_ty changes.

These.interlocks. will be d.es,igne\\!A.,1=0 prevent the.q~cur-::-

I

~ *.'.

rence of DNB in the core_ as a result of a cold wat.er,slug ent.ering. the core,

).*

t, *

• I*

a:,f t:~r loop startup.

-,. -~~.:

' ' ~*-

We have also considered the pos_si~ili ty of_ ~he. exis te_nce. of, ~o~~urrent.'

low boron concentratton in the isolated. loop..

• If,lopp isplation.v1ere.<.:.iJ

... ' -. ~-*

occur near the end. of core life and_ continue through a. ~efueling operat~on,,..

b

'~ **

it, _is.concei~a'ble, the>ugh_ highly unlik~:J.y,.that,t;h~ i_l!lolate,~ *. 1001' could_ b.e *:":

unborated with the remainder of* the primary* system f!llly. pe>ra,ted. :, ~~ this

e cold, unborated water were rapidly injected into the core inlet plenum, a severe reactivity excursion could result.

Reliance is placed on admini-strative procedures to prevent this occurrence.

These procedures require that prior to re~start of a loop, the following be performed in order:

1.

The active portion of the. primary system be borated to the extent that.sub-criticality can be maintained in the event of startup of an unborated loop.

t.

The boron concentratien of the iselated loop be verified to be at least equal to the active pertion of the primary system by two independent

.. analyses.

3.

The-isolated loop temperature be at least equal to that of the hottest

• active loop.

4.

The pump in the isolated loep be stopped prior to epening of the last isolation valve.

5.

The isolation valves be opened to permit reverse flew until mixing of loop water with that of. the active portion of the primary system is assured.

6.

The iselated loop coolant pump be started wit_h the discharge valve closed when returning the loop tp service.

We have evaluated the censequences ef failure of the operator to observe portions of these procedures.

Since items (3) and (6) above*are interlocked as discussed above, violation of these procedures will *not result in any perturbation of the core*reactivity status. A.potential reactivity insertion exists enly if both items (1) and (2) abeve are violated.

If only these two

r violations are considered, a reactivity insertion is generated by the slow dilu-tion.of primary system water by reverse flow through the inactive loop after the isol.ation valves are opened.

The rate of dilution will be determined by the rate at *:which. the isolation valves are opened and the flow characteristics of the loo'p during reverse flow.

The reverse flow under these conditions has not yet been determined a11:d, therefore, the reactivity insertion rate caused by dilution cannot be stated.

The design criterion will be that no DNB will occur from.the maximum reactivity insertion resulting from violation of any two of these six*administrative procedures.

We conclude that. this objective will he met during detailed system design.

As recommended by the ACRS, we will continue to review the detailed design of the protective instrumentation andthe*scope of the administrative procedures.

4.2 Steam Line Rupture The power level and flux distribution attained following the rapid.primary I

system: cooldown resulting *from a steam line rupture are functions 6'£ ro'd' patterns and their prediction requires*detailed physics *parameters.

The applicant's calcu-lations indicate that the most severe transient would occur if the break were to occur with the reactor at. hot standby.

• For breaks exterior to* con*tainment, down-stream of the flow restrictor, the maximum power l~vel which would occur is 30 percent o*f full power, with no DNB occurring.

With breaks inside containment, upstream of the flow restrictor, calculations predict a maximum power level of 52 percent would be attained,*coupled with approximately.10 percent clad per-forations.

A rupture inside containment would result in venting of the*steain generator to the containment structure.

Since VEPCO proposes to operate with

1 a maximum primary-to-secondary leakage of 10 gpm, release of fission products to the containment would result.

The stored energy in one steam generator is sig-nifi.cantly less than that in the primary system; and the fission product release is much lower than that assumed for the design basis accident.

• Thus, off-site doses resulting would be substantially below those of the design basis accident.

Fo:t' breaks outside containment, no fuel damage is anticipated.

However,*

because of the allowable primary-to-secondary leakage, release of radioactivity entrained in the primary system to the atmosphere would occur.

The effect of this*leakage has been considered, assuming (1) instantaneous secondary system depressurization to atmospheric pressure, (2) linear primary system pressure decrease to 350 psia in four hours, equivalent to a cooldown rate of 50°F/hr, (3) pressure constant at 350 psia for the following four hours, (4) primary pump shutoff after eight hours with the primary system depressurized to atmos.;..

pheric pressure, (5) leakage rate proportional to the square root of the primary-to-secondary differential pressure, and (6) equilibrium activity in the secondary system prior to the rupture resulting from a 10 gpm primary-to-secondary system leakage rate with a primary system.fission product inventory equivalent to that resulting from 1% failed fuel.

We have evaluated these assumptions and COll,clude that they are conservative.

With these assumptions, 4.6 percent of the primary system inventory would leak from the primary system and flash to the atmosphere during the eight-hour period. Off-site doses (54 rem thyroid and 50 mr whole body doses) would remain well below the 10 CFR 100 guidelines even if all iodines flash.

4.3 Steam Generator Tube Rupture The applicant has analyzed the consequences of a steam generator tube rupture assuming isolation of the condenser after reactor trip. This trip would occur approximately six minutes. after occurrence of a rupture and result from a safety inject.ion system actuation signal. Prior to reactor trip, steam generated*. in* tbe steam generator,

• together with entrained fission products, would pass through the turbine to the condenser.

Any.gaseous fission products' which leak.t;o the secondary system would be drawn from the condenser by the

• air ejector.

Upon air ejector high activity alarm, air ejector effluent would be *aut:omatica1iy *c:1::i..verted to.the contaimnent'. Thus, activity released from the pl."imary ~y~tem would be prevented from :reaching the atmosphere.

If.it is.*assumed that off-slte p*otier. is iost concurrent with the tubcf.

rupti:i:re, the circulating water pumps stop and the condenser-and air ejectors are>isoi~t~d.

Decay heat.removal is accomplisheci by boiloff of.emergency feedwater in the e;teain ge~e.rators*, with relief of the steam produced to the atmosphere.

During this period, the.steam generator experiencing the ruptur~d tube -can be identified through use of the steam generator blowdown monitor and observation of steam generator levels.

Once the faulted steam generator is identified, :Lt cari be'isoiated using the primary system.isoiation valves, thus terminating leakage in a short period of time.

The applicant has estimated that the defe~tive.steam generator can be detect.ad and isolated within 10 minutes.

The off-site. ~onsequ~nces have be.en evaluated assuming 30 minutes 'ialapse before isolation.

We conclude that this t'i~e interval is conservative in view *of the nature of the sampling and clos~re operations which must be.performed.

During this 30-mi.nute J>eriod, less than 20% of the primary system f°luid-will be trans-ferred to the secondary side of the steam generator*.

The applicant has performed a preliminary analysis of the consequences of a tube rupture assum;i.ng an iodine partition factor of 6~5 x *10-:1, a *fission*

• product.inventory equivalent to 1% failed fuel, and a 30-minute leakage period.

With these assumptions, a site boundary thyroid dose of 0.4 rem has been calcu-lated.

We conclude that the off-site doses will be well below the gui9elines of 10 CFR 100.

At the operating license stage of our review, we will use the consequences of this accident to obtain primary coolant activity limits for the

. Technical Specifications to ensure that the resulting site boundary doses are within acceptable limits.

4.4 Design Basis Accident The design basis accident: analyzed results from a major loss.,.of-coolant accident coupled with a release of 100% of the noble gases, 50% of the halogens, and 1% of the volatile selid fission products.

As.previously stated, the Surry containment will normally operate at a pressure of 10 psia.

Since the contain~

ment pressure will be brought to subatmospheric conditions by the spray systems at same time following a loss-of-coolant accident, out-leakage ceases when the containment pressure falls below atmospheric pressure.

The design specification of the containment structure is a leakage rate of less than 0.1% of the contained free volume per day when pressurized to the design pressure of 45 psig.

However, since VEPCO has proposed a leakage requirement of 0.125%/day at design pressure, we have evaluated radiological consequences assum-ing this containment leakage rate, a fission product release as stated above, a leakage duration of 3090 seconds, and meteorology as discussed in Section 3.1.2.

The choice of leakage duration time is discussed in Section 3.3.1. With these assumptions, we calculate the thyroid dose at the site boundary to be 526 rem and the whole-body dose to be 3.6 rem if no credit is given for iodine removal by the

e spray.

Thus, a reduction factor of 1.75 in the thyroid dose is required to meet the.10 CFR 100 guidelines at the site boundary.

Since leakage is assumed to cease after 3090 seconds, the 30-day doses at the low-population zone distance of three miles are well below the 10 CFR guidelines without credit for halogen removal.

These doses are 4.6 rem to the thyroid and 0.4 rem whole-:-body.

As _previously discussed, a chemical additive will be present in the spray solution which will raise the pH to* *a minimum value of 11.

We have evaluated the calculational technique and the assumptions made by the applicant in perform-

. -.. *' :* ~-* -.,:'

._*.. :*~....,.

ing his analysis of the degree of effectiveness of the spray in removing halogens from-the containment atmosphere and consider them to be conservative.

In addi-tion, all available experimental evidence on the effectiveness of chemical additive sprays indicates that some dose reduction factor is attainable.

On

• ' '-? *~.

these bases, we.have confidence that a reduction of at least 1.75 will be readily attainable. Nevertheless, space will be provided in thecontainment'fer

'. *** {."~.

