ML19340B705

Forwards Draft Safety Evaluation Re Getr Landslide Hazard & Seismic Design.Addl Info Needed to Determine Soil Property Effects on Seismic Analysis

ML19340B705
Person / Time
Site:	Vallecitos File:GEH Hitachi icon.png
Issue date:	10/27/1980
From:	Eisenhut D Office of Nuclear Reactor Regulation
To:	Darmitzel R GENERAL ELECTRIC CO.
References
NUDOCS 8011110686
Download: ML19340B705 (81)

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f, UNITED STATES y ;)c(i j NUCLEAR REGULATORY COMMISSION

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October 27, 1980 Docket No. 50-70 Mr. R. W. Damitzel, Manager Irradiation Processing Product Section Vallecitos Nuclear Center General Electric Company Post Office Box 460 Pleasanton, California 94566

Dear Mr. Darmitzel:

We have, in all but one respect, completed our review of the General Electric Test Reactor (GETR) with regard to landslide hazard and seismic design of structures, systems and components important to safety. Our Draft Safety Evaluation is enclosed. The draft status has been assigned because additional work by both the staff and licensee is necessary regarding soil property effects on the saismic analysis (See Evaluation section II C, page 8 and Appendix B). At the time this work is completed, a supplement finalizing this safety evaluation will be issued. Assuming satisfactory resolution of this issue, thB evaluation, in its final form, along with our Safety Evaluation, issued May 23, 1980, regarding the proper geologic and seismic desi n bases for GETR, will set forth the NRC staff's position on issues (1) and 2) of the October 24, 1977 Order to Show Cause.

Contingent upon satisfactory resolution of the outstanding issue discussed above, it is the staff's position that, upon completion of the proposed modifications, GETR can be operated safely considering the geologic and seismic design bases determined procer by the staff. The enclosed Draft Safety Evaluation is being submitted to the Advisory Committee on Reactor Safeguards and the Atomic Safety and Licensing Board assigned to this proceeding.

It should be noted that changes imposing operability and surveillance require-ments, and operating restrictions as discussed in Sections II A, B and C of the enclosed Safety Evaluation, must be incorporated into the GETR Technical Specifications prior to future operation. You should inform us of your schedule for proposing such changes.

In addition, you are requested to submit a description of your pre-operational test program for the seismic scram and fuel flooding systems.

incerely, Darrell G. Eisenhut, Director

-Division or Licensing

Enclosures:

As stated cc w/ enc 1: See next page i

6021220 h

General Electric Company October 27, 1980 CC

• California Department of Health

's ATTN: Chief. Environmental Radiation Dr. Harry Foreman, Member

. Control Unit Atomic Safety and Licensing Board Radiologic Health Section Box 395, Mayo 714 P Street, Room 498 University of Minnesota Sacramento, California 95184 Minneapolis, Minnesota 55455 Honorable Ronald V. Dellums Ms. Barbara Shockley ATTN: Ms. Nancy Snow 1890 Bockman Road General Delivery, Civic Center San Lorenzo, California 94580 Station Oakland, California 94604 Advisory Committee on Reactor Safeguards Friends of the Earth U. S. Nuclear Regulatory "ommission ATTN:

W. Andrew Baldwin, Esquire Washington, D. C.

20555 Legal Director 124 Spear Street San Francisco, California 94105 Jed Somit. Esquire 100 Bush Street Suite 304 j

San Francisco, California 94104 Edward Luton, Esquire, Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Mr. Gustave A. Linenberger, Member Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555 George Ed' gar, Esquire Morgan, Lewis & Bockius 1800 M Street, NW Washington, D. C.

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.E WASHINGTON O. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR THE GENERAL ELECTRIC TEST REACTOR GENERAL ELECTRIC CCMPANY DOCKET NO. 50-70 Introduction On Octocer 24, 1977, the NRC issued an Order to Shcw Cause to the General Electric Company (GE or the licensee) requiring +. hat the General Electric Test Reactor (GETR or the facility) be placed in cold shutdown pending further Order af the Comission. The basis for this action was new geologic evidence which placed in question the adequacy of the GETR's seismic design.

The issues of the Order are as folicws:

(1) What the proper seismic and geologic design bases for the GETR facility shculd be; (2) khether the design of GETR structures, systems and components important to safety can be mcdified so as to remain functional, considering the seismic design bases deter lined in issue (1) above; and (3) hhether activities under Operating License No. TR-1 should be suspended pending evaluation of tne foregoing.

In its Safety Evaluation dated May 23, 1980, the staff concluded that the folicwing seismic design parameters are appropriate for the GETR (Issue 1):

1.

The Regulatory Guide 1.60 spectra anchored to 0.75g as the maximum effective vibratory ground motion at the site. Th ?s is set by motion on the Calaveras faul t.

2.

A surface displacement of one meter of reverse-oblique net slip along a fauit plane which could vary in dip from 10 to 45 degrees and which could occur on a Verona fault zone strand (splay) beneath the GETR during a single earthquake event.

3.

An effective vibratory ground motion of 0.6, anchoring the Regulatory Guide 3

1.60 spectra, togetner with a fault displacement of one meter as described in 2. above.

The May 23, 1980 Safety Evaluaticn did not review

• he landslide hazard. The staff nas new completed its review of the landslide. hazard as well as tne seismic design of GETR structures, systems and components important to safety (Issue 2).

These are the subjects of the follcwing evaluaticn.

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PART I i

LANDSLIDE HAZARD

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LANDSLIDE HAZARD

1.0 Background

On September 27, 1979, the staff issued a report entitled "Geosciences Branch Safety Evaluation Report Input on the GETR." That report documented the Geosciences Branch review up to September 6, 1979 and defined the NRC staff position regarding the proper seismic and geologic design bases for the General Electric Test Reactor (GETR). At that time the staff position regarding the landslide hazard was: "In the absence of a definitive evaluation, we conclude that the GETR could be impacted by a landslide."

On May 23, 1980, the staff issued its Saiety Evaluation (Reference 6) regarding the proper seismic design bases for GETR. The staff position at that time was: "The proposed GETR landslide investigation program, dated April 17, 1980, provides a generally acceptable investigation program and should provide sufficient data regarding the stability of the hillside deposits for the staff to evaluate the potential hazard."

General Electric submitted a landslide stability analyses report (Reference

1) on August 29, 1980; the staff review of the landslide hazard is provided in the following discussion.

2.0 Evaluation 2.1 Geologic Evidence of Slope Movements The southern slope of the Vallecitos hills has been interpreted by both General Electric and the California Division of Mines and Geology (Reference 5) as an old landslide complex. While it is the i

staff's position that the evidence strongly supports tectonic origin j

iof the offsets observed in the trench exposures at the site, a landslide origin has not been ruled out. Geologic data indicates that the offsets displace Holocene (less than 10,000 years old) soils (Reference 6, Page A5). The staff has concluded that the offsets of the youngest soil horizon could have occurred within about the last 2,000 years (Reference 6, page A10). The younger soil horizons are displaced approAimately. 3 feet (Reference 6, page A9).

Future slope displacements have a higher likelihood of occurring along the existing shears or offsets because existing planes of weakness i

in an old landslide deposit form a natural-surface for water to flow along and produce a weathered zone which will have lower shear strength.

The conclusiva that future displacements are likely to occur on existing shears is based on general geologic oburvations and probabilistic considerations presented in our May 23, 1980 Safety Evaluation.

. 2.2 Geotechnical Investigation and "esting General Electric has' completed the landslide investigation as proposed in a letter dated April 17,1980 (Reference 2). This program included field and laboratory investigations, as well as an analysis of the hillside stability with the data collected. The main objectives of the field investigation were: (1) to define, if possible, the location of any subsurface landslide failure surfaces; (2) to obtain representative samples of subsurface materials; and (3) to determine subsurface ground water conditions.

Sampics obtained during this investigation were tested to detemine the appropriate parameters to be used in the stability analyses. Data from the testing program were utilized to perfom slope-stability analyses for both static and dynamic conditions. The results of this investigation and analyses are presented in the General Electric landslide stability analysis report (Reference 1).

The subsurface investigation involved drilling, sampling, and standard penetration testing of four borings within the hillside. The borings included RD-1 thru RD-4, as shown in the encicsed Figure 1 of Reference 1.

Pitcher Barrel and Modified California Drive samplers were used to obtain samples for laboratory testing. Although samples were obtained nearly continuously from within the expected shear zone at RD-1, no shear zones were detected in the samples. No shears were noted in borings RD-2, -3 and -4, which were sampled intermittently. Therefore, no additional information regarding the shape of any subsurface shear surfaces was obtained.

A number of soil samples were obtained for index property determination and laboratory strength testing. Four material types were identified and selected for testing: material types 1 and 2 are classified as clayey siltsilty clay (ML-CL); material type 3 is a sandy clay (SC);

and material type 4 is a clayey gravel and sand (SC). The distribution of each material type is shown in Figure 2 of Reference 1.

Piezometers were installed in all borcholes to help define groundwater conditions. Near the tN of the hillside, ground water was encountered at a depth of about 40 to 80 feet. Midway up the hillside, ground -

water was measured at about 175 feet below the surface. The groundwater profile shown on the aclosed Figure 2 of Reference 1 is judged to be reasonable for use in stability analyses.

. A comprehensive laboratory testing program was performed to determine the strength characteristics of the varicus soils within the hillside.

These tests included consolidated-undrained triaxial tests with pore-pressure measurements and consolidated-quick direct snear tests en both undisturbed and remolded test samples.

In addition, test data frcm previcus investigations were used to augment the results of this program.

Both total-stress and effective-stress strength envelopes were provided, based on interpretation of test results.

Effective-strength enveloces were based on an axial strsin of 10 percent, and total-strength envelopes used remolded test results to model residual strength conditions.

These tests and interpretations follow procedures which are accepted by the staff for large emt'ankments and natural slopes. Mcwever, due to uncertainty in the strength of these natural soil deposits, the staff considered the possibility of even lower shear strength values in its evaluation of slope displacement.

2.3 Sloce Stability Analysis The results of the field invest'gation and laboratory testing program were used as input to the slope-stability analysis. A design earthquake of Richter magnitude 7-l/2, with a peak acceleration of 0.75 g, which coincides with the maximum effective vibratory ground motion determined proper by the staff (Reference 6, page C-2) was used in this analysis.

The methcd used to assess the stability of the hillside deposits as a result of an earthquake is based on a procedure developed by N. M. Newmark (Reference 3) and further reconrnendations provided by F. I. Makdisi and H. 3. Seed (Reference a). This procedure provides an estimate of the amount of permanent slope displacement that may occur during a seismic event.

This procedure is based on a comparison of the minimum seismic forces required to just overceme the available shear strength within the slope, to the actual force induced in the sicpe by the earthquake acceleration. The minimum seismic force is represented by a yield acceleration, K, which is defined as that average acceleration of a y

mass of soil which procedes a slope failure factor of safety of unity.