• _1 ::.

charcoal filters, sh~uld the spray development program indicate that the 10 CFR 100 guidelines cannot be met-using the sprays.alone.

Accordingly, we conclude that the effectiveness of the spray system will be adequate to ensure that there

, I

~

is reasonable assurance.that the guidelines of 10 CFR-100 will not-be exceeded.

r.. :.*

**' ~

,. 5.0 QUALITY CONTROL The quality control program will be conducted by the Stone and Webster.

Engineering Corporation (Stone & Webster). It is described in Supplement Volume 2 and summarized below.

The field inspection organization will be supervised by the Quality Control Engineer.

He will be under the direction of the Quality Control Manager at the Stone & Webster headquarters and will be directly accountable to the Supervisor of Design and Construction for the Virginia Electric and Power Company.

The Quality Control Engineer is not responsible to either the on-site Stone & Webster general superintendent or the Construction Manager at Stone & Webster headquarters.

_ The Quality Control Engineer will have a staff consisting of competent civil, structural, mechanical, electrical, and welding inspectors.

In addition, all*quality control testing laboratories will report to him.

The field inspec-.

tors will have at least five years of experience in their respective fields.

These inspectors will have the authority to order supervisors or general foremen to halt work and repair faulty workmanship, replace material not to specifica-

tion, as well as refuse acceptance of manufactured equipment that fails to meet specification.

The original of all test reports by the inspectors or by indepen-dent inspection organizations will be submitted directly to the applicant with copies provided to Stone & Webster.

The shop inspection organization will consist of inspectors located at eight district insp"ection offices throughout the United States and controlled by a Manager and Chief Shop Inspector at Stone & Webster headquarters.

The Chief Shop Inspector will have a direct line of conununication to both the Quality Control Engineer and to the Supervisor of Design and Construction for the Virginia

N Electric and Power Company.

The shop inspectors will ensure that equipment and materials conform to their specified requirements by means of inspections con-ducted at the manufacturer's plants.

They will also accompany the applicant to witness cirtical inspections.

Shop inspection reports will be issued after equipment has been accepted.

The originals of all reports will be forwarded to the applicant with copies provided to Stone & Webster.

We conclude that (1) VEPCO will exercise adequate control over the quality control organization proposed and (2) this program is adequate to ensure that the plant will be constructed in accordance with the application.

~ -*

• 6.0 STATION DESIGN WITH RESPECT TO THE 70 GENERAL DESIGN CRITERIA In November 1965, the Commission published its General Design Criteria for.

Nuclear Power Plant Construction Permits~ and on July 11, 1967, published in the FEDERAL REGISTER its revised General Design Criteria taking into account comments received on the initial c;iteria and further development of the criteria by the regulatory staff.

The applicant has stated that it will comply with all criteria.

We have evaluated the application for conformance with the revised criteria and have concluded that except for General Design Criterion 22, the proposed unit conforms with the revised criteria.

Recognizing that the proposed revised cri-teria may be modified, we will review the proposed unit at the operating license stage in light of the criteria as formulated at that time.

The capability of satisfying Criterion No. 22 which relates to separation of control and protection instrumentation systems was discussed at length with the applicant and commented on by the ACRS.

As recommended by the ACRS and discussed in Section 2. 5 of this evaluation, we will c*ontinue to review the design of this portion of the plant to assure that the proposed protection system is modified to eliminate or reduce to a minimum the interconnection of control and protection instrumentation.

• 7.0 RESEARCH AND DEVELOPMENT In Amendment No. 10 to the application, the applicant has indicated the scope of the research and development (R&D) programs to be conducted during the completion of the design of the Surry_ facility.

These are summarized below:

1.

Chemical Additive and Containment Spray -- VEPCO will rely primarily on the Westinghouse developmental work in consideration of chemical characteris-tics, iodine removal characteristics, materials compatability, and radi-olysis. Experimental investigations of the relationship of absorption rate to containment atmospheric conditions, the effects of process variables on spray nozzle performance, the extent of radiolysis, and the nature of the radiolytic products are being conducted by Westinghouse, Oak Ridge National Laboratory, and Battelle Memorial Institute.

2.

Emergency Core Cooling System -- The applicant's program consists of develop-ment of system design and the improvement of the analytical techniques used to calculate the core thermal response following a loss-of-coolant accident.

R&D effort will be done to confirm assumptions made in such areas as blow-down forces, heat transfer during blowdown, fuel rod perforation, fuel distortion or swelling, metal-water reaction, and retention of molten fuel.

3.

Core Thermal,Hydraulic, Nuclear, and Mechanical Design Properties Developmental work to be performed includes the following:

(1)

Evaluation of the reactivity worth and effect on power distribution of borosilicate-glass burnable poison.

(2)

Evaluation of the mechanical performance of the burnable poison material and rod configuration by in-pile testing.

• (3)

Determination of operating program for the partial length.rods.

• (4)

Evaluation of DNB ratios for normal and transient conditions to assure they meet design criter~a.

(5)

De:velopment of refined methods of determining Doppler effect... :'.

(6)

Ev,aluation of the effects of blowdown forces on reactor internals.-

A schedule of the research and development program is presente~ below:

1.

Development of the design details of the containment spray.

a.

System design details including piping layout, pipe sizing, and component selection, to be completed by mid-1968.

b.

Selection of spray additive including studies on the effects of the accident environment on.the spray additive and its performance, to be completed by mid-1968.

2.

Development of the design of the emergency core cooling system.

a.

System design details including piping layout, pipe sizing, and component selection, to be completed by mid-1969.

b.

System performance evaluation including parametric studies on performance and the rod burst program to determine fuel rod behavior under simulated loss-of-coolant accident conditions, to be completed by mid-1969.

3.

Development of the final core thermal, hydraulic, nuclear, and mechanical design parameters.

a.

Development of parameters to be completed by late 1969.

b.

Use of part-length control rod -- analytical work to be completed by late 1968, in-reactor work to be completed by mid-1969.,

c.

Use of burnable poison rods.. Critical experiments are completed.

Reactor work to be completed by* late 1968.

d~

Effects of blowdown forces.

Analytical work to be completed by the

• end of 1969.

• We conclude that the development programs, including those proposed by the applicant and those being pe~formed at otherfacilities, aie reasonably designed to resolv*e any *safety *questions associated 'with the abo.Je -feature's of Surry Power *stiation and te provide the data necessary to 6onst+/-ucit the. Surry Station in accordance' with the. criteria and specifications. set forth*. in' the i>sAit

• .*. ~-1.

• 1' _*

~-

• 8.0 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS In a letter to the Commission, dated April 29, 1968, the Advisory Com-mittee on Reactor Safeguards (ACRS) reported on the proposed Surry Power Station Uni ts 1 and 2.

A copy of this letter is att.ached as Appendi~ B.

The letter con¢ained a riumber of comments and recommendations noted in appropriate sections of this evaluation.

The ACRS also noted that those matters identified as warranting careful considerati0n with r~gard t.o recent reactors of high power

. d_ensity. and other matters of significance for all large, water-cooled,. power reactors in its letter to the_ Commission on the Dia~lo Canyon facility apply similarly to Surry Station Units 1 and 2.

These items.have been discussed in this evaluation and will. be resolved to the satisfaction of the staff and the ACRS prior to ~he issuance of an operating license *.

The ACRS letter concluded, "The Committee believes that, if due considera-tion is. given 1;0 the foregoing items, the nuclear units ~reposed for the Surry Station site can be constructed with reasonable assurance that they can be operat~d without undue risk to the h.ealth and safety of the public."

9.0 TECHNICAL QUALIFICATIONS We have reviewed* th~ *appfrcatioh"with-respect.to the'* adequacy of the technical qualifications 'of VEPCO and its contra~tors!.. The.eji;*e~u'tion of the

';Surry P~wer 'st.ati'on p'roject is th*e :sole res~onsibility. o"f° the Virgi;ia Eiectric

• and Power Company/ The. applicat1t: ha:s' had previous *~uclear experience through

• its. partfcipa,tion in. Carolina~* Virginia. Nuci'ear Power*'Associates' Inc.,- ~hich operated the Carolinas Virginia: T~be 'Reactor. cc:HtrR). _. Severai VEPCO.' e~ptoy~~s were assigned *to CVTR on a resident: b~asis*.

";!" t.:

• VEPCO has 'also instituted a :traini~g* program fol' k~y* super~isory* i,er~6il.nel operation.

\\ *,

• I*

_'. 1;~

Operators and operating* supervis*ors** will undergo. an extens'ive train-

. ing pro*gram consisting.hf basic classroon{ lraihirig ih the.o'ry,:., aitual opeti.Ji.ng experience at a pressurized'water read:*ort aild *bii.-stt'e*'traihing as 'stiticif:-aon-structicin is completed~

• 'VEPCO* has* erigag~d Stone & Webst,.er Engineedhg:*cb'rporation as: its'*4gehtfor engirieedng and constructlon,' and has contiactid with 'Westingliou~~-' El~i::ttfc

• Corporation foi:: furnishing' the ~u-cle~r* stea~, s\\ippi.y** syst~Dis, **,*£\\led: \\ ahi(.bitbi'.ne generators.

Stone & Webster Engineering Corporation is an engineering construction firm serving the electric utility industry in the design and construction of power stations. It has been associated with the nuclear industry since,1942 and has provided varied design and construction services at the Shippingport Nuclear Station and has had major responsibility for engineering and construc-tion in*the Yankee Atomic Electric Plant, the Carolinas Virgi~ia Nuclear Power Associates Plant, and the Connecticut Yankee Atomic Power Plant. It is also

• providing construction management for the Nine Mile Point Nuclear Station which is currently under construction.

The Westinghouse Electric Corporation has accumulated considerable experience in the nuclear industry.

It has designed and constructed several pressurized water nuclear power reactors that have been licensed by the Com-mission. *The more recent of these include the Diablo Canyon Nuclear Power Plant, H.B. Robinson Unit No. 2, San Onofre Nuclear.Generating Station, and Connecticut Yankee Atomic Power Plant.

VEPCO has also retained the following consultants:

Dmnes & Moore, Inc, Geology Pritchard & Carpenter Associates Hydrology and Ecology Weston Geophysical Research, Inc.

Seismology NUS Corporati,on Meteorology, Climatology, and General Nuclear Consultation The above contractbrs and consultants are recognized to be competent in their areas of specialization.

On the basis of the above considerations, and based upon our evaluation of the responsible personnel and.of the quality control organization discussed in Section 5.0, the applicant is qualified to design and construct the proposed facility.

e

,..56....

10.0 COMMON DEFENSE AND SECURITY The application reflects that the activities to-be conducted would be within the jurisdiction of the United States and that all of the directors and principal officers of the applicant are American citizens.

We find nothing in the application to suggest that the applicant is owned, controlled, or dominated by an alien, a foreign corporation, or a foreign government.

The activities

• to be conducted do net involve any restricted data, but the applicant has agreed to safeguard any such data which might become invelved in accordance with paragraph 50.33(j) of 10 CFR Part 50.

The applicant will rely up'on obtaining fuel as it is needed from sources of supply available for civilian purposes, so that*no diversion of*special nuclear material from military purposes is invelved~

For these reasons, and in the absence of any information to the contrary, the activities to be performed will not be inimical to the common defense and security.

11.0 CONCLUSION

S Based on the proposed design of the Virginia Electric and Power Company's Surry Power Station Units 1 and.2; on the criteria, principles, and design arrangements for systems and components thus far described, which include all of the*important safety items; on the calculated potential consequences of routine and accidental release of radioactive materials to the environs; on the scope of the development program which will be conducted; and on the technical competence of the applicant and the principal contractors; we have concluded that, in accordance with the provisions of paragraph 50.35(a),

10 CFR Part.50, and 2.104(b), 10 CFR Part 2:

.(1)

The applicant has. described the proposed design of the facilities,,

including the principal architectural and engineering criteria for the design, and has identified the major features or components for the protection of the health and safety of the public; (2)

Such further technical or design information as may be required to complete the safety analysis and which can reasonably be left for later con-sideration, will be supplied in the final safety analysis report; (3)

Safety features or components which require research and development have been described by the applicant and the applicant has identified, and there will be conducted, a research and development program reasonably designed to resolve any safety questions associated with such features or components; (4)

On the basis of the. foregoing, there is reasonable assurance that (i) such safety questions will be satisfactorily resolved at or before the latest date stated in the application for completion of construction of the proposed facility, and (ii) taking into consideration the site criteria con-tained in 10 CFR Part 100, the proposed facility can be constructed and operated

• • at the proposed location without undue risk to the health and safety of the public; (5)

The applicant is technically qualified to design and construct the proposed facility; and (6)

The issuance of a permit for the construction of the facility will not be inimigal to the common defense and security or to the health and safety of the publi~.

~,.,

'. *~;.*.... '

... *~, :.*.

. --~}

APPENDIX A CHRONOLOGY REGULATORY REVIEW OF THE VIRGINIA ELECTRIC AND POWER COMPANY

1.

March 20, 1967

2.

April 27, 1967

3.

May 15, 196 7

4.

May 18-19, 1967

5.

June 2, 1967

6.

June 7, 1967

7.

June 21, 19.67

8.

June 30, 1967

9.

July 5, 1967

10.

July 17; 1967

11.

August 4, 1967

12.

August 14, 1967 Submittal of Preliminary Safety Analysis Report and License Application.

Meeting with applicant to discuss s.chedule of regulatory review of application.

Request to applicant for additional financial data, Meeting with applicant to discuss site characteristics and plant design and containment structural design.

Meeting with applicant to discuss pressure response of the containment atmosphere to a loss-of-primary-coolant accident.

Meeting with applicant to discuss site seismicity, geology, and seismic structural design.

Submittal of Amendment No. 1, revised pages to be substituted in application.

Meeting with applicant to discuss instrumentation and control, electrical power systems, and containment isolation criteria.

Submittal of Amendment No. 2, revised pages to be substituted in application, and additional financial data.

Request to applicant for additional information on reactor design, reactor coolant system, engineered safety features, auxiliary systems, steam and power conversion equipment, waste disposal system, radiation and monitoring protection, and loss-of-coolant incident containment pressure analysis.

Meeting with applicant to discuss accident analysis.

Request to applicant for additional information on site and environmental considerations, structural design excluding containment, containment isolation, containment structural design, instrumentation and control system, and electrical power.

13 *. August 24, 196 7

14.

September 5, 1967

15.

September 27, 1967

16.

October 6, 1967

17.

October 27~ 1967

18.

November 24, 1967'

19.

December 7, 1967

20.

December 8, 1967*

21.

January 4, 1968

22.

• January 19*, 1968 :.--

23.

Janum7 23 2l1, 1968 24., '-February -14, 1968

• Submittal of Amendment No. 3, information-on site and environmental considerations in reply to the AEC request of August 14, 1967.

Site visit by the*staff and ACRS Subcommittee meeting.

Request to applicant for additional information on safety* analysis* and- -site considerations'.~

Submittal of Amendment No. 4; answers to ACRS's request for additional information of July*:17;,.*1967*'

and report on Foundation Dynamics by Robert B.

Whitman, MIT~

Meeting with applicant to discuss Amendment No~,4

. to PSAR *

. Request to applicant for additional'* information Lon.'

reactor design, engin*eering and safety features, and adequacy of proposed site.

Submittal of Amendment No. 5, in response to the AEC_

request of Septeinber,27, 1967.

-Submittal-of Amendment No. 6, in response. to-AEC

• requests* of August -14, 1967 and September 27, 1967.

Submittal of Amendment No. 7, in response to AEC letter of November 27, 1967.

Revised'. pages to:.be substituted in application *.

• Submittal of* Amendment No. *a* which*contains answers to DRL requ~st for additional information.

Meeting with applicant to discuss seismic design core internals

• and engineered safety fe:atures., th.ermal shock, containment structural design, and engineering

• safety features, flood potential, instrumentation and control, electric power, safety analysis, etc.

• Submittal of.Amendment No. 9 which contains answers

• to DRL request for additional information.

• 25.

March 1, 1968 Submittal of Amendment No. 10 which contains revised pages to be substituted in the application.

26.

March 18, 1968 Submittal of Amendment No. 11 which contains revised pages to be substituted in the application.

27.

March 26, 1968 ACRS Subcommittee meeting.

28.

April 4, 1968 ACRS meeting,

29.

April 29, 1968 ACRS letter to Chairman Seaborg on VEPCO.

APPENDIX B e

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON, D. C.

20545 Horiorable Glenn T. Seaborg Chai:nnan April 29, 1968 U.S. Atomic Energy Commission Washington, D. C.* 20545

Subject:

REPORT ON SURRY POWER STATION UNITS 1 AND 2

Dear Dr. Seaborg:

At its ninety-sixth meeting, on April 4-6, 1968, the Advisory Committee on Reactor Safeguards comPleted a review of the application by the Virginia Electric and Power Company for' authorization to*construct two nuclear units at its Surry Power Station in,Surry County, Virginia.

This project.had previously been considered at Subcommittee meetings at the site on September 5, 1967 and in -Washington, D. C. on M~rch 26, 1968.

During its review, the Committee had the benefit of discussions with represent-atives of the Virginia Electric and P,ower Company and their consultants, the Westinghouse Electric Corporation, the Stone and Webster Company and, the AECRegulator}' Staff and their consultants.

The Committee also had the benefit of discussions with its own consultants and of the documents listed.

The Surry Station site comprises approximately 840 acres, located on a small peninsula which juts into the James River; 4.7 miles northwest of the nearest corporate limit of Newport News, Virginia.

Newport News has a popuiation of approximately 114,000, located from ten to twenty miles southeast of'the site. Wiliiamsburg, Virginia, a major tourist' attraction, is located seven miles north of the site. The region surrounding the site i~ rural and agricultural.

Surface deposits at the site consist of layers of sand, silts and clays ranging in thickness f:rom,approximately 50 to 80'feet.

Below this are Miocene, Eocene, Paleocene and Cretaceous sediments extending to bedrock, about 1300 feet below grade.

The reactor buildings are to be founded on ten foot thick, reinforced 'concrete mats, supported on the Miocene deposits, approximately 70 feet.below the surface.

The fuel building, between the reactor buildings, is supported on concrete-filled piles driven into the Miocene deposits.' The auxiiiary building and control room area are sup-ported,on four foot thick, reinforced concrete mats, about 36-feet below grade.

e Honorable Glenn T. Seaborg

... 2 April 29, 1968 The Surry Station units are to be identical, three-loop, pressurized water reactors operated at maximum power levels of 2441 MWt.

With respect to core design and other features of the nuclear steam supply system, the reactors are similar to the Diablo Canyon reactor.