The earthquake induced acceleration, Krax, is defined as the average acceleration of a potential sliding mass.

Thus, when the earthquake induced acceleraticn (Knax) exceeds the yield acceleration (X ), movements y

will occur along, and in the direction of, the failure surface.

The critical failure surface identified in the GE analysis intersect the ground sarface near the toe of the slope, in the vicinity of the shear zone identified by the GE field explorations.

In this analysis the maximum design ground acceleration of 0.75 is taken at the crest of the hill north of the GETR. The resulting induced acceleration, hax, varies between 0.34 and 0.15. GE conservatively selected Kmax=0.34 We agree with this selection.

In its evaluation of the yield acceleration, K,

y GE shcws a range between 0.18 and 0.23.

In additional calculations made by the staff using lower values of soil shear strength, the staff finds

. that the 4 values could go as low as 0.1.

These results are shown in the enclosed Figure 3 of Reference 1.

The staff agrees that the licensee's value for Kmax=0.34 is reasonable and conservative. However, using a value of K =0.1, the staff finds that the maximum displacement ranges uo to 1 m. y 3.0 Surnary and Conclusions investigations and reports provided by General Electric have been reviewea by the staff and found to satisfy the requirements of 10 CFR Part 100, Appendix A,Section V, Seismic and Geologic Design Bases ((d) Determination of Other Design Conditions; (2) Slope Stability).

In addition these investigations and reports are in agreement with Standard Review Plan Section 2.5.5, Stability of Slopes.

Based on a review of the General Electric site reports, independent studies, and evaluations by the staff, the staff concludes that an earthquake-induced slope displacement of 1 m is conservative. Ground surface displacements resulting from these slope movements would be expected to occur near the toe of the slope, in the vicinity of the observed shear zone, and at some distance (approximately 300 feet) from the GETR plant. Therefore, ground surface displacenents due to the postulated landslide must be considered in the desig: of safety related eyaipment located near the toe of the slope (e.g., fuel flooding system piping) but need not be considered in the design of the GETR reactor structure.

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REFERENCES l.

Earth Sciences Associates, "GETR Landslide Stability Analysis," prepared for General Electric Company Vallecitos Nuclear Center, Pleasanton, California 94566, 1980.

2. ' Letter from R. Darmitzel, General Electric Company, to D. Eisenhut, NRC,

Subject:

Docket No. 50-70, April 17,1980.

3.

N. M. Newmark, " Effects of Earthquakes on Dams and Embankments,"

Geotechnique, Volume 5, No. 2, Geotechnique, June 1965.

4.

F. I. Makdisi~, and H. B. Seed, " Simplified Procedures for Estimating Dam and Embankment Earthquake-Induced Deformations," Volume 104, No. GT7, Journal of the Geotechnical Engineering Division, ASCE, July 1978 5.

California Division of Mines and Geology, " Geologic Evaluation of the General Electric Test Reactor Site, Vallecitos, Alameda County, California," Special publication 56, 1979.

6.

NRC Safety Evaluation dated May 23, 1980.

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f PART II SEISMIC DESIGN GETR SAFETY RELATED STRUCTURES, SYSTEMS AND COMPONENTS

a i PART II SEISMIC DESIGN, GETR SAFETY RELATED STRUCTURES, SYSTEM 3 AND COPP0NENTS 4 Introduction The staff has evaluated the GETR with respect to issue (2) of the Order. Issue (2) is: Whether the design of GETR structures, systems and components important to safety can be modified so as to remain functional, considering the seismic design bases determined in issue (1). The evaluation is presented in several sections as follows: A. GETR Structures, Systems and Components Important to Safety B. GETR Electrical, Instrumentation and Control Systens C. Seismic Design of GETR Structures, Systems and Components Important to Safety D. Offsite Radiological Impact of Design Seismic Events 4 4 i i we g

4 A. GETR STRUCTURES, SYSTEMS AND COMPONENTS IMPORTANT TO SAFETY -1.0 Introduction The GETR is a high-flux, pressurized water reactor which operates at a maximum power of 50 MW thermal. Pressurizer pressure is maintained by uitrogen gas. The reactor core is contained in a 2 foot diameter cylindrical pressure vessel positioned on the bottom of a 9 foot aiameter pool. The pool is flooded with demineralized water to a level 11 feet above the tap of the reactor vessel or 23 feet above the core. Demineralized water is pumped through the reactor vessel and out to heat exchangers for cooling. Coolant enters the pressure vessel near the top of the reactor vessel via two 12 inch diameter inlet pipes, flows down-ward through the core and out near the bottom via two 12 inch diameter outlet pipes. The reactor coolant operates at a me>imum temperature of 180 degrees F and maximum pressure of 150 psig. Tht coolant is subcooled at atmospheric pressure. An isanetri.. trawing of the primary coolant loop is shown in Figure A-1. 2.0 Accident Analysis The method adopted by the licensee of providing protection during and following the design basic seismic event is comprised of three basic phases: 1. Reactor scram at the onset of the seismic event. 2. Initial removal of decay heat by evaporation of existing water l from the reactor pool and fuel storage canal. 3. Long term cooling by-providing makeup flow to the reactor vessel and fuel storage containers. In detcrmining the structures, systems and components necessary to perform these functions,the licensee has considered the possible failures resulting from the seismic event as well as the most limiting irradiated fuel locations. This is necessary to define makeup cooling water requirements in terms of flow rates and time of initiation. The licensee has identified rupture of the primary coolant piping as the most limiting accident cciacident with a design basis seismic event. With such a postulated break, water will drain from the vessel and pool through the primary return lines until the water reaches the level of the return line outlet from the vessel. The fuel is covered by a minimum of 5.5 feet of water at this point. Further drainage due to a siphon effect through the outlet is prevented by the anti-siphon valves (PRI 190 and 191).

A-2 The assumptions made for esaluating this postulated accident include: 1) the worst postulated earthquake occurs with reactor trip initiated by the seismic scram system;

2) simultaneous non-mechanistic rupture of the primary system piping; and

3) heat transfer and decay heat rates based on 25 day power run of the reactor operating at 50 MW.

Results of the analysis of the primary pipe rupture show that the water level drops to the top of the core at 45 hours after the event assuming no makeup flow. At that time, the boil-off from decay-heat requires makeup water at a rate of.8 gpm. Boil-off due to decay heat of the fuel in the fuel storage tanks requires makeup at a rate of 0.74 gpm at about 32 hours after the accident. The licensee has also considered the case in which a freshly discharged core has been placed in the fuel storage canal following reactor shut-downs. The assumptions made for evaluating this fuel storage situation include: 1) the seismic event occurs a tx hours after shutdown from a 25 day run at 50 MW; 2) the temperature of the canal water is assumed to be 130 F;

3) heat transfer calculations for the stored fuel are based on decay heating equivalent to an infinite irradiation of a single core at 50 MW with a 6-hour decay prior to the seismic event; and 4) the primary pipe rupture discussed above is assumed to occur due to the seismic event.

The results of the analysis show that following approximately 34 hours with no makeup, water must be added to the fuel storage canal at a rate of 1.64 gpm to account for boil-off due to decay heat. This makeup flow rate requirement decreases with time. There is no makeup requirement for the reactor pool in this case. Because of the reduced power density of the GETR fuel following a reactor scram, heat transfer due to pool boiling is sufficient to maintain the cladding temperature low enough to prevent fuel damage. Thus, to prevent fuel damage it is sufficient to shut down the reactor and keep the fyel l in the reactor and storage canals imersed in water. Based on our review

) A-3 of the analyses provided, we conclude that the most limiting accident during the seismic event, for determination of cooling requirements to the reactor, is the double-ended break of tne primary pump discharge. The initial system response during a seismic event will be the same regardless of what size pipe break or other transient occurs. The initial seismic scram and short term cooling by evaporation of pool water is the same for all credible accidents and transients as long as the seismic scram is triggered. Since the largest amount of pool water is lost most rapidly due to the double-ended break at the primary pump discharge, all smaller breaks and other transients are enveloped. We further conclude that the cooling water makeup requirements for stored fuel are set by the case which considered a freshly discharged core. It is necessary, based on the assumptions of the analysis, that full core discharges be limited to no earlier than,5 hours af ter shutdown. To keep the core covered for the long term it is necessary that makeup water be provided to replace water evaporated by decay heat. The flow rates of makeup water to both the reactor vessel and the fuel storage tanks, calculated by GE, are adequate to replace l the amount evaporated by decay heat. This was verified by conservative, approximate calculational methods. 3.0 ' Procosed Safety Related Structures, Sys+ ems and Components The licensee has identified the systems recessary to shut down GETR, maintain the reactor in a safe shutdown condition and to cool stored fuel assuming the accident and fuel storage location discussed above. These systems include new systems, existing systems and existing systems with modifications and are listed on the following pages. The reason that the system, component, or structure is safety related is stated. Those systems, components, and structures which require modification from the original design include a description of the modification. This list was provided in the licensee's report "U dated' Response to NRC Order to Show Cause Dated 10/24/77" (Reference 16. l

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SEE C liefd EL* Hit WlflCaslML Hone reqialred 1. I:aactor Concrete 5truc ra Used to support other systems. components and structures important to safety Contains pool anil canal water 2. kesclcr primary piping and - piping (within the pool) contains - MJ standpipes to top of essergency Assuclated Restraints the water coolant necessary to besp coollag check velves, reactor fuel cool. Restraints prevent high forces - AJJ new rest:alats to primarv from being applied to the reactor piping. pressure vessel and safety-related primary piping 3. Itcactor psussure vessel - Rpy contains tl.e water coolant Strengthen one of three lateral an.l Associated Restraints necessary to keep reactor fuel supports which hold the reactor cool pressure vessel 4 position. - Restralats prevent high forces fro.e heing applied to the SPV 4. Primary lleat fachanger Restraints Restraints prevent damage to reactor - Md new restraints to the primary primary piping and associated piping heat exchanger to provide additional restraints assurance that the heat exchanger cannot damage the primary systess piping.

• f or altigation of maalemme postulated selsmic events

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1. 't El - %Iutritalitelh111m. Aratattratratmt - cnatis.aca lim 8::na 11 !!flaum mh 5.

Poul float !achan9er R<stralats - Pra: vent ds.xg: to reacto. ;rimary - AJJ pew restraints to tiie pool q ' riping and associated piping restraints heat enchanger to provide addl. tlunal assurance that the heat p" = 3 enhanger cannot damage the primary system piping. 6. Reactor Solsmic Scram and Irly - Tri0gers replJ shutdoinn and depress-Noute signals from aulating selselc System (includes circuitry to urtration of the teactor. Iriggers scram system to new control units actuate control rods and fuel flooding Systene santssion valves. which open el.e emergency coolleg safety-related valves) valves and ll.a fuel flooding System asalssion valves, and close the primary system pressuriser isolation valves. Install new. seismic

triggers, 7.