The units have a power level and average heat flux about 16% higher than the H.B. Robinson reactor with a power density a little less than that of;th~ Diablo Canyon reactor.

Each of the primary system loops is equipped with two valves to enable isolation of the pumps and ~team generators fo;r purpos~s of mainten:ance.

Further consideration of the instrumentation and.administrative pro~edures proposed for protection against. p()teµ_tial.. reactivity *transients.. initiated by the introduction of cold and/or* ~~borated water into the core from a previously isolated loop may be appropriate at the operating.,licemse *review stage.

In connectioµ with postulated loss-of-coolant, acciclents,, the appl~cant stated that, using conservative assumptions and.afiowing appropriately

. f~~ fuel element distortion from the original* core geometry' the emergency core-cooling. systems. will.be designed to keep. f~el:.:clad temperatures below the -poJ.nt *:S.t whic~ the clad may disinteg~ate upon ~ubsequent co-~ii~~.:::

.. Each r*eactor ~~d it:,s steam.generators are enclosed\\n.a.steel.,-1i11~/i~in-

, fo-:rced concrete containment structure o,f 45 psig design pre~sure:. *. A;.

-;routine oper.?t:ing pressure of 10. o + o. s psia is maint~iµe~Lwith vac~UIII

,pumps.

The applicaD:t,has,s.tated, th;t e,iq1er. o.~, tt.e: ;wq con'ta:1,.n,m~nt. spray subsystems, employing chilled, slightly alkaline water, together.with two of the four containment recirculation spray subsystems will return the

,containment to subatm.ospheric pressure within 40 _minutes :l'..n the.unlikely event o_f a. lc:>ss::-of.:._coolant ac~idept:...

The appli~an*t has stated that protection wiii; he. :S.fforded against* the m~ximu~ wave' ru~mp expected during hur'ricanes' in.,the' vi.dn':i.,ty of: the :

.. : station.

The applicant has proposed using signals from certain' protection,instru-

, m.ents.for con_troL_purposes.

The Co_rnmiJ:;tee continue~ to l:>elieve that controL and protection instrument.ation:* siw.uld.. be sep.arated to the'. fullest

.exteU:t practicable._ The Co~ittee* believ~s t~ai: the-propos_~*p.- p_;rot~rc'tion sys:t:em, can a:11.d sh_ould be modified. t? el,iminate_ or reduce. to. a mini~um.

the interconnection ~>f controf ~nd protectio.n instrumentat1Pn*.* The

. modifie'd system should be. reviewed by._ the AEc: Regulatory Sta.ff.

~.*

_The Comm;Ltte_e continues to call '~ttention to\\n~i:ters that warrant careful

,:,,.~onsidera,t_ion with ~~ga:r:d-to.re~*ent. reactor~ o'f.kigb, p'o~er' d.eiisi'ty *. {:lnd other matters of si.gnificance for all large' water-'cooled\\ power reactors.

These matters, stated in our report to you of December 20, 1967 on.Diablo Canyon, apply similarly to Surry Station Units land 2.

e Honorable Glenn T. Seaborg 3 -

April 29, 1968 The Committee believes that, if due consideration is given to the foregoing items, the nuclear units proposed for the Surry Station site can be con-structed with reasonable assurance that they can be operated without undue risk to the health and safety of the public.

Sincerely yours, Original signed by Carroll W. Zabel Carroll W. Zabel Chairman

References:

1. Letter from Hunton, Williams, Gay, Powell & Gibson, dated March 20, 1967; Surry Power Station Units 1 and 2 License Application, Part A; Part B, Preliminary Safety Analysis Report, Vols. I, II, and III.

2.

Letter from Virginia Electric and Power Company, dated June 21, 1967; Amendment No. 1 to License Application.

3. Letter from Virginia Electric and Power Company, dated July 5, 1967; Amendment No. 2 to License Application.

4.

Letter from Virginia Electric and Power Company, dated August 24, 1967; Amendment No. 3 to License Application.

5.

Letter from Virginia Electric and Power Company, dated October 6, 1967; Amendment No. 4 to License Application.

6.

Letter from Virginia Electric and Power Company, dated December 7, 1967; Amendment No. 5 to License Application.

7.

Letter from Virginia Electric and Power Company, dated December 8, 1967; Amendment No. 6 to License Application.

8.

Letter from Virginia Electric and Power Company,. dated January 4, 1968; Amendment No. 7 to License Application.

9.

Letter from Virginia Electric and Power Company, dated January 19, 1968; Amendment No. 8 to License Application.

10.

Letter from Virginia Electric and Power Company, dated February 14, 1968; Amendment No. 9 to License Application.

11~

Letter from Virginia Electric and Power Company, dated March 1, 1968; Amendment No. 10 to License Application.

12.

Letter from Virginia Electric and Power Company, dated March 18, 1968; Amendment No. 11 to License Applicaiton.

APPENDIX C Comments on Surrey Power Station Units land 2 Virginia Electric and Power Company Preliminary Safety Analysis Report Amendment No. 3 dated August -24, 1967 Prepared by Air Resources Environmental Laboratory Environmental Science Services Administration April 1, 1968

• Our original comment on this amendment was made in a submission dated September 26, 1967 and was in the fonn of a question on how the vir.tual source distance of 300 meters was applied to the long period av~rage concentration.

• Infonnal discussion with *the applicant~ s meteorological consultant has since indicated that the virtual source correction only amounted to an appropriate increase in the vertical concentration distrib,ution Caz).

Thus at 500 m, a Oz of 27 m was used instead of 19 m for Pasquill Type D -which amounts to* an added dilution factor of 1.4. This seems to be reasonably-conservative in view. of the fact that in applying a correction factor due to a volumetric source, it does not seem appropriate to*account'for the crosswind horizontal dimension of the source when the concentrations are averaged over a crosswind sector.

137?

APPENDIX C Conuuer,ts on Surry Po~ver Station Uni ts 1 and 2 Virgi~ia.,'.Electric*.S:nd.Power C_ompany Prel:i.-minary Safe,ty An.alysi s R~port.*

A>-nendment No: 3'~ *dated August 24, i967 Prepared by Envi~~nn~~nt~l, ~Mete9ro~ogy Branch

. - Air Rese;,:rch Laboratory Environmental Science Services.Administration September 26, 1967 Question 9. 1-It is not cl,e.p.r: f1;ow.(he !=ins,Jer.to, th~ quest~on how the virtual sourc~

di St13.11~~ of,)Oi:*l\\1eters W~E>-,e.pplied to the long,period *average *conceri~

tratiopr.~.~quation~

If:. i~:.'.the**eqUS.tiQn,.both... X *anc~!' crz :wer)e c.orrect:ed,

~ :-:*

for the ~~~.distaI;f~ -* (X: ~ctual + x0 ) an additional.diluti~_I\\ ~acto-r *of. __ -

about 1~. '1'.,'0µld.. re,su.lt:,p-t a di stanGe. of lOOm_ during neutral conditions.

If onJy :f:he ~z:. was,--.corre~:'ted. for the. aµded *vfrtuai' di sta,nce. art. added..

dilution-fa-~tor ~f: about. 3 ~ould. ~esul t ai: lOOrri.

He believe 'the' ia:tter method JP; be, ~~r;, physic;aliy coi;r~6i: and i~or~_* i'n agrJe... ~nt with. I:ecent .

build~ng:_d_i~ut;9,n fie-ld t.es~s* at the-Nati~:ma.l R6-ac.tor Testing S_t'a.ti'on*~-;

~

I * ' :. :, * ~. ;

\\...,..

,. ;i

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'l

• APPENDIX C Comments on Surry_ Power Station Units 1 and 2 Virginia Electric and Power Company Preliminary Safety Analysis Report Volumes 1 9 11 9 and III dated March 20 0 1967 Prepared by Environmental Meteorology Br~ch Institute for Atmospheric Sciences May 16 g 1967 From the data presented, the Surry site seems to offer no meteorological, factors which might adversely affect the diffusion climate.

Surface-based inversions are computed to occur between 25 to 30% of the time which *agrees with Hosl~r 9 s, [l]

.value for that_ area and is about average for the countryo -

i -~.

• }

- The recommended meteorological assumptions for the hazard analysis of an uncontrolled release seem.appropriately conservativeo For the first 12.hours 9

, the *mean wind° direction is assumed to be invariant and a centerline concentra-

-tion is computed using Pasquill "F 11 (inversion) diffusion rates a a 1 m/sec*wind 9

• a ground source; and a virtual source distance correction to account fo-r '*building turbul~nc:e.

In view of the persistence statistics presented in Figo :.2~-2-Sg it

_would seem unrealistic to carry these assumed conditions beyond a 12-hour: period a especiaily*without ave.raging the concentrations over a 22-1/2 de_gree arc_:?;,,The assumptions -for the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and for the remainder of the 30-day petiod _

also are reasonable a Since, excep_t for the annual average, no ground~levei relative' con:centrat*ion values (-x/Q) '~.ere pregented in the report; no,ltlpot'**checks were made on the dose calculations.

The appropriateness of the virtual i;ource -

distance corre'ction could not be checked because a value for the -bupdi~g cross-sectional area (A in eqc 3) could not be found.

[l]

-.Reference

.,t; Hosler 0 C. Ro O 1961~

"Low-level Inversion Frequency in the Contiguous

• United_ States." Monthly Weather Review 5 Vol. ~9 0 pp~ 3i9-;339 o

)

e CEREN*

Mr. Roger S.* Boyd APPENDIX D DEPAR.TMENT OF THE ARMY COASTAL ENGINEERING RESEARCH CENTER 5201 LITTLE rALLS ROAD, N.W.

WASHINGTON, D.C. 20016 Asst. Director, Reactor Projects Division of:Reactor Licensing U. S ~ Atomi_c Energy Commission Washington,, D. C.

20543

Dear Mr. Boyd:

11 March 1968 Ref~rence is made to your letter of 18 May 1967 and various discussions and meetings between Mr. R. A. Jachowski of the CERC staff, and a member ofi the AEC staff regarding Docket Numbers 50-28.0 and 50-281, Virginia Electric and Power Company's Surry Power Station, and Amendments 1-~ thereto.

As indicated above, Mr. R. A. Jachowski has been assigned the review of this app:j-icat;i.on from the viewpoint of storm surge level.

wave height and wave period, wave runup and overtoppi:ng, and water level setdown as associated with the Probable Maximum Hurricane (PMH) and their relation to the station site.

The design still water level of 20.1 feet above MSL, based on the ap-plicant's analysis, would appear to account for the meteorological paramet~rs associated with the Probable Maximum Hurricane (as defined.

by the applicant).

However final decision on the acceptance should await the results of a study presently being conducted by the Hydro-meteorological Branch of the U.S. Weather Bureau to more clearly define the PMH parameters. It is expected that the results of the Weather Bureau study will 'become available about 1 May 1968.

It is therefore our suggestion that the applicant (Virginia Electric and Power Company) should reevaluate the selection of maximum design water level, wave runup, and wave overtopping based on the results of the Weather Bureau Btudy of. PMH para.meters when they become available, and that the final appraisal of this data by CERC await this reeval-uation.

Sincerely yours, r

r,,.,....,(' /

i

\\" I

l 0~~-P-1<. 1, \\ \\,__.(:);/__3,i't.:J-f:...e c

'-JOpEPH M. CALDWELL Acting Director 864

APPENDIX E Su_-r:r:;r ?0... ;.:.;-:_* s.:co/~ion Surry cv*.A:-~ty) \\/'5...:---bi:i...ia P:2.C Doc:::::~ 50-2cJ, 281 The site i::.: located. or. a pc::1.ir.::;*c:.lc. :'or:::ed. cy 1Bo 0 turn of' the James Rive:r bet'\\,Tceri. Je.r:.E:stow~1 an::l ::e-,:_;c::.~.; 2rews :;.:-"d a.bout 25 :miles upstream fror:1 C"::.csal)ca~:e Bay.

T::e rive:.:.* ::.s ~rackis:. in tr.J.s reach and :f'lm1 reverses \\.rite ti::e tide.

Data of* fre3t~~*rater ru:.rioff r" -*.:- (),

u_v....

in the Preliminary The data however do not include the d.roug:i1t of lS*.sc--::,::._ :~*..:.:::::.:-as ~-fr:..icr. flow::. occurred that were conside;raoly lower thar. t:::.c :-:..:.::.ir:n..:.:.:1 r,o:rthly flow of 857 cf's shown.

Thus the :r;,ea..11 :flow for tl1e pe:-::.o-::.. ".:..:.Gust tl".roue;h October 1930 was less than 700 cfs ( cubic feet per se:cc.:-.::), lowest me.:m ~onthly flow was about 620 cfs and rtlnimu.rn flow fo:* 7 co~1secuti ve days was about 440 cf's on the basis of records of' Jarc.es ~..iver at Cn..."-"tersville a."'ld Appomattox River near Petersburg.

Except during periods of high :.~*~:110:t'f; tic.al flow priraarily accounts for the upstrearr, and downstream cu::-r;;;.--/.:; s :p:::.st the site. The oscillatory flow does not directly indicate tr:e: rate of flushing out from the James Estuary of effluents added to it. In cc~eral, flushing out of effluents that remain near the-surface wcul:l be :;.t a r;reater rate than is indicated by the runoff, while effluents the:.::. Gin:~ to a lower depth of the river would be flushed out at a slowe1* rc.teo No quantitative data are presented in the report but it may be esti,;:atcd. tn2..t heated effluent from the plant would remain near the surface ar:d. would. be flushed out at a rate that is considerably greater than the runoff would indicate.

The site is underlain by unconsolidated. beds that dip toward the sea..

Ground-water is utilized from several deeper sandy strata in these beds as well as from the uppermost quaternary depositse The latter are extensively tapped for domestic or other small volu.~e uses, except near the James River where pumping wc*..:.ld. lead to the intrusion of' brackish water. In the lower strata wate::..* is found. at artesian pressure because of' overlying less permeable deposiJ.:;s o

. Water enters these strata principally near the Fall line,-:l:e:ce they are exposed. Wells along the James River between Claren:0::-t ::.~-:.d i_,:orfol~ :~r..vG "been drawing water from this artesian aq_uif'er si:::-1(;.::: *:..:_e ea1*ly :;:,~::;.*-..; o::.~ this century and the artesia.YJ. head has been dro:r:~:.:.. *_6 stewiily" :.~-..:!:.*e this head is lower than the overlying surf'icial grc*.::-.d.-water table loccl recharge from the upper ground water through t:-.-2: coni'ining deposits to tt.e artesian aquifer can occur.

ii O i Rad.ionuclides s:pi],.lecl at the sit-2 or :ie})::isi tcd. on the c:round in the vicinity would be ex!)ected to ent,~::..~ -~:"c surficie.l eround water and migrate slowly towards the nearc::;-:;. st!'e~ cha..;mel.

Entry into the lower aquifer could occur in ares.:; where p~-:;ping has lowered the artesian nead. Migration times ti.11"'0-:..1.gh the less permeable confining beds is a matter of ccn~ccture but could be eA-nccted to be a matter of decades.

The nearest well to the site is a"v :iOg Isla:.1.d. State Water Fowl Refuge about.. one mile from the reactor. other nearby wells are at Jainestown, Fort Eustis, and Ba.cons Castle *.

References:

Cederstrom~ D. J., 19~*5, Geology and. Ground-Water Resources of the Coastal Plain of Southeastern Virginia: Virginia Geol. Survey, Bull. 63 *

.., 1957,.. Geology and Grm.md...::*J'ater Resources of the York-James

__......, _____ Peninsula, Virginia.: u. s. Geel. Survey; Water Supply Paper 1361.

The analysis. of the geology of the Surry Nuclear Power Station site, Atomic E.~ergy Commission Docket ( 50-280, 281) was reviewed and compared

~Tith the. available 1i terature

• Ti1e geolcci C analysi S i S carefully derived I

an~ pr~sents an adequate apprais~l of those aspects of th~ geology which wo.ulc.. bej,ertinent to an engineering evaluation of the site..

Although there are no identifiable. geologic

• structures which could be, expected to localize seismicity in the irranediate vicinity of the site, the area. is underlain at depth ca - 1500 feet by crystalline rocks contimicms. with those which crop out to the ~;est w:i thin the Piedmont Province.-

Hence it.must be assumed that earthq,uakes with bedrock intensities comparable to th9se. characteristic :of' the Piedmont Proy:i.nce may occur.. in the area_ of the plant.

APPENDIX F REPORT ON THE SITE SEIS:MICITY FOR THE SURRY POWER STATION, VIRGINIA At the request of the Division of Reactor Licensing of the Atomic Energy Commission, the Seismology Division of the Coast and Geodetic Survey has evaluated the seismicity of the area around the proposed reactor site in Surry County, Virginia, and has reviewed a similar analysis made by the applicant in the "Preliminary Safety Analysis Report," of the Virginia Electric and Power Company.

The applicant's report of the seismicity which necessarily included an ex-haustive valuation of the local geology, the surface foun-dation conditions, and the possibility of liquefaction is satisfactory for an evaluation of the seismic factor.

In reviewing the seismicity two significant factors were considered, first, the differences in maximum intensi-ties in the Piedmont Province and the Coastal Plain* Province in which the proposed reactor site is located, and, second, the intensities-of distant earthquakes, New Madrid, Missouri, of 1812 and Charleston, South Carolina, of 1886.

None of the Piedmont earthquakes in Virginia of intensity VII* (MlVI) at distances of 90 to 250 miles from the proposed site were felt in the vicinity of the site.

Within 50 miles of the

\\. site in the Coastal Plain Province the reported earthquake

.i.

~ :.

~ * !

intensities do not exceed IV and approximately 25 earthquakes of intensities I-IV have been reported since 1925.

~_In..:'"~:x;a,m~n.ing -the ef-fect_s_ of tb,e Ne\\\\! Madrid and Charies-t-on ear~J:iquakes at the proposed site intensiti~s ra.nging _

from.V to-~ _weak VI,were repor~ed at Will:1am.13burg, N9rfolk and,James.tow_n... A wel_l clocuII1ented,-intensity VI w?,s reported at E*icpmon.ct,: c:!ipprqxima te ly, 50 miles. from: the proposed. Su_rry site.

_, -,The Qoqst and Geodetic ~urve?, after consideri:rl.$_,th~. _

site geol.ogy _arid the _proposed soil conditions, agr,ee _wi.th_,-:-

tl:1.~: -q.pp_l_:t;cant that an acce lera_t?-on_ of O. 07 g at ground sul:':'."

f&Qe,wou);;d,*-1:>e.- adequate for r,epr.esenting earthquake distur-

~

bance f:! lil~ely* to occur within the lifetime of ~he fac:i.li ty.