Control No a and Associated - Assure replJ shutdoimi of the reactor None requircJ N I'"S Containt seals to psevent leatage of prirury coolant 8. Reactor Psessuro Vessel and Pool - Piping contains the water coolant AJJrestraintstotheselines(In psain lines and Pulson injectfun necessary to keep reactor fuel cool sut,pileroom)topreventpostulated line and Associatej Restraints Jamage upstseem of block valves. s Piping contains pool water (line - AJJ a redundant clieck valve to the Integrity prevents rapid pool water liquid poison lujection line loss) litem13) - Restraints psevent Jamage to the piping 4

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A-4 We have reviewed the licensee's identification of safety related structures, systems and components as well as the proposed modifications i to assure ourselves that the licensee has identified all the-safety related structures, systers and components necessary to shut dcwn the facility and maintain the reactor in a safe shutdown condi~.f on during and following the design basis. seismic event. Our analysis of the GETR as modified indicates that 'the principal safety related structures, systems and components are those presented below. a. Seismic Trip of the Reactor and AL.couatic Operation of the Emergency Cooling Valves (List item #6) To assure emergency cooling either by natural circulation of pool water or from the proposed Fuel Flooding System, the primary system must be shut down and depressurized. A seismic trip which would scram the reactor, open the emergency cooling valves and isolate the ~ pressurizer at a low seisnic activity level of approximately 0.01 g effective peak ground acceleration was proposed. The depressurization would be accomplished within one second of seismic i scram actuaticn prior to any significant seismic load being reached. In the event of a loss of power the emergency cooling valves fail open and the pressurizer isolation valves fail shut. GE has stated that the seismic trip system response has been assessed to ensure l that protective actions would have been completed before any damage to the reactor facility could occur. The staff's evaluation of this system is in Section B. b.s Reactor Concrete Structure, Reactor. Pressure Vessel and Canal Fuel Storage Tanks (List items fl, 3, 9) These structures serve as the containers for fuel cooling water. Integrity of these structures must be maintained to assure that ccolant leakage will not exceed that assumed in the analyses (60 gph from reactor pool; 400 gph from storage canal) and, in the case of the reactor concrete structure, that support for other safety related equipment is retained. Water contained within these structures at the time of the seismic event serves as the initial heat sir 3 for fuel decay heat. [ The canal is separated from the pool by a 3 ' piece removable gate to i allow underwater pool and canal transfers. All irradiated fuel, not in the care, is stored in racks designed to maintain a subcritical configuration. The racks are inserted in stainless steel tanks. To replace.the water removed by boiling, the proposed Fuel Flooding System will supply adequate water flow.to the fuel stored in the canal in the event of a seismic event, without operator action. Modifications to the fuel storage tanks include redundant supply line and nozzles for each tank. The. nozzles are installed to act as siphon tubes to maintain all tanks at the same level. The reactor pressure vessel supports the core and other internals which must maintain their integrity. l

A-5 c. Control Rods and Reactor Vessel and Storage Canal Penetrations (List item #7, 8, 13) Control rods must function properly to shut down the reactor and maintain the reactor in a shutdown condition. All systems penetrating the reactor vessel or storage canal whose failure would result in an unanalyzed coolant leak path, must maintain their integrity. Thes e systems include the pool and vessel drain lines, poison injection lines, capsule coolant system, canal emergency recirculation system, control rod drives and isolation valves associated with these systems. Restraints will be added and valves seismically qualified to assure the necessary integrity. d. Emergency Ccoling System (List item #2, 3) A pneumatically closed, spring opened, solenoid-tripped, emergency cooling valve is provided on each of the two primary inlet cooling lines (PRI 130 and PRI 150 in Figure 1A). A check valve is provided on each of two primary outlet cooling lines (PRI 140 and PRI 160). On receipt of the seismic trip signal or a loss of power to these 4 valves the emergency cooling valves open the primary system to the reactor pool. System depressurization is assured by closing the primary system pressurizer isolation valves and pressurizer supply valve. Depressurization does not cause flashing and blowdown of the primary system because the coolant is subcooled at atmospheric pressure. If a rupture occurs in the primary piping water will drain from the pool and reactor vessel until the level drops to the level of the alti-siphon valves (PRI 190 and 191). Standpipes will be added to the top of the check valves (PRI 140 and 160) to insure that the water level in the reactor vessel remains above the core regardless of the water level in the pool. The standpipes se* ',e as the injection points for makeup from the fuel flooding system. e. Fuel Flooding System (List item #12) The fuel flooding system is initiated automatically by the seismic trip described above to provide water to the core and to the fuel storage tanks without operator action. The system will consist of two identical redundant legs each capable of delivering the required flow rate. The required flow rate of 2.44 gpm is the maximum evaporation rate from the irradiated fuel subsequent to postulated canal and pool drainage. Sufficient water is provided for seven days of operation at this flow rate. The reservoirs will be situated on a hill adjacent to the containment building at an elevation to provide adequate gravity feed flow. Each supply leg will approach and penetrate the containment building from a different angle and will be routed to the fuel storage baskets and to one of the stand pipes to be installed on the emergency cooling system. The flow control valves are air operated and fail open on loss of air. The solenoid air centrol valve will vent air pressure from tne flow control valve operator on loss of power, making the system fail safe.

1 A-6 The structures, systems and components discussed above are adequate to snut down GETR and maintain the reactor in a safe shutdown condition provided the equipment identified works as described. The licensee has proposed additional mcdifications to insure that failure of other equipment during the seismic event will not effect the capability to safely shut down. These modifications include adding restraints to the primary coolant and pool heat ext.. angers (List 4tems 14 and 5), the ~ missile impact system (item #10), primary piping (item #2) and the permanent pool shielding (item 114). In addition, a canal impact pad (item fil) will be added for protection of the fuel storage tanks. Based on our review of the GETR we agree that by insuring the integrity of the equipment discussed above the safety related equipment will not be jeopardized by other equipment failures. In the course of the staff's evaluation of the currently installed and proposed safety systems which must not fail during a seismic event, we analy:ed the required capability of the various systems, components and structures with respect to active and passive functions and with respect to tne time frame of that capability. A summary of our analysis is presented belcw. a. passive components that must not fail structurally due to a seismic event and for which there is no backup system. 1. Reactor pressure vessel and internals. (no leakage) 2. Reactor pool and liner. (leakage to 60 gph) 3. Reactor vessel bottom head penetrations, centrol red drives, pool and reactor vessel drain lines and drain line valves. (no leakage) 4. Fuel storage tanks. (noleakage) 5. Fuel storage canal. (leakage to 400 gph) b. Ccmponents that must not fail structurally due to potential impact on essential equipment. 1. Third ficar missile impact system. 2. permanent pool shielding restraints. 3. Canal ficar missile impact system. 4. Primary coolant system restraints.

l l A-7 1 c. Passive components that must not fail due to a seismic event for which there is redundancy. l. Fuel flooding system tanks. 2. Fuel flooding system piping. 3. Emergency cooling standpipes. d. Ccmponents that must operate actively during the initial low magnitude tremor and must not fail passively during the main shcck. 1. Emergency cooling valves PRI 130 and PRI 150. 2. Fuel flooding system ficw control valves. 3. Pressuri:er isolatien valve PRI 110. 4. Pressurf:er nitrogen supply valve GNI 112. 5. Centrol Rods. 6. Seismic Reactor Scram Cricuit. (No operability or integrity requirements after reactor scram and valve operation) e. Components that must remain operable following the main shock. 1. Primary cooling check valves PRI 140 and PRI 160. 2. Anti-siphon valves PRI 190 and PRI 191. 3. Fuel flooding system check valves. 4. Fuel flooding system anti-siphon devices. 5. Liquid poison check valve. 6. Capsule coolant anti-siphon valves. 7. Canal emergency recirculation system anti-siphon valves. 4.0 conclusion ~ The evaluation of the currently installed and proposed safety systems identifies the equipment wnich must not fail during a seismic event. If tnis equipment satisfies the seismic design criteria for the GETR site and the operability criteria described above, the reactor core and irradiated material in the storage canal will remain submerged in coolant and adequately cooled during and following the design basis seismic events, i

4 i A-8 1 i i Technical Specifications imposing operatility and surveillance requirements consistent with the staff's evaluation must be propmed by the licensee t 'and approved by the NRC staff prior to GETR power operation. In addition, .a requirement must be added to the Technical Specifications limiting core 4 discharge to occur no earlier than 6 hours after reactor shutdown. J l l i I i i 1 1 i t 1 i I I I i . ~. -. .,.,m_,

- l ] FIGURE A-1 ISOMETRIC OF PRIMARY PIPING r Pw. iso oiscwAne.a in poet <g >.rg=:f?/ Q'**?.... s ggy omcuAnca io poob y Pc mo sg g v-4 "'h 3 \\ PgGLSL f.- ) / i,. 1 ac t.d 4-t will(04 ( t ii1 f Q % t* b i x i i i i pa: t f) ~p s:- r x C'A n' n ~ we. to t i4* e.maa! b mE l w. ~> dPCIAC ) l a-QD k,,. -o s . ---- -- w, r D '75. -. -a ED= A$wu., rY w- ~a s ? LEGEND COMPONENT NO. DESCRIPTION PRI 130 & 1EO EMERGENCY COOLING V. sES SD BURST DIAPHRAGMS PRI 190 & 191 ANTI-SIPHON VALVES PRI 140 & 160 EMERGENCY COOLING CHECK VALVES P 101 PRIMARY PUMP H E 101 PRIMARY HEAT EXCHANGERS 3 ~

B. GETR ELECTRICAL, ZNSTRUMENTATICN AND CCNTROL SYSTEMS 1.0 Introduction As noted in Section A of this evaluation certain automatic operations must occur for safe GETR operation during and following a seismic event. These operaticns consist of actuation of the seismic scram system and subsequent reactor scram; opening of emergency cooling and fuel flooding system vaives; and shutting of pressurizer isolation valves. The staff has performec a detailed review of circuit diagrams, equicment operating nistories and written system descrioticns in evaluating the systems, including proposed modifications, used to perform these coerations. Our evaluation follows. 2.0 S/ stem Descriotions 2.1 Seismic Scram System Figure 1 is a block diagram of the modified seismic scram actuation circuit. The scram circuitry is activated by two kinemetrics triaxial seismic triggers. The three component triggers (two horizontal and one vertical) will replace the existing Odo component (two horizontal) triggers. Tne sensitivity of these seismic triggers is such tnat they will initiate trip signals at ground acceleratiens as low as 0.01 g and the range extends to ground acceleraticns up to 0.5 g. Each kinemetrics triaxial seismic trigger system consists of tuo modules; a seismic switch and a seismic switch power supply. The seismic switch contains three orthogonal electremagnetic transducers. Tne transducers are small moving coils that produce a voltage proportional to acceleration. Tnis coil voltage is amplified to energize the output relay. The acceleration trip level is adjusted by changing the sensitivity of the amplifier. The seismic switch power supply contains a dc cower supply and a battery. The dc power supply powers the seismic transducers and serves as a charger for the battery. The battery is connected to float on the pcwer supply output, thus it is a backup power supply. Tnerefore, the de pcwer supply and the battery provide redundant power to the triaxial seismic trigger system. The licensee reports that the seismic triggers and cower sucply units (DC 3attery and charger) nave been demonstrated qualified to.5 g. Since action by these comconents is designed to occur at.01 g the staff finds the level of qualification acceptable (See Section C). The system output is a relay contact change of position when the seismic switen is accelerated either vertically or horizontally to a level greater than the preset level. The relay requires a manual reset to return it to normal pasition. A test switch provides a voltage across the transducer coil that simulates an acceleration that would trip the system. This relay will be connected into the GETR scram system in the same way as the present two ccmponent seismic switch. The relay outputs frcm each of the two seismic switch sysiams each send signals to two process scram logics and to redundant fuel flooding system (FFS) control units. Tne process scram logic consists of process relay contacts connected in a one out of two configuration.