In add:1._tion, _:we *agree that the applica,nt's proposal to

-- d,e,~ign fp:t'., q.n_:acceleratio:q. of Q.15 g,at ground surface would

,. \\*.*

-.aq.e,quat~.l:Y :Pr~t.. ~C:t ~h~_ f,'ac.~li ty\\ ~~or.gro~Zf~- -~?,t;icm, from the maJ:Cimum _ ~a~tl}.qtJ,ake_. likely ~o affect_.this -~~-it,~.

.. ' ~- '

,.;. *~..

U.r "-S ~ ;c.Coast,,:,and Geodetic Survey Rockvfiie/'MaI'yiand' *-20852 - *:

.. ?:j/::-),>J

.

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APPENDIX G REPORT TO AEC REGULATORY STAFF ADEQUACY OF THE STRUCTURAL CRITERIA FOR THE SURRY POWER S.TATION. UNITS 1 AND 2 VIRGINIA ELECTRIC AND POWER COMPANY (Docket Nos. 50-280 and 50.:..281)

A.

NATHAN M. NEWMARK Consulting Engineering Services 1114 Civil.Engineering Building Urbana, Illinois 61801 by N. M. Newmark,

w. J. Hall, and J. Hendron, Jr.

March 1968

ADEQUACY OF THE STRUCTURAL CRITERIA FOR THE SURRY POWER STATION UNITS 1 AND 2

• • ,.,... :.*:**~*t... '.:.. by N. M. Newmark, W. J. Hall, and A. J. Hendron, Jr.

INTRODUCT!ON

• *1, 1"

~-.. '!.

This report _concerns the* *adequacy *of the containment structures and components, reactor piping and reactor*. _int.ernals:,.. for th.e Surry Power Station Units 1 and 2, for which applicat:!~ri 'for,{'construction permit has been made to the U. S.

Atomic Energy Commission (AEC Docket Nos.,50-280 and 50-281) by*the Virginia Electric and Power Company.

The facility is to be located in Surry County, Virginia on a point of land called Gravel Neck which juts into the James River.

The site is *approximately 30 miles northwest of Norfolk, Virginia, and seven miles south of Williamsburg, Virginia.

Specifically, this report is concerned with the evaluation of the design criteria that determine the abilit*y of. the containment system, piping and reactor internals to withstand a design earthquake acting simultaneously with other applicable loads forming the basis of the design.

The facility also is to be designed to withstand a,ma/x!mt11I1_ eart_qquake simultaneously with other applicable loads to the extent of insuring safe shutdown and containment.

This report is based on information a~cl cri,teria set forth in the Preliminary Safety Analysis Reports (PSAR) and supplements and amendments thereto as listed at the end of this report.

We have participated in discussions with the AEC regulatory staff and the applicant and-. its consultants, in which many of the design criteria were discussed in detail.

DESCRIPTION OF THE FACILITY The*Surry Power Station is described in the PSAR as a pressurized water reactor nuclear ~.team supply syslem furnishec;l by the Westinghouse Electric Corporation 1:1nd designed for an-initial power output of 2441 MWt (816 MWe net) for each unit.

The reactor coolant system for each unit consists of three iocips, each loop having components (steam generator, pumps, and piping) generally s:i,.milar to th<>i:ie for Indian Point Unit No. 2.

The reactor vessel wiil.have _an inside diameter of about 13. 0 feet, a height of 42. 3 feet,.. and is designed for a pressure: of 2485. psig. and a temperature of 650°F.

The.

vessel is made of SA-302 Grade B low alloy steel internallf. clad with*:type 304 austenetic stainiess steel.

'l*

e

• The reactor containment structure which encloses the, -reactor and steam generators for each unit, consists of a steel lined totally reinforced concrete vessel with cylindrical walls, a flat base, and a hemispherical dome.

The cylinder

• will be about 126' -0" inside diameter with a 4' -6" minimum wall thickness.

The spring line of the dome will be about 128 ft. above the inside surface of the foundation mat.

The dome.will have an inside radius of 63'-0" and a thick-ness of 2 1-0".

The liner will be made of 3/8-inch carbon steel sheet conforming to ASTM.

A-432 Grade 60 specification having a guaranteed minimum yield strength of 32,000 psi.

The reinforcing in the cylindrical portion of the shell will consist of horizontal and v'ertical bars' and diagonal bars placed at 45 ° to the hori-zontal in both directions in the._ plane of the wall to resist tangential shear.

Radial shear will be resisted by stirrups or diagonal bars.

The reinforcing will conform to ASTM Al5 or ASTM A408 specifications.

For size 14S and 18S bars, the Cadweld method of splicing will be employed except for a minor number of splices which may have to be made by welding.

Personnel and equipment access hatches are provided for access to the contain-ment vessel.

In addition, there are other penetrations for piping and electrical conduits.

The facility includes* a cooling water intake structure located on the James River.

The intake pumps discharge the cooling water into a paved canal approxi-mately 1-1/2 mi. in length.

The cooling water is approximately 16 ft. deep in the canal, and we believe tpe design to be adequate.

The information on the geology at _the site indicates that the surface deposits are sediments of the Norfolk Estuarine formation of Pleistocene age extending to depths of abou_t. 50 to 80 *feet.

The upper 20 to 35 feet of this formation consists of layers.of brown and mottled brown sand, silty sand, organic and inorganic silts and clays, with interspersed thin lenses of iron oxide cemented sands.

The lower part of _the formation consists of layers of gray sand, silty sand, and organic ~~d inorganic silts and clays; many of which contain decayed vegetation and -shell. ffagments.

The Norfolk formation just described uncomfonnably overlies the Chesapeake group of Miocene age.

Within the site area, the surface of the Miocene sediments are estimat~d to be_ about. 240 feet thick.

This layer consists of compact very stiff tough clays', green to dark gray in color, with occasional compact sand and silt*members.

These soils are noted to be strong and stable with moderate to high shearing strengths.

Underlying the Miocene sediments are Eocene, Paleocene and Cretaceous sediments estimated to be about 45, 55~ and 800 feet in thickness, respectively.

From seismic investigations about two miles southeast of the site, the location of the crystalline bedrock.is estimated to be at a depth of about 1300 feet below ground surface.

e There is no known fault near the plant site.

The nearest known *major fault is

• located southwest of Richmond about 60 miles from the site.

SOURCES OF STRESSES IN CONTAINMENT STRUCTURE AND CLASS I COMPONENTS The containment struct:ure-is to be designed for the following loadings:.dead load _,if the structure including effect of hydrostatic pressure, ice and snow loads; internal pressure corresponding to a loss of coolant accident of 45 psig; an internal temperature of 280°F; wind loadings; tornados (for Class I-*

structures_ *and systems whose failure might prevent shutdown) for a tangential wind velocity of 386 mph, an external vacuum of 1.5 *psig, and missiles; *and seismic loads as desctibed next.

The earthquake loading will be based on a design earthquake of 0.07g maximum horizontal ground accelerationo The*reactor also is to be designed to allow safe shutdown.under a maximum ground acceleration of 0~15g.

Class I piping and equipment will be designed for normal live loads com-bined with pipe rupture loads and earthquake loading.

The reactor internals are to be designed to'resist earthquake combined with blow-down loadings and other applicable loadings.

COMMENTS 0N ADEQUACY OF DESIGN Foundations The containment structures are-*--to be founded on the stiff Miocene clay layer at about elevation -45 to.~50 ft.

The fuel building is to be founded on pipe piles extending into the Miocene clays with the base of the structure on a con-tinuous mat at about elevation 0.

The-auxiliary *building is to be founded on a continuous mat at about elevation +2.

The control area is to be founded on a continuous mat at about elevation +2.-

The basement floor of the turbine building will be at elevation +9 with the structure founded on a system of.

soil~bearing strip footings; the turbine-generator will be founded on pipe piles driven-into the Miocene *clays~* The service building will be founded ori spread footings in densely compacted granular backfill at about elevation+2q.

The detailed foundation studies that have been completed indicate that th~

factor of safety against liquefaction occurring during the maximum credible earthquake (based on 0~15g m4ximum horizontal ground acceleration)"is.con-sidered adequate for the foundations:and drainage conditions proposed at the Surry site. We concur in this evaluation.

However, from the studies that have been made thus far, it appears;-t:hat*there is a possibility of significant relative deformation in both. th_~ horizontal and vertical direction between,various building components of this facility.

The applicant* ad~ised in Amendment 8 that after evaluating pile test results

- to be conducted as a part of the foundation evaluation program, if the actual pile displacements are such that the design safety factors would be reduced (with reference t<? rattle space requirements), the fuel building foundation will be revised to assure that the clearance between buildings.is adequate.

Aiso, the design of piping running between buildings must inc'iude provision for with-standing the possible relative motions.

We believe tha:*t the approach outlined by the** applicant in Amendment 8 is satisfactory.

Seismic *nesign The information presented in the Amendment.s.to the PSAR indicates that the reactor will be designed to withstand within the elastic range, the effects of an earthquake based on a maximum horizontal ground acceleration of 0_.07g. It is also noted that the reactor will be designed to allow safe shutdown under an earthq~ake based on a maximum horizontal ground acceleration of 0.15g.

These values are in agreement with those given-in the.report of the U. S. _Coast and Geodetic Survey (Ref. 5).

In the case of safe shutdown and containment under the maximum credible earth-quake, we concur in the basis of designing for an earthquake based on 0.15g maximum horizontal ground acceleration.