B-2 i The associated relays are energized during normal operation. A seismic trip will de-energize relays which send signals to the three reactor protection system (RPS) logic elements, and to the relay scram logic. The relay scram logic system directly interrupts the control rod magnet circuit and sends a scram signal to the protection system logic elements. The logic elements trip their associated power switches when any one of the three input signals calls for a scram. When any two of three power switches indicate a scram, the magnet circuit is interrupted. Thus, control rod latch magnet current is interrupted either directly via the relay scram logic system or by action of the logic elements and power switches. Power to the entire seismic scram system is supplied from the plant 120 VAC vital bus which is powered by offsite, utility supply power or one of two onsite diesel generators. The facility license presently i requires one diesel generator to be operating paralleled with the offsite power to supply the vital bus when the reactor is not sFv? down. Sufficient protective devices, indicators, controls and alarms are provided to ensure that the diesel generator and the vital bus are adequately protected and monitored in this mode of operation. The control rod drive circuitry is separate in function from the safety system or scram functions. A scram causes each control rod to disconnect from itc drive mechanism and drop to the fully inserted position. In order to re-engage a control rod to its drive mechanism, the drive mechanism must be driven to the fully inserted position and the scram circuitry). I must be manually reset (only possible if no scram signals are present In the particular case of_a seismic scram, the seismic trip system must be manually reset to clear the scram signal. The licensee has investigated the possibility and effect of any spurious electrical signal that could initiate a control rod withdrawal during a seismic event. The investigation considered the situations where the control rods and drives are either fully inserted, or with-drawn from the fully inserted position when the seismic event occurs. The analysis of the drive circuitry indicates that no credible sequence or consideration of failures during a seismic event could cause spurious electrical signals to initiate a control rod withdrawal. Further, the separation in function between the control rod drive circuitry and the control rod scram circuitry insures that during a seismic event the control rods will rapidly and fully insert and will not be inadvertently withdrawn, e

B-3 - 2.2 Fuel Flooding System (FFS) Control Units The proposed FFS control system consists of two separate, redundant control units located in the control room. Either unit will be capable of actuating the Fuel Flooding System. The FFS control system consists of the controls for emergency cooling valves (ECV 130 and ECV 150), the pressurizer isolation valves (Pressurizer Control Valve PRI-110, and Nitrogen Supply Control Valve GNI-ll2), and the control of the FFS flow control valves (see attached block diagram - Figure 2). Each control unit is designed to initiate FFS actuation upon receipt of a trip signal from the siesmic trip system (contacts of the seismic follaner relays). When tripped, the control ur.ts send signals to open the Fuel Flooding System Flow Control ialves and the Emergency Cooling Valves, and close the Pressurizer Control Valve and Nitrogen Supply Control Valve. The control units are designed to lock-in the trip condition until they are manually reset. The licensee has . submitted information showino that the control units are seismically qualified by type tests. The control units will receive unregulated, single-phase,120V ac power frcm the plant energency load center (which also supplies power to the seismic trip system as described in Section 2.1 above). The control units are designed to fail safe upon loss of power. 2.3 Emergency Cooling Valves, pressurizer Isolation Valves and Fuel Flooding System Valves The Emergency Cooling System Va',ves (ECV 130 and ECV 150) are solenoid trip, spring loaded, air-operated valves. These valves fail open upon loss of power and loss of control air. These valves have operated reliably over the past 18 years. On five occasions, cut of approximately 1300 valve operations, one of the valves failed to operate; hcwever, adequate cooling was provided by the other valve (the valves are oversi:ed and provide fully redundant capability). Periodic inspection and testing of these valves will be performed in accordance with existing plant preventive maintenance procedures and reactor startup procedures. Once each reactor operating cycle, the operators for these valves are inspected and the valves are functionally tested for proper operation. l l I

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1 B-4 i Each FFS flow control valve will be supplied operating air pressure through a three-port, solenoid operated valve. When the solenoid-operated valves are energized, air pressure will be supplied to the flow control valve operators to snut the valves against spring pressure, 3 j If power is removed from a solenoid-operated valve, it will re-position to vent air pressure from the flow control valve operator causing the flow control valve to cpen by spring pressure. These valves fail open j upon loss of control air and/or electric power. The FFS will consist of two redundant sections each capable of delivering the design flow rate that will assure that the irradiated fuel will remain submerged under water. Each section consisting of a reservoir, piping, and valve will be sized so that the design flow rate is achieved for seven days of uninterrupted makeup. The operation, maintensnce, and testing of the flow control valves as prwosed are similar to those for the eftrgency cooling valves. i The Pressurizer Control Valve (PRI-110) is a remete-operated plug valve with an electro-pneumatic controller that isolates the primary system form the pressurizer after a scram signal. The Nitrogen Supply Control Valve (GNI-ll2) is a diaphragm control valve that isolates the pressurizer tank from the high pressure nitrogen supply bottles. These valves are solenoid trip, air operated valves that close upon loss of power and/or loss of control air. The Emergency Cooling Valves, Pressurizer Isolation Valves and Fuel Flooding System Valves have been seismically qualified by proof tests and analyses. 1 i 3.0 Seismic Scram System Reliability Subsystems contained in the seismic scram and emergency cooling trip system and fuel flooding system are: Power Switch Seismic Relay 86/SC (Relay Scram) Seismic Switch Logic Element FFS Control Unit Process Scram Logic 3.1 Redundancy and Fail Safe on Loss of Power The character of the redundancy and fail safe nature of these subsystems are discussed below. A s i

3-5 3.1.1 Power Switch There are three separate, identical redundant power switches whose output contacts are connected in a two-out-of-three twice configuration. The control rods will scram on loss of power to any two of the three power switches because the relay car. tacts within the power switch open on loss of power. This places the reactar in a safe condition in the event of a seismic occurrence. 3.1.2 86/SC (Scram Relay) The scram relay (86/SC) output contacts are connected in a one-out-of-two configuration. The control rods will scram on loss of pcwer to the scram relay because the contacts cpen on loss of pcwer. The 36/SC relay is in the control room under the surveillance of tne eactor operator. 3.1.3 Logic Element There are three separate, identical redundant logic elements. Loss of pcwer to any two of the three logic elements will cause the scram of the control rods because the input power is coupled through the logic element (i.e., if there is no input voltage, there is no output voltage to ';he power switch). This then places the reactor in a safe condition. The logic elements are in.the control room under the surveillance of the reactor operator. 3.1.4 Process Scram Logic There are two separate, ident1 cal redundant process scram logic biccks and associated relays. Loss of power to tre process scran logic will cause the control rods to scran because the relay contacts open on loss. of power. This places the reactor in a safe condition. Process scram relays are in the control rocm under the surveillance of the reactor operator. 3.1.5 Seismic Relay (Latching) There are two separate, identical redundant seismic relay (latching) subsys tems. On loss of power to eitner one of the seismic relays the relay contacts will open causing the control rods to scram, the depressurization of the reactor, and the initiation of the Fuel Flooding System. These actions place the reactor in a safe condition. i w"' w

B-6 3.1.6 Seismic Triggers There are two separate, identical redundant seismic triggers, described previously. 3.1.7 FFS Control Unit 2 There are two separate, identica' redundant control units. Loss of electrical power to only the control unit would cause reactor depressuriza-tion and scram and the initiaticn of fuel flooding. The two redundant control units are new. They each contain only cne relay. It has been seismically qualified and has a mean time between failure of 1 x 100 hours, a reliability adequate for the GETR safety system. 3.2 Operating Histories A brief examination of the historical record has detennined the number of operations of the GETR scram system and certain of the sensors which cause scram. A single cycle (a cycle is of five weeks average duration) review determined the split between the number of process scrams and those produced by the nuclear instrumentation. For example, the 86/SC relays operate only for process sensor scrams,and the power switches and logic units operate on all scrams. A more comprehensive review of twenty-one cycles for the approximate two-year period before October 1977 revealed that the process scram system or representative portions of it were made to function 1,791 times. Extrapolating these data to the full 18 years of GETR operation, much of the process scram system was operated as many as 14,000 times. The two 86/SC scram relay contacts are estimated to have been operated more than 11,500 times over the past 18 years. 't should be noted that the logic elements and power switches were installed and operated for 11 years before October 1977 and are estimated to have operated 8,500 times based on the above infonnation. The process scram logic and relays are estimated to have been operated more tnan 11,500 times over the past 18 years. In addition. as discussed in the licensee's letter of February 14, 1978, the seismic scram system has shut dom GETR eight times since GETR began operation in 1958. Five of these scrams were due to seismic events (Magnitude 3.5 to 4.9), two due to improper seismic switch gap adjustment and ane due to seismic relay tube failure. It is to be noted with all these operations there has not been one unsafe failure; and that the scram system has operated properly for all l the occasions (test and actual) in which a reactor scram was requirs, 3.3 Surveillance of Seismic Scram and Related Ecuicment During a seismic event, the safe shutdown of the GETR requires automatic operation of the following reactor systems: I 1. Seismic Scram and Emergency Cooling Trip System 2. Reactor Control Rods 3. Fuel clooding System

4 B-7 Y Survet1~ &.se testing asscciated with these systems is described belcw. 3.3.1 Seismic Scrmi and Emergency Cooling Trip System TFe . actor seismic scram and energency cooling trip system consists tne sensors and circuitry. The seismic sensors and scram circuit ..are functionally tested prior to every reactor cycle startup (average 5 weeks). The test consists of manually tripping the sensor and verifying loss of control red magnet power, opening of the Fuel Flooding System admission valves and the emergency cooling autcmatic valves, and closura of the requirea pressurizer valves. This test is performed with alternate halves of tiis redundant circuit in bypass each time. Redundant circuit components, therefore, are checked every other cycle. 11e seismic sensors are calibrated annually. The emergency cooling automatic valves provide rapid depressuri:ation of the primary system. The primary pressurizer safety related valves include the valve wnich isolates the pressuri:er frem the primary system and the valve which isolates the nitrogen supply frem the pressuri:er. In-service surveillance of these valves consists of a fune"onal test every reactor shutdcwn. 3.3.2 Control Rods The control.'ods assure rapid shutdcwn of the reactor. In-service surveillance of the control rods consists of the folicwing tests and checks : Red scram check:. and latch integrity tests are performed prior to e'._ y reactor cold starrf; drcp times for each centrol rod are measured and recorded at least semiannually; preventive maintenance is performed on eacn centrol rod drive unit at least every 15 months; and centrol red bank reactivity worth is routinely checked at least annually. 3.3.3 Fuel Floeding System The Fuel Flceding System assures a long-term supply of cooling water to irradiated fuel at the GETR. The surveillance program ter the Fuel Floeding System (FFS) components consists of the folicwing tests and inspections: Continuous and periodic in-service surveillance. Continuous in-service surveillance of the FFS will consist of the folicwing: Mcnitoring admission valve position, monitoring reservoir water level, monitoring critical point temperatures, and monitoring pipe heating tapes. In addition to the cyclic (every 6 weeks) operability checks discussed in section 3.3.1 above, periodic in-service surveillance of tne FFS will consist of the follcwing: Instrument and Control Test and calibration (annually); water flow (annually); si;nen breaker test (annually); reservoir water samole (at least annually); stand pice pneumatic test; and systems and stand pice visual insoections. 1

o 3-8 3.4 Alar ns and Monitors The seismic trip system status is monitored by its telltale lignts and by a panel alarm providing an audible and visible alarm. Each seismic cutout relay extinguishes its telltale light wnen it actuates. The panel alarm is activated when either or both seismic relays (latcning) de-energi:e. Limit switches on cr.e emergency cooling valves and FFS flow control valves provide valve position indication. Level indications and alar ns for the FFS water reservoir will also be provided. Tne heat tracing of above-ground piping in the FFS is monitored by indicating lights shcwing availability of heat trace power and icw temcerature visual alarms activated oy thermal switenes at various locati cns. Tne nomally open manual valve on tne proposed fuel ficoding system will be locked in positicn and administratively controlled. We find tnat the indicatien and alam provided to monitor the proposed modifications to the seismic trip system, and the prepcsed fuel flooding system to be adequate for the systems concerned. 4.0 Resocnse Acticn Time for the Scram System It is the licensee's position that the autcmatic actions asscciated with the seismic scram system (i.e., seismic scram actuation, valve operation and centrol rod insertion) will be ccmpleted prior to significant eartnquake loadings. This allows the licensee to not qualify portions of the seismic scram system to the seismic design basis level or, for other safety related equipment, to qualify in the post-operation condition (e.g., core with control rods inserted). As a result, the response times after seismic switch trip for scram action events such as control rod disengagement, reactor shutdown, and valve opening have been examined. The acceptability of these respcnse times, with respect to seismic leading, will be addressed in Section C. The GETR scram system operates when (among other events) the seismic switcnes close. The reactor control rods' are disencaged from the drive mechanism 180 milliseconds arter eitnet of these two seismic switches make electrical contact. That is, all the electrical and electronic scram circuitry have operated and the control red magnetic laten circuit has been intermpted and the control rod begun its drop by the end of the 180 millisecond peried. The control red then drops by the forces of gravity and primary coolant flow so as to be fully inserted from a 36-inch withdrawn pcsitien within 500 milliseconds from the time the centrol red is disengaged frcm tne drive. Based on available rod drop data, it is conservatively estimat.:d that within 300 milliseconds l from ne time the control rod is disengaged from the 36-inch withdrawal starting position, or a80 milliseconds from seismic switen trip, the centrol t l l \\ \\

B-9 rods will be at or below the 12.2-inch withcrawn position whereupon the reactor is considered to be shut down. The energency cooling power-operated valves, pressurizer valves and fuel flocding system admissicn valves are the only valves for which initiating action is by seismic trip or scram circuitry. The emergency cooling power-ocerated valves and the fuel flooding system admissicn valves begin to open and the pressurizer. valves to close within 190 milliseconds after triggering of the scran system. The remainder of the valve operation is complete within a total of one second from scram seismic t?ip. 5.0 Conclus ions The licensee has descrioed in detail the slectrical, instrumentation and control equipment, as well as proposed modifications, necessary #cr automatic cperations at the initiation of a seismic event. We have reviewed this equipment as well as the reliebility of the scram and valve actuaticn circuitry in the context of redundancy, power loss, operating experience, and functional testing and in-service survefilance of scram systems and ccmponents. Furthermore, we have reviewed the response times for the scram action events for safe shutdcwn of the reactor. Based on our evaluation, it' is concluded that: 1. The electric, instrumentaticn and control,equipnent, mcdified as proposed, will perform the necessary automatic actions of reactor scram, pressurizer isolation, emergency cooling valve operation and FFS initiation; 2. The reliability of the scram and valve actuation circuitry provides reasonable assurance that the necessary automatic actions will be performed when reqyj red; and 3. The response times for the scram action events and the safe shutnawn of the reactor are reasonable for use in evaluating the status of equipment during significant seismic loadings. 4 Tecnnical Specificaticns imcosing operability and surveillance requirements consistent with the staff's evaluation must be proposed by the licensee and approved by the staff price to GETR power ope it n J 1 rn v +, -,

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l l C. SEISMIC CESIGN CF GETR STRUCTURES, SYSTEMS AND 2CMPCNENTS IMPORTANT TO SAFET( 1 1.0 Introductien Tne staff has perforned an evaluatien of the seismic qualification of the GETR structures, systems and ccmponents, identified in Sections A and B, necessary to assure that GETR can be shut down and maintained in a safe shutdcwn condition if subjected to a design seismic event. Ors. W. J. Hall and N. M. Newmark have assisted, as consultants to the staff, tnroughout the review and evaluation process. Tne reviews and reccmmendations of our consultants support the staff's conclusions and are attached as Appendix A. 2.0 Loads, Load Combinations and Acceotance Cri ceria Our review of this existing facility is based upon the following general criteria. In the case of nuclear facilities, safety for seismic excitation does imply that certain elements and ccmconents of the system must continue to remain functional. Resistance to seismic motion does not mean ccmolete absence of permanent deformaticn in all cases. Structures, piping, and equipment may deform into the inelastic range, and some elements and components may even be permitted to suffer damage, provided that the entire system can continue to achieve and maintain a safe shutdcwn condition. Hence, the safety criteria for an existing facility may differ for the various elements and ccmconents of the system. Significant differences exist between a test reactor, such as GETR, and a nuclear pcwer plant. Therefore, different requirements result with respect to operating cnaracteristics and safety features necessary for the protection of public health and safety. Despite the differences, the staff has perfarned its review of the GETR in the same depth as one perf0rted for an operating power reactor seismic design review. Tne review concept was not envisioned as one based upon demcnstrating ccmpliance with specific criteria in the Standard Review Plan or Regulatory Guides, since individual criteria do not control broad safety issues. However, current licensing criteria were utilized with respect to the level of design they dictate, and as baselines frcm which to measure relative safety margins to support broader integrated assessment. The principal objective was to provide reasonable assurance that the public health and safety is protected. A-

C-2 In this case, the evaluation involved an assessment of the capability of essential structures, systems, and components to withstand the specified seismic hazards. The GETR is unlike a commercial power reactor in that, due to its small size, tnere is no need to rely on the operaticn of a large number of active electrical and mechanical systems to maintain the reactor in a safe shutdewn condition after the occurrence of the design seismic event. Given the seismic design parameters, only the folicwing structural and mechanical requirements must be satisfied: 1. The structural integrity of the massive cor. crete structure wnich succorts other systens and components important to safety, such as the spent fuel canal and the reactor pool, must be maintained. 2. The structural integrity of the reactor vessel and the c?nal fuel storage tanks must be assured. 3. A source of water, including the associated piping system, must be available after the seismic event to provide water to the spent fuel canal storage tanks and the reactor pressure vessel to replenisn that lost tnrough boil off and evaporation in the process of cooling the fuel, assuming that the fuel coolir.g system piping and associated heat exchangers have failed. It should be noted that once the reactor is shut dcwn, or scranced, residual heat can be removed provided only that the fuel remains covered with water and makeup water is available. 1 )

C-3 Tne GETR, as mcdified, has been reevaluated by the staff to detemine i wnether the staff has acequate assurance that the GETR can safely with-stand the effects of the maximum credible seismic events. The acceptance criteria used in this reevaluation are consistent with 10 CFR 50 Appendix A Criterion 2, and specify that the integrity of essential structures, systems and componenL required to safely shut down the GETR and maintain it in a safe shutdcwn condition (including remcval of residual heat) during and after the design seismic events be assured. To accomplish this goal, the load combinations for the specified essential structures, equipment, piping systems, and components have been deternined to be as follows: 2 1. Concrete Core Structure and Miscellaneous Systems and Compons.?ts (a) Nornal Service i)D+L 1 5 (b) Max 1 mum Credible Accident i)D+L*E S'L ii) D + L - W 1 0 4 2. Reactor Pressure Vessel Shell and Internals (a) Nceral Service i) D + T + P < S (b) Maximum Credible Accident

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C-4 The load symools above are defined by GE as folicws: 0 = sustained vertical loads wnich include the dead loads of the structures and pemanent loads such as water in the piping systems and equipment weight. L = live loads wM:h consists entirely of snow load apolied only to the roof areas of the equipment building and Control Rocm. Tnis loading was that defined by the Uniform Building Ccce for the region which includes the GETR si te. P = Nornal operating pressure Pt = test pressure Pa = maximum accident pressure T = Nor:tal operating temperature Tt = test temperature Ta = Maximum accident temperature W = Criterion wind load which is considered to ce a 100 year return pericd fastest mile wind velocity of 80 miles per hour. E = Criterien earthquake loads which consist of the most adverse loadings resulting frem the surface rupture and the maximum vibratory ground motion postulated at the site, including the most adverse and physically consistant ccmbination of these events. Be safety related structures, piping systems, and ccmponents itemized in section A-3.0 have been reviewed to assure that they will not fail as a result of tne maximum credible seismic event and that non-essential structures, piping systams or ccmconent failures will not result in their failure. Supports for thee piping system and the other safety related components have been analyzed in accordance with the requirements of the American Society of Mr.cnanical Engineers (ASME) Soiler and Pressure Vessel Ccde, Section III, Subsection hF. The piping systems have been evaluated against the loading combinatiens and acceptance criteria based upon the ASME Boiler and Pressure Vessel Code, Section III, Subsection NC for Class 2 piping. The definitions of the acceptance criteria previously listed for piping systems are:

C-5 piping basic material allowable stress Sh = pipe allowable temperature stress range = 1.5Sh Sa = "he allowable stress limits for structures, piping systens, and ccmponents are determined on the basis of material properties at temperatures l corresponding to the specific load combinations. When appropriate, the procedures in the following concrete and structural codes have been utilized to evaluate the structures and components: l 1. ACI 318-1971, " Building Code Requirements for Reinforced Concrete," American Concrete Institute,1971. a 2. AISC, "Specificat!cns for Design Fabricatien, and Erection of Structural Steel for Buildings," American Institute of Steel Cons tructi on,1969. In genersi, the following definitions of the acceptance criteria apply for structures and components: S = working stress threshold - the load which causes a principal structural member to reaca the allowable working stress as defined by the applicable code establishes this limi:. S tructural memoers include beams, columns, shear walls, wall bracing, etc. ultimate capacity threshold - this limit is based upon either U = buckling, initiation of concrete cracking, brittle fracture, or a ductile mechanism of failure. ine calculated threshold is based on the failure of one or more critical structural members or sectionc. As stated previously, the GETR facility is unlike a commercial power reactor facility in that there are no ccmplex systems which must remain functional in the event of the cccurrence of the maximum credible seismic event. Following a seismic scram signal it is only necessary that the previously cited structures, systems, and components required to maintain reactor pool and refueling canal integrity and to provide makeup water, perform 1 their functicns. The containment shall need not maintain its integrity provided that the above conditions are satisfied, thereby assuring the spent fuel and the reactor core remain immersed in water. However, it has been adequately demonstrated that any damage sustained by the containment will have no effect on the leak tight integrity of the reactor pool and the fuel storage canal, or the capability to supply makeup water to these water pools. The concrete walls of these cools have been evaluated 4

C-6 utilizing the in-situ compressive strength of the concrete. Concrete allowable shear strengths are based upon values of shear stresses at first crack initiation in the massive, concrete core structures, as detemined from the most current experiments performed by the Portland Cement Association (Reference 5) and earlier work done at Stanford University (References 6-9) regarding deep beam shear capacities. Meeting these allowable strengths will provide adequate assurance that structural integrity of the concrete core region will be maintained and that cracking in the concrete will not cause strains in the liners which would be detrimental to their leak tight integrity. Also, it has be assured that foundation instabilities due to the mest critical combinations of earthquake loadings will not cause significant water loss from the pools or compromise the integrity of the core structure to support other systems and components important to safety. 4.0 Review Analyses Detailed state-of-the-art review analyses have been performed by General Electric Company and their consultants to verify the adequacy of the GETR essential structures, systems, and components (listed in Section 2.0) to withstand the effects of the design basis seismic events. We have reviewed the information provided in References 4 and 10-44. The following information has been provided to assure that the acceptance criteria requirements are satisfied. The concrete core structure supports other systems and components imptrtant to safety. Analyses have been carried out using linear elastic analysis techniques. The integrity of the core structure is assured in tems of its allowable shear strength and its stability. The analyses performed to verify the capability of the core structure to withstand the effects of the seismic design criteria have included the following: 1. Time histories which envelope the Regulatory Guide 1.60 defined-response spectra anchored at the appropriate acceleration levels as defined. 2. Three dimensional linear elastic time history dynmic analyses utilizing a spring-mass model, with soil spring constants based on Reference 2 and 3, which incorpcrate the effects of the approximately 20 feet embedmeitt of the building. 3. Consideration of the influence of torsion on response. 4. Parametric studies of the effects of shear modulus, foundation contact areas, and variations of modal damping ratio, due to soil-structure interaction, on the response of the linear elastic model. The material damping coefficients used for the concrete and steel were 7 and 3 percent, respectively, which conform to the positions in Regulatory Guide 1.61.

C-7 5. Nonlinear time history dynamic analyses to investigate the effects of nonlinearities associated with potential uplift and sliding of the reactor building slabs and the behavior of the building on the structural response of the GETR. 6. Stress analyses of the concrete core portions of the core structure, arising from the seismic loadings imposed on the structure, using a three dimensional finite element program. 7. A combined surface rupture offset and vibratory motion analysis, to account for the effects on the structure due to the design criteria hazard. i Analyses of the reactor building for the effects due to the design seismic event on the Calaveras fault were performed using a three dimensional spring-mass model. Loads determined from these analyses were then used as input to a three dimensional finite-element stress analysis to verify that the core structure is adequate to withstand motions induced by the design criteria. Analyses of the reactor building for the effects of the design parameters related to the Verona fault were performed by canbining the effects resulting from the vibratory motion with those resulting from surface rupture. The effects resulting from the vibratory motion were determined in a manner similar to that described above for the Calaveras event. That is, a spring-mass model was used to determine dynamic response, which was then used as input to a finite element analysis to determine stresses and deformations. The effects of the surface rupture were determined using a finite element model of that portion of the reactor building which supports and protects the safety related equipment and components necessary for safe shutdown. The fault orientation used in the analyses was that which produced the most critical loading case for the concrete core structure. The fault was considered to pass through the structure at several locations, and the corresponding effects were addressed. The fault location which produced the highest stresses in the core structure was that where the fault plane intersects the vertical plane containing the center of gravity of the reactor building, causing the building to act as a cantilever. Other fault locations which may cause more excessive deformations of the outer, non-essential reactor building walls were evaluated, but were found to cause less stress in the concrete core structure. The concrete core stresses were computed using a linear elastic three dimensional finite element program which included the considerat Sn of potential cracking and yielding of the floor slabs.

~, C-8 Analyses were perfonned to determine representative and con 5ervative input 1 i parameters to be used which would be consistent with the seismic design criteria defined by the Verona fault ha:ard. Soil pressure analyses were 4 evaluated to deternine the physical load limits on the combined load case comorised of a grcund acceleration vibratory motion and a surface rupture offset, the latter represented analytically as an unsupported cantilevered length of the reactor building. The results of.these analyses indicate that there are physical limits on the combined loading of vibratory motion anc unsupported length of the reactor building. That. is, Figure C-1 provides a representation of the limiting load combinations resulting from the specified design basis events. Our evaluation supporting tnis conclusicn is attached as Appendix B. l l Analyses were perfonned to assure that the facility can withstand the load combinations defined above. The capacity of the faci?;;y was determined j based en evaluations for various sets of load combinaticns selected to l conservatively represent the input parameters defined in Figure C-1. These included evaluations for the following combined input parameter cases: a) Ground acceleration = 0.75 g unsuppcrted length = 0 ft. 1 b) Grcund acceleration = 0.0 g Unsupported length = 20 ft. c) Ground acceleration = 0.30 g Unsupported length = 17 ft. These analyses reasonably bcund the limiting load c::mbinatiens representing the ha:ard defined by cur. seismic criteria. In addi tien to the investigations performed to verify the adequacy of the GETR with respect to our specific design criteria, a post-offset analysis i using an input acceleration of 0.8 g was perfonned to demonstrate that the facility could resist a major ground motion which might occur subsequent to a surface offset event. It was conservatively assumed that only the safety related portion of the core structure could be relied on to resist the input acceleration. It was assumed that the remainder of the structure, including all concrete slabs and walls exterior to the core structure, had lost their structural resisting capacity due to the surface rupture offset effects; however, the. total mass for these assumed failed portions were included in the model. Nonlinearities due to potential uplift at the foundation slab-soil interface, as well as at the interface of the interior cencrete structure and the foundation slab were considered. 'I w,~-- a

C-9 The intagrity of the reactor vessel and the canal fuel storage tanks was evaluated by assuring the integrity of the supporting cencrete core structure as discussed above, and by assuring t... 'abili ty of all essential ccmponents and equipment to meet the seismn triteria. Evaluations of the reactor vessel lower head penetrations (References 18 and 29) indicste that maximum stresses do not increase significantly during the design seiWc events and remain less than 10". of allcwable. Tnerefore, failure due to seismic effects is not expected. In addition, it was assured tnat the fdlure of any non-safety related components or equipment would not compromise the integrity of essential items. General Electric has evaluated the reactor vessel and internals, including ne fuel and experiment capsules, for the loads resulting frcm the design seismic criteria. The fuel assemblies used in the core are flat-plate, uranium-aluminum alloy assemblies, consisting of 19 fuel plates each 0.050-inch thick (ncminal), 2.80-in, wide and 37.25-in. long. Tne fuel plates are roll-swaged into 6061-T6 aluminum side pieces, which act as protactive skin containing the fuel. Tne allowable stress for this aluminum skin has been appropriately determined to be 2000 PSI. Tnis allowable stress dces not take credit for the increased yield strengtn of tne aluminum due to irradiation. The results of the seismic analyses (Reference 44) indicate displacement 3 at the core region to be minimal, and stresses on the aluminum fuel cov(ring, about 70 PSI, to be significantly below allowable. The experiment capsules used by GE are constructed of ASTM 304 stainless steel. Based on loads resulting frem the seismic analysis of the reactor building, GE has calculated the stresses en these experiment capsules and has shown the stresses to be belcw the appropriate ASME III allowables. Although altr.inum experiment capsules are not generally used at the facility, GE has stated that tneir use has not been ruled out. Based on our evaluation, we find that tne stresses on the aluminum capsules resulting f-a seismic loadings and operating conditicns have not been snown to bt celow code allowables. Therefore, we find that use of the aluminum capsules snould be prchibited, unless additional detailed analyses indicate that the aluminum capsules are capaDie of withstanding the combined seismic and operating loads, or release of these capsule's inventories is acceptable. Response spectra which envelope the results of the various seismic analyses of the GETR reactor building were und to determine the adequacy of the safety-related systems and ccmponents specified previously. In addition, in the pool region, the experiment table, the pcol shield, the beam port, the thennal shield and the blast mat, the fuel experiment capsules and asscciatec items, and the wall pads supporting miscellaneous piping and equipment have been analyted. In the storage canal region, the canal gate frame and the canal gate rack have been analyted. For all items except the ~. -. - -. -

C-10 4 miscellaneous piping and equipment, the natural frequency was determined, the corresponding spectral accelerations obtain'ed, and an equivalent tatic force analysis was performed. For the small items of piping and equipment, equivalent static loads were determined based on a multiple (1.5) of the peak spectral acceleration of the floor response spectra to conservatively determine the adequacy of their design. Due to the rigidity of the inner core structure, relative support deflections were negligible in these analyses. l As a result of evaluations performed certain restraints have been modified or added, as described in Reference 16, to meet the seismic design criteria for the items listed below: 1. reactor pressure vessel 11. reactor primary piping iii. reactor pressure vessel and pool drain lines and poison injection line iv. permanent pool shielding ) v. primary heat exchanger vi. pool heat exchanger In addition, standpipes were added above the emergency cooling check valves to ensure that water remains over the fuel in the reactor vessel in the unlikely event of loss of water from the pool. For the canal storage tanks, new tanks were constructed and will be used, and structures were added and equipment modified to prevent potential missiles from being generated or causing damage. These included: 1. polar crane impact structure

11. restraints on the polar crane trolley; missile shield, and refueling bridge iii. canal impact pad to prevent damage due to postulated cask tipping.