However, we believe that this criterion, coupled with the use of standard spectra, will not lead to design criteria in the velocity controlling'region with the desired degree of con-servatis~.. it,. :i~- our recommendation that the design spectra for the maximum credible earthquake be based on a maximum horizontal ground acceleration of 0.15g and a maximum ground motion velocity of 9 in/sec.

In this manner, the spectrum can reflect a reasonable degree of amplification in the velocity con~

trolling region, a region in which a small numb~r of critical items fall with regard to design.

These ground motion bounds will not lead to difficulties, in our estimation, with liquefaction in the foundation soils.

Our reason for recommending this additional degree of conservatism in the seismic criteria results from our evaluation of the site con.ditions which encompass large depths of sediment overlying basement rqck. It is our belief that significant amplification might occur, and.it is our recommendation that some degree of conservatism*be incorporated in the design to account for such possible amplification.

The*spectra presented in Figs. 59.15.-1 and 59.i5"'-2 reflect th~ above noted criteria and are acceptable to us~

The general method of dynamic analysis for the containment structures.has been outlined in. the PSAR *and supplements and we concur in the approach described.

The damping factors to be employed in the,dynamic analysis are summarized.

in Table 2.5-2 and in answer to Question i2.2.4(3 of Supp~einen,t - Vol. 2.

'4/ For reinforced concrete the damping_factor ~f 5 perce~t will be associated

<~ith stres*s levels at *or* slightly below :yield and -at le~st a moderate degree

• of* cracking~:*

• Damping* values of 10 percent, as noted in. answer to the question

_:* :.,.: cited wdtild be :*ass*ociated with a high degree of cracking of the concrete. -

iHowever, as noted in* the discussion, the damping values are over-alLvalues

*which include the* damping in both. the reinforced concrete structure and the

• <soiL Irideed, in* this situation with the containment vessel founded on* the Miocene clay, one would expect a rather high degree of damping from the foundation *~ystem, and this damping by itself has not been singled out for attention in the PSAR and, supplements and amendments.

We believe that damping including rocking of the containment vessel on soft soil should not exceed 10* percent* arid--the

• applicant concurs as noted in.Amendment 10.

• As indicated

• **** in* Table*

12. 5-2 i: *the *applicant intends to* use l:ewer damping values for. Class. I..

• ,/systems and components.

Generaf.."Design:Pfovisions:fer Containment The loading combinations and allowable stresses are discussed in the PS.AR,

__ Section 5, and in answer to ~uestiori 12.2.3 in Supplement - Vol. 2.

We are

-< "in**agreement with *the load combinations noted *. It is also st.ated that *the.

. miritiuium -'allowable* stress* and* tension under Case* 3, whi'ch includes the* hypo-*

theticai ; earthquak~ ~ 'will be "limited to 80 percent of tl).e minimum tensile *

• tsfreri.gtli of"the'reinforcing steel which does not exceed 90 percent-of.the minimum yield 'sfrength. of the *reinforcing *steel.

This criterion is acceptable

to rui!r:, ::* :-,. >* '. *.. *.

• • .,*,_*-*] -:.*";*:*

(

I.

' The* design' of the liner' is *discussed* in,,the* *PSAR and** in* the supplement.. It*:_ is

\\' note~:':that: the attachment; spacing will be such that the critical buckling.

. *., : stress wilFbe above the yield* point-of the liner material.

Moreover, it ts noted'that'the*limitirigstresses will *be in accordance with Section 3. of-the

.ASME.Boiler and'Pressure V~ssel Code *. We.a:reled to believe from these state-ments that;the stresses and strains will be limited to values below yield.

for~both the design and maximum *earthquake* conditions *. *. *Hence we concur in the-approach ~dopted. *

- I

. "'*The"- analysis and. d~sign of the penettations, including the large penetrations,

is d'escribed--ii;i.-:the-PSAR and supplements.. We coni::ur in.the tnethod of:analysis for analydng the stresses, *as

• described by the applicant;~-- -Attention,is.

called to the necessity for providing continuity of loads and*deformations

from* the stiffening ring surroundirig the opening into the shell_ under the various loading conditions imposed.

Pipi'ng', :. Vessels; Supports*; Reactor Ve~sel Internals 1 *and Other Applicable Compbn'ents. **.

The discussion presented in Section 10 of Supplement - Vol. 2, and on p.

s1oa.:..r:of:Amendmetit,.1-,, notes that piping and vessels will :_be designed in i;..

,*<.'J..,.

e

• accordance with Westinghouse Report WCAP-5890, Revision 1, with modifications.

We are in agreement with the approach outlined.

The criteria presented for the design of the reactor internals appears satisfactory to us.

CONCLUSIONS In line with the design goal of providing serviceable structures and com-ponents 'with a reserve in strength and ductility, and on the basis of the information presented, we beli*eve the design criteria outlined for* the con-tainment and other Class I components, including the reactor internals, piping, vessels, and supports, can provide an adequate margin of safety for seismic resistance.

• REFERENCES

1.

• "Preliminary Safety Analysis Report -- Vols. 1, 2 and* l, II Surry Power S~ation.Units.1 and 2, Virginia Electric and Power Compa~*-196 7.

2.,"Preliminary Safety Analysis Report -- Supplement Vol. ~1.c Ul,d,2,"

Surry

• power Station Units 1 and 2, Virginia Electric and Power-~any, 1967.

3 *. IIPreliminary Safety Analysis Report -- Amendments 1, 2, 3,. 4"';_._ 6, 7.. 8,

• • 9,-10," Surry P.ower Station.Units 1 and 2, Virginia :El~tclc'"'and *Power i

• Company, 1967 and 1968 *

. 4 *.. *,.Additional submissions:

a)

Revised Repert

• Environmental Studies, Proposedc.. Ntjclear Power Plant, Surry, Virginia, by Dames and Moore, dated November 17, 1967.

b)

Report on Seismicity Analysis, Surry Nuclear P~~r.* Plant: Si~e, by Weston Geophysical Research, Inc., dated April 1, 1967.

c)

Report on Test Pile Program, Surry Power Station, NUS-1424, by Stone and Webster, dated June 28, 1967.

5.

"Report on the Site Seismicity for the Surry Power Station Virginia,"

U.S. Coast-and Geodetic Survey, Rockville, Maryland, February 23, 1968.

'11. ~.**q.~~

tJ. )

f(,<l-0{

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APPENDIX H IN REPLY REFE!'f TO:

UNITED STATES DEPARTMENT OF THE INTERiOR.

FISH AND WILDLIFE SERVICE WASHINGTON. D.C. 20240 APR 8 1968

• Mr. Harold Price Director of Regulations*

U. S. Atomic Energy C~ssion

. Washington, D. C.

20545

Dear Mr. Price:

This_ responds to_the questions of your staff as to whether we would object to the issuance of a construction permit for the Surry Power Station, Units 1 and 2, Docket Nos. 50-280 and 50-281, before the answers are provided to questions raised in our letter of February l2 concerning the applicant's proposed radiological program.

We understand that. lack of finality in design and operating criteria for these units makes it essentially impossi~le-to.

present clear answers to these questions at this time. Therefore, we have no objection to the issuance of a construction pe?mit before the questions are answered.

We suggest that the ~pplicant provide the answers to questions raised in our letter of February 12 during consultation with the Service as contemplated under recommendation No. 1 of our letter of September 8, 1967.

Sincerel.1' yours, 1.2~0 S.:

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APPENDIX H UNITED STATES DEPARTMENT *oF THE INTERIOR.

IN IUCPLY IIIEP'IEII T01 Fl~H AND WILDLIFE SERVICE

. WASHINGTON, D. C.

20240 Mr. Harold L. Price Director of Regulations U. S. Atomic Energy Commission Washington, D. C.

20545

Dear Mr. Price:

a* 1 21968 This is in response to Mr. Boyd's letter of August 29, 1967,

• • transmitting a copy of Amendment 3 to the Prel:iminary *safet;t

':Analysis Report of the application for a construction permit

_for the Surry Power Station, Units Nos. l and 2, ~urry Count)'.,

. Virginia, AEC Pocket Nos. 50-280 and 50-281.

The applicant has submitted a general proposal* for a radio_;,

logical program for the aquatic.environment. More specific.

• infoma.tion will be requir.ed so that our Radiobiological

• Laboratory staff 'cari make an adequate evaluation of the

• • • • proposed* radiological.monitoring program. It is requested.

'thatthe applicant be *required to provide t~e following information:

(1) a list of the controlling radionuclides t

selected for study, (2) an estimate of the composition and.. *

. reiative concentrations of the various radionuclides in the' liquid; *effluent,* (3) the locations of the sampling stations.,

*,. (4) the reconcentration factors for the various.'species of**.

fish and shellfish, and (5) a list of proposed indicator species for the'various radionuclides.

-~-

The opportunity to express our comments on the*a:m.endment is

appreciated.

Sincerely yours,.

492 r-

-3.:.

APPENDIX H UNITED STATES DEPA.RTl\\/lEf\\[T OF THE INTERIOR FISH AND WILDLIFE SERV!CE WASHINGTON, D, C.

10240 Hr. Harold L. Price Director of ~egulations U. S. Atomic Energy Commission Washington, D. C.

205L~5 Dear Hr. Price~

'lhis is in reply to Er. Boyd's letter of April 3, 1967, request.ing our comments on the application by Virzinia Electric and Power Compar,y for a construction permit and faci1it,J license for its proposed Surr-f Nu-clear Power Station Units 1.md 2, James River., Surry County, Virginia, Docket Nos. 50-280 and 2tn.