In addition, to assure the integrity _of the reactor pressure vessel and canal fuel storage tanks, to keep all fuel covered with water, a source of make-up water to replenish that lost through boil off and evaporation is required. To achieve this goal, General Electric has proposed to install a Fuel Flooding System with redundant gravity flow (no power required) supply capability. The Fuel Flooding System is described in References 16, 21 and 24. The system consists of two redundant legs each capable of delivering the design flow rate. Each reservoir site consists of two 50,000-gallon polyurethane flexible " pillow" or " bladder" tanks situated on a hill adjacent to the containment building at an elevation which provides adequate gravity fed

C-ll ficw. Each supply leg is constructed frem 1-1/2" I.D., reinforced snythetic rubber. The line is " snaked" in a shallow trench providing line slack and permitting the line to accomodate postulated surface faulting. Through the yard area, the line is buried in a 4" stainless steel pipe wnich protects the line in the event of postulated surface faulting due to either a seismic event or seismic initiated landslide. Each supply leg approaches and penetrates the containment building from a different angle, and is routed to the irrsdiated fuel storage tanks in the canal and to the reactor pressure vessel. Each supply line inside the containment building is allowed to move within a orotective cover. Tnis arrangement protects the line and orevents unacceptably high seismic stresses. The lines inside the containment building are a combination of: (a) high pressure, high vacur rated reinforced rubber, (b) stainless steel flexible hose, and (c) rigid stainless steel pipe. Reactor pressure vessel water addition (frcm the FFS) is to the reactor vessel standoipes previously discussed, and therefore, to the bottom of the pressure vessel. An in-service surveillance and in:pection program has been developed (Reference 16) for tne Fuel Flooding System frcm the source tanks to the points of connection at the reactor pressure vessel and the spent fuel storage tanks, including the interface with the containment structure. The design and analyses of the Fuel Flooding System together with the implementation of the in-service surveillance and inspection program, provide reasonable assurance that required makeup coolant ficw to the reactor vessel and the fuel storage system is available following the design basis seismic events. To provide additicnal assurance as to the operability of the system, aopropriate pre-operational testing of the systen (including actuation of the seismic scram switches), will be required to be performed. Automatic actions, consisting of activation of the seismic triggers and subsequent reactor scram and emergency valve operations, must take place at tne onset of a seismic event. Section 3 discusses the ccmoonents wnich are relied upon to perform these actions as well as the time to complete these actions. Tne licensee argues that the actions will preceed significant seismic loadings which mignt adversely affect tne active cocoonents. The licensee supplements this argument by stating that the seismic triggers, including de batteries and chargers, are seismically qualified to levels of.5 g; the valve actuation circuit (FFS control units) and the subject valves and valve coerators have been qualified to the design basis seismic events; and, that the system has operated satisfactorily during experienced seismic events. Our seismologists have reviewed tne licensees position with the following results: I N

C-12 Review of. Representative Time Histories for Seismic Scram Analysis a". GETR In order to determine the adequacy of the seismic scram system witn regard to the trigger level (0.01 g) and time required to complete the scram action (1 second) the licensee has submitted i a study of near field time histories. The main object of this study was to determine whether consequential horizontal or vertical accelerations would be reached before completion of the scram action. i The earthquake threat at the GETR site comes from two main sources, strike slip events (up to magnitude 7.5) on the Calaveras fault-2 bn away and thrust events (up to magnitude 6.5) in the immediate vicinity of the plant. Thirty-six sets of records from well recorded everts up to surface wave magnitude = 6.9 for strike i slip and surface wave magnitude = 7.0 for thrust faulting were ( analyzed. Several sets of accelerograms were recorded at distances less than 1 kilometer from the fault. The data set can be considered a representative sample of all available data in the magnitude and distance range of interest. Envelopes of all j horizontal and all vertical accelerations during the first second j after recording 0.01 g (the seismic trigeer level) were cocputed and plotted. The highest peaks were as .tated with the pacoima dam 4 record from the 1971 San Fernando eat ke. These were 0.13 g 4 f for the horizontal component recorded sec after reading 0.01 g ]' and 0.24 g for the vertical component recorded 0.52 seconds after reaching 0.01 g. We are not aware of, nor has GE provided, any physical arguments which would preclude the occurrence of such peaks earlier in the time history of earthquake ground motion. It is the staff's position that in determining the adequacy of the seismic scram system that high frequency (310 Hz) peaks of this amplitude (approximately 0.25 g) could occur anytime during the first second after 0.01 g on either or all componc-nts of motion. This conclusion is based upon our relatively limited knowledge of earthquake source models and their relation to near field motion and the limited data base available for statistical studies. Based on our reliability assessment of the scram system, tests performed on the control rods ar.d. internal components (Reference 43), and evaluations performed, we find that reasonable assurance is provided that the circuits required to perform automatic actions will function satisfactorily, considering the minor loadings postulated during the first second of the design seismic events. Some inodifications and strengthening to the control cabinets will be completed prior to.the restart to provide additional assurance that the seismic event will not adversely effect the scram circuits at the onset of the seismic event.

a. 3 C-13 5.0 Summary of Staffs Evaluation The seisnic review analyst; and design of the GETR essential structures, systems and components are in confermance with accepted codes and criteria. In the case of structures and structural components, based on the information reviewed by the staff and our consultants, we find that the analyses performed are of the state-of-the-art that would be used fer existing nuclear facilities. Analyses have been carried out using linear elast', analysis techniques, and where appropriate, nonlinear effects have F considered. We find that General Electric has considered all applicable loadings and effects of imposed deformaticns resulting fecm the design basis faulting and/or shaking in a manner consistent with current practice. It was demonstrated that allowable strengths were ad quate to acccmmodate the effects of the seismic design criteria, with consideratien of standard linear analysis techniques of calculaticn. The results of the analyses showed tnat the concrete shear strengths remained below threshold levels considered appropriate for first cracking. Limited local distress in the reinforced concrete core structure could result frem streng shaking and/or faulting, hcwever, we find that adequate assurance has been provided that its integrity will be maintained to carry out its intended function. In additien, stability analyses performed shew that sliding and overturning of the reacter building is not of concern, due to its dimensicns and enbednent. We hr a reviewed the methods and results of extensive analyses performed to verify the adequacy of the GETR equipment. Necessary restraints have been added or mcdified. Results of analyses and qualification testing of valves similar to these in service demenstrate their abili:y to functicn during and after the design basis events. Operability, as specified herein, means the ability of an active comocnent to perform its required mechanicai motion. It has been demonstrati s tsat either the required action of safety related components will be ccmpleted in the early part of the earthquake after initiation by the seismic scram $1gnal and prior to the initiation of significant vitratory moticn, or that anal, ses and/or tests have been perfonned to adecuately dssure the operation of those safety related components required to operate during and/or after the design seismic events. Functionability, as specified herein, means the ability to remain in a required configuration. Extensive testing and/or analyses have been provided to assure that safety related piping and components will remain functional during and after the design seismic events. (

l l l i t C-14 6.0 Conclusion On the basis of our evaluation of the seismic design criteria, analyses methods and criteria. employed, and the results obtained, we conclude that the GETR structures, systems and components important to safety, modified as proposed, will remain functional / operable, considering 4 the seismic design bases determined proper by the staff. I As noted earlier additional analyses must be provided by the licensee and reviewed by the staff prior to use of aluminum experiment capsules. In addition, the licensee should describe to the staff its pre-operational test program for the fuel flooding system. l 4 i l I

f. '

y wv.--wq- 'm y ww w y

1 e a 4 I 0.6 - 2 LIMITING LOAD N!: COMBINATIONS BASED j m ON LOCAL SOIL

i PRESSURES 9

Q 0.4 /- c:w d " INCIPIENT d LOCAL YtELDING" AT EDGE OF E SUPPORTING SOIL 4 i 3 0.2 l O C J l 4 I I 0' l 0 5 10 15 20 "UNSUPPO RTED LENGTH," FT, FIGURE 5. RESULTS OF SOIL PRESSURE ANALYSES (Reproduced from Figure 4 of Reference 2) i i t Figure C-1 EDAC

D. OFFSITE RADIOLOGICAL IMPACT OF DESIGN SEISMIC' EVENTS 1.0 Introduction The staff has evaluated the offsite radiological impact associated with the design seismic events. 2.0 Evaluation The seismic event is assumed to result in breach of the containment above and below grade. Although our analysis shows that the structural integrity of the pool and canal would be maintained (see Section C), a release of the radioactive contaminants of the pool water was assumed in order to provide a bound of the radiological consequences of this event. Such releases could occur as a result of leaks of the pool liner or evaporation from the pool. A Fuel Flooding System being installed at the GETR will automatically actuate at a low seismic activity level and supply water to the reactor core and canal fuel storage tanks so that irradiated fuel will remain submerged without operator action. ( Section ; of this Safety Evaluation addresses the Fuel Flooding system). Furthermore, an analysis of the mechanical integrity of the core (see Section C) shows that no mechanical failure of the fuel would occur as a result of the seismic event. Therefore, no fuel failure, and hence, no f ision product release from the fuel was postulated. During normal operation of the reactor, as many as five BWRSD test capsules could be located within the pool. As a result of the seismic event, it is postulated that all five capsules would fail, thereby releasing the fission products which could have accumulated within the capsule. When these experiment capsules contain fuel pins, it is assumed that previous pin failure has resulted in the accumulation of the pin's fission products within the outer barrier of the capsule. Any releases to the Experiment Effluent Holdup System (EEHS), which is assumed to fail during the earthquake, also results in a release to the atmosphere due to the assunption that the seismic event triggers containment breach above and below grade. The source tem for the potential radiological consequences of the postulated seismic event, therefore, is bounded by the radionuclide inventory of the pool / canal water, and the content of as many as five test capsules. In order to bound the potential offsite doses, it was assumed that all radioactive contaminants in the pool / canal water invent.ory, the EEHS, and of five experimental capsules is released directly to the atmos phere. This is censidered to be a worst case assumption for atmospheric release from the pool /caral water and is made in lieu of an accurate descriptive atmospheric release model from this source.