The project would ;Je located 0:::1 c:. point of land called Gravel Neck on the southern bank of the James River approximately 14 miles northwest of Newport News, Virginia.

'I'he proposed plant would consist o:r two pressurized water reactors desi2,ned for an ultimate combined output of appro::dmately 5,092 thermal megawatts, with a: gross electrical output of about 1,695 megawatts.

A raaioactive waste disposal system and other facilities required for a complete and operable nuclear pmrer plant would be provided.

qondenser cooling water would be taken fror,1 the James River on the east side of the site (downstream), conveyed to the plant via a paved intake canal, I

and discharged on the west siue of the plant (upstream.) after absorbing radioactive and heat wastes.

Cooling water would be circulated at a rate of about 1,722 c.f.s. per unit, for a total of 3,444 c.fos.

The applicant plans to conduct radiological stuaies of the environment prior to, and during, reactor operation, although details of the proposed studies are not giv-en in t,he,Safety Analysis Report.

The James ?..iver in the project area supports, extensive and valuable commercial nnd sport fishery resources.

Tne James River estuary, just below this proposed plant, is the major seed oyster. producing area in Virginia.

1-Iog Island Waterfonl 1-Ianagement Area, operated by the Virginia Cormnission of Ga.-rne and In1a.'1a *:i'isheries, is located immediately north of, and adjacent to, the plant site.

Tne applicant indicates that the release of radioactive wastes would not exceed maximum permissible limits prescribed in Title 10, Part 20., of the Code of Federal Regulations.

1\\lthough :!:.hese limits refer to ma.."'Cimum levels of radioactivity that c.s.n occur in drinking water for man, without resulting in any known harmful effects, operation within these limits may not always guarantee that fist and wildlife will be protected from-ad-verse effects. If concentrations in,the receiving water were the only

consideration, maximum permissible limits would be adequate criteria for detennining the safe rate of discharge.

However, radioisotopes of many elements are concentrated and stored by oreanisms that require these ele-ments for their normal metabolic activities.

Some organisms concentrate and store rao.ioisotopes of elements not normally required but which are chemically similar to elements essential for metabolism.

In both cases, the radionucliaes are transferred from one organism to another through various levels of the food chain just as are the nonradioactive elements.

These transfers may result in further concentration of radionucliaes ana a wide dispe-rsion from the project area, particularly by mieratory fish, mammals, a."1.d birds *

• • ~,

In view of the above, we believe that pre-operational and post-operational radiological surveys planned by the applicant should include studies of the effects of raa.ionuclia.es on selected organisms indigenous to the pro-ject area which require* these waste elements or sirnilar elements for their metabolic activities.* These surveys should. be planned in cooperation with the appropriate Federal and St;te agencies. If it is determined from pre-operationai surveys that the* release of radioactive effluents at levels permitted. under the Code of Federal Re6>11lations would result in harmful concentrations of radioactivity in fish and. wila.lif"e, plans should be made to reduce the discharge of radioactivity to acceptable levels. Post-operational surveys should be conducted to evaluate the-predictions based on the pre-operational surveys ana to serve as a basis for reduction of raaioactive levels to insure* that no unforeseen damage occurs.

In view of the importance of the sport and co:mmercial fisheries and. wild-life resources of the James River, it is imperative that every possible effort be made to protect these valuable resources from rauioactive con-

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tamination. -The~fore, it is reco:rrunended that the Virginia Electric and Power Company be required.to:

1.

Cooperate with the Fish and v!ildlife Service, the Federal Water Pollution Control A~~inistration, the Virginia Co:rmnission of Game and Inland Fisheries, the Virginia.Institute of Marine Sciences, the Virginia '\\*Tater Control Board, and. other interested State agencies in developing plans for radiological surveys.

2.

Conduct or arrange for the conduct of pre-operational radio-logical surveys of selected organisms indigenous to the area that concentrate and store radioactive isotopes, and. of the environment including water and sediment samples.

These surveys should be conuucted by scientists knowledgeable in the fish ana *wilalife fiela *

. 3.

Prepare a report of the pre-operational raaiological survey ana provide five copies to the Secretary of the Interior for evalu-ation prior to project operation.

4.

i.:ake moaifications in project structures and operations to re-duce the dischar::;e of -radioactive 1*mstes to acceptable level

.f it is detern1ined in tnc pre-operational or the post-operational surveys that the release of r2aioactive effluent per,nitted under

'.title 10; ?2.rt 20) Coa.e of ~cederal Regulations~ would result in nannful concentrations of radioactivity in fish and wildlife.

5o Conduct -radiologic cl ::m.Tveys, s:udlar to those specified in re-co:rmnendation 2 above, ru1alyze the data, and prepare and submit reports ever:1 three r.10~1.ths during the first year of reactor operation and every s:J_::.;. *.:ionths thereafter or until it has been coaclusively demonstrG.ted that !10 significant adverse conditions exist.

Submit five co)ies of these reports to the Secretar-J of the Interior for distrfoution to the appropriate State and ?ederal ci6encies for eva.luatio~1.

1:-re understand it is the Cora::1ission 1s opinion that its regulatory authority over nuclear power plants involves only those hazards associated with radioactive materials.

However, we recommend and urse that before the per-mit is issued, thermal pollution and any other detrimental effects to fish and wildlife which,nay result from rilant construction and operation be called to the applicant I s attention. '!/e reco1mnend further that the ap-pl:j_cant be requested to discuss this matter with appropria.te State con-servation officials and the ?ish and_Fildlife Service anct to develop mea-sures to minimize these hazards.

G

"'.."e are particularly concerned over the possible effects of increased water temperature on aquatic organisr,1s in the James River.

• Company engineers have made stud.ies relating to this addition of heat to the river by uti-lizing a hydraulic model located at the Corps of Engineers Waterways Ex-periment btation, Vicksburg, i*lississippi.

Operation of the hydraulic model indicated that the dominant water move-

.nent in the river at the plant site results primarily from the oscillating ebb and flood of tid.eso In order to estimate the effects of condenser discharges on river temperatures, under these flow conditions, a model of the two unit nuclear pow~!'. plant i*Jas installed and operated at full capa-city with a. temperature rise of 15° /.*across the condenserso The ambient river temperature during these studies was held at 80° F., which is an averaE,e July and J'..ugust condH,io:::..

1*'ind velocities were assumed to be 2 m.p.h.

Under these con~itions, condenser cooling water discharged into the river would not exceed 95° /.

These heated waters were lighter and, therefore, remained on the surface where* heat dissipation was rapia.

The movement of this heated surface 1vat.er was directed ei t.her up or downstream by the tidal flows.

Evidence of the cooling* rate and extent of trans-portation was demo;.1.strated by the location of the 84° surface isothermal.

This isot.hennal extended offshore to ::nidstrea.m, or about 1. 5 miles out from the outfall, and either up or downstream approximately 6 miles de-penaing on the tide cycle.

The mctXimum surface temperature increased along the opposite river banl{ ranged from 2 to 3° F.

3

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'...,..a Under these conditions it is anticip&ted that the effluent would have no aaverse effects on lare;er forms of aquatic life; however, planktonic or-ganisms may be destroyed when passini; through the condenser coils as a result of the rapid temperature increases.

Ecological surveys should be conducted prior to and following plant operation to measure the effect of plru1t operation on the bioloc;y of the river.

T"nese surveys should be planned in cooperation with the a.ppropria.te Federal and State agencies.

If it is determined from the pre-operational investigations that the heated-water to be discharged into the J*a..'nes River would result in changes in the environment that would be si,:niflcantly detrimental to fish and wild.life, plans should be maae to reduce tl1e temperature of the effluent to acceptable levels. Post-operational surve;:,,s should be conducted to evaluate the pre-dictions* based on the* pre-operational surveys and to ensure that no un-foreseen damage occurs.

Another potential hazard to fishery resources in the river is the cooling water intake.

Unless the int<Jke is auequately screened, fish may be dra,;m in* and destroyed.

Suitable fish protective facilities should be installed to prevent loss of i'is~ through the intake structure.

In view of the Administra.tion' s policy to maintain, protect, and :improve the quality of our environment and most p~rticularly the water and air media, we request that the Commission urge the Virginia Electric and Power Company to:

1.

Cooperate with the :i.i'ish and ~.-~ildlife Service, the Federal Water Pollution Control Adrninistr2.t,ion, the Virginia Commission of Game and Inland Fisheries, the Virginia Institute of Narine Sciences, the Virginia Water Control Board, and

• other interested State

/

agencies in developing plans for ecological surveys, initiate these surveys at least two years before reactor operation, and continue them on a regular basis or until it has been conclusively demonstrated that no significant adverse conditions exist.

2. }ieet with the above mentioned Federal and State agencies at fre-*

quent intervals to discuss new.plans and to evaluate results of

. existing surveys.

Construct, operate, :md maintain such fish protective facilities over the intake structures as needed to prevent significant dam-age to fishery resources.

4.

Make such modifications in project structure and operation in-cluding facilities for cooling discharge waters as may be de-terrained necessary as a result of the pre-operational or post-operational surveys to 9rotect the fish and wildlife resources of the area.

The opportunity for presenting our views on this subject is appreciated.

Sincerely ~s, i~i.v.~L Acting ~ommissioner 4