D-2 The offsite radiological consequences resulting from this postulated release are only fractions of the 10 CFR Part 100 guidelines. The 0-2 hour thyroid dose at the exclusion area boundary is 29 Rem, less than ten percent of the 10 CFR Part 100 guideline values. The maxiqum 50-year organ dose from ingestion of water at the well nearest the site boundary is less than 10m rem to the GI tract - lower large intestine, from non-sorbed 106Ru. It is concluded that no offsite radiological impact detrimental to the public health and safety will result from the postulated seismic event. 2.1 Radiological Effects on Local Water Sources The introduction of radioactive water from the pool, canal, and EEHS to the containment building basement, where it enters the ground through a breach in the containment, was also analyzed as an alternate pathway for the release of radioactive contaminants from the pool / canal water and EEHS. This assumption will result in worst case well water concentration and radiological ingestion consequence as to the GI-LLI. The accident scenario consists of loss of pool, canal, and EEriS (Experiment Effluent Holdup System) water to the basement of the containment building, where the water is assumed to enter the ground water instantaneously through a seismically-induced breach in containment below grade. Ion exchange takes place as the water moves through the soil, and a chromatographic model is appropriate - that is, selective - filtration / absorption of cations by the soil occurs. Some elements, such as ruthenium, are not absorbed, while others are strongly absorbed. It is assumed that a horizontal chromatographic column to describe the water flow from the GETR to an offsite well 875 meters southwest of the reactor is formed; the column diameter is conservatively assumed to be 100 meters. The travel time of the water from the GETR to the well is about two years. Any radionuclide which has largely decayed in two years will not reach the well in any meaningful abundance. Theradionuclidesib6 ion are Ru (one potential concern as far as well-water contaminggC (5730 year half-life). year half-life), 3H (12.3 year half-life), and i

D-3 By far, the credominant radionuclide that can reach the well water is non-sorbed 106Ru, with a concentration of 10-7 uCi/ml; this is two orders of magnitude below MPC, with a corresponding dose rate of <10m rem to the GI tract-lower large intestine, the most significant critical organ dose. Although CFR, Title 10, doo act address limiting doses for the GI-LLI, a comparison of this dose with the corresponding maximum pennissible dose specified in ICRP Publication 9, shows that the calculated dose is two orders of magnitude less than the recomended limiting dose for the GI-LLT. There is, therefore, very little impact on an individual drinking this water.

2. 2 Thyroid Dose (0-2 hr) Due to Five Failed BWRSD Capsules in the Pool The most substantial offsite consequence is due to hypothetical release of the volatiles from five failed BWRSD capsules. These capsules are nominally 101-inches in length,1.375-inches 0.D., with a 304 stainless steel.049-inch wall thickness. They are used to test BWR zircalloy-clad fuel pins, each capsule containing one pin.

They are irradiated at 40.4 KW power. Although the one percent halogen, ten percent noble gas release assumed would not result from the seismic event, some experiments require ramping to slight capsule cladding this time; a X/Q of 9 x 10 g a one percent halogen release occurs atis assumed failure. It is assumed tha 3 sec/m 131, 105 CI of g I, and 88.5 Ci of gous release of 39.8 Ci of credit given. T means the instant 1 I. The resulting 0-2 hour offsite thyrcid dose is 29 Rem. Of the analyses considered, this result is the n.ost significant, but it is still less than ten percent of the guidelines of 10 CFR Part 100. Thus, the consequences of this accident do not pose an undue risk to the public health and safety. 2.3 Instantaneous Atmosaheric Release of Pool / Canal Water Contaminants The most significant radiological consequenqe of an instantaneous release (X/Q = 9 x 10-4 sec/mh of the pool / canal water contaminants to the atmospge is release of thgodines, most significantly 5.07 mci of I and 46.9 mci of I. The resulting 0-2 hour site boundary radiological consequence is m3m rem to the thyroid. An instantaneous release of the pool / canal water radioactive contaminants is assumed because no appropriate conservative model to describe " steaming" of the canal water with concomitant tra sport of radioactive species to the atmosphere is available. A semi-infinite cloud is assumed. The whole body dose is negligibly small. Thus, the atmospheric dispersal of the radioactive contents of the pool / canal water poses no health hazard, resulting in offsite radiological consequences five orders of magnitude less than the 10 CFR Part 100 guideli 'e values.

D-4 2.4 Direct Gamma-Ray Excosure at Site Boundary The direct gamma-ray exposurg due to 225 2-hour decayed GETR fuel elements and 500,000 Ci of 6uCo stored in the canal for convenience with the canal water surface two feet (the depth restraint at the top of the lower gate) above the top of the source region are considered. The staff used a very conservative equivalent point source intensity of 1.8 x 1016 photons /sec of 1.3 Mev energy. (Thelicensee used a 1.1 x 1015 photons /sec source term). The resulting 0-2 hour dose at the exclusion area boundary is <100 mrem; less than two orders of magnitude below the whole body exposure guidelines of 10 CFR Part 100. Thus, it is concluded that the direct gamma-ray exposure at the site boundary does not pose an undue risk to the public health and safety. 3.0 -Conclusion Using appropriately conservative release fractions, meteorological, and hydrological modelling, the staff finds that, even in the event of a major seismic event leading to the types of releases mentioned herein, and even if these releases were to occur simultaneousely, which is not physically possible, the offsite radiological consequences are a small fraction of the guidelines of 10 CFR Part 100. 1 l

References 1. "GETR Safety Analyses Report," Report NED0-12622, Vallecitos Nuclear Center, General Electric, June 1977. 2. Novak, M. " Vibration of Embedded Footings and Structures," ASCE National Structural Engineering Meeting, San Francisco, Meeting Prepring 2029, April 1973. 3. Novak, M., and Y.0. Beredugo, " Vertical Vibration of Embedded Footings," Journal of the Soil Mecnanics and Foundation Division, ASCE, December 1972, pp. 1291-1310. 2 General Electric Ccmpany, letter frcm Mr. R. W. Darnit:el to USRNC, " Responses to Additional Information Recuest Regarding Seismic Scram System for One General Electric Test Reactor,' October 13, 1980. S. Barda, Felix, John M. Hanson, and W. Gene Corley, " Shear Strength of Low-Rise Walls with Soundary Elements," Portland Cement A:sociation, Researcn and Develocment Sulletin RC043.01D, Preprinted with permis-sian from ACI Sym;;osium, " Reinforced Concrete Structures in Seismic Zones, American Concrete Institute, Detroit, Micnigan,1976. 6. Williams, H.A., and J.R. Benjamin, Investigation of Shear Walls, Part 3 -- Experimental and Mathematical Studies of the Senavior of Plain and Reinformed Concrete Walled Bents Under Static Shear Loading," Department of Civil Engineering, Stanford University, Stanford, California, July 1,1953. 7. Benjamin, J. R., and H. A. Williams, " Investigation of Shear Walls, Part 6 -- Continued Experimental and Mathematical Studies of Reinforced Concrete Walled Bents Under Static Shear loading," Departnent of Civil Engineering, Stanford University, Stanford, California,1 August 1954 S. Benjamin, Jack R., and Harry A. Williams, " Investigation of Shear "t, Part 9 -- Continued Experimental and Mathematical Studies of .,...rced Concrete Walled Bents Under Static Shear Loading," Decartment of ~ivil Engineering, Stanford, Uiversity, Stanford, California,1 September 1955. 9. Benjamin, Jack R., and Harry A. Williams, " Investigation of Shear Walls, Part 12 -- Studies of Reinforced Concrete Shear Wr.ll Assemblies," Department of Civil Engineering, Stanford University, Stanford, California, December 1956.

10. General Electric Company, " Response to NRC Order to Show Causa dated 10/24/77," Novemcer 11, 1977.

11. General Electric Comcany, " Seismic Criteria and Basis for Structural Analysis of Reactor Building - Attacnment 1, " December 16, 1977.

12. General Electric Company, letter frem Mr. R. W. Dannitrel to USNRC, "Resumpticn of Operation of the General Electric Test Reactor,"

January 6, 1978.

. 13. Engineering ' Decision Analysis Company, Inc., " Seismic Analyses of Reactor Building, General Electric Test Reactor - Phase I," Prepared for General Electric Company, EDAC-ll7.217.02, February 3, 1978.

14. General Electric Company, letter from Mr. R. W. Darmitzel to Mr. Victor Stello, USNRC, " Update of Analygtical and Modification Information,"

i February 24, 1978.

15. General Electric Company, letter from Mr. R. W. Darmitzel to Mr. Victor 1

Stello, USNRC, " Resumption of Operation of the General Electric Test Reactor," April 17, 1978.

16. General Electric Company, " Updated Response to NRC Order to Show Cause Dated 10/24/)7," June 1978.

17. Engineering Decision Analysis Company, Inc., " Seismic Analysis of Reactor Building, General Electric Test Reactor - Phase 2", prepared for General Electric Company, EDAC 117-217.04, June 1978.

18. Engineering Decision Analysis Company, Inc., " Seismic Analysis of Primary Cooling System and Reactor Pressure Vessel, General Electric Test Reactor,"

Prepared for General Electric Company, EDAC 117-217.05, June 1978

19. Engineering Decision Analysis Company, Inc., " Seismic Analysis of Primary Heat Exchanger, General Electric Test Reactor" prepared for General Electric Company, EDAC 117-217.06, June 1978.

20. Engineering Decision Analysis Company, Inc., " Seismic Analysis of Reactor Pressure Vessel and Pool Drain Lines and Poison Injection Line, General Electric Test Reactor", prepared for General Electric Company, EDAC 117-217.07, June 1978.

21. Engineering Decision Analysis Company, Inc., " Seismic Analysis of Fuel Flooding System, General Electric Company, EDAC 117-217.08, June 1978.

22., Engineering Decision Analysis Company, Inc., " Qualification of Safety-Related Valves, General Electric Test Reactor", prepared for General Electric Company, EDAC 117-217.09, June 1978.

23. Engineering Decision Analysis Company, Inc., " Analysis of Lateral Restraints to Contain Heat Exchanger HE 102", prepared for General Electric Company, EDAC 117-217.10, June 1978.

24. Structural Mechanics Associates, " Structural Analysis of New Fuel Storage Tanks and Support System, General Electric Test Reactor", prepared for -

General Electric Company, June 1978.

25. Structural Mechanics Associates, " Structural Analysis of Third -Floor Missile Impact System, General Electric Test Reactor", prepared for General Electric Company, June 1978.

h-

. 26. General Electric Company, letter from Mr. R. W. Damitzel to Mr. Victor Stello, USNRC, submitting responses to NRC requests for additional information on the Phase 2 report, July 26, 1978.

27. General Electric Company, letter from Mr. R. W. Darmitzel to Mr. Victor Stello, USNRC, " Resumption of Operation of the General Electric Test.

Reactor," January 22, 1979.

28. General Electric Company, letter fromlMr. R.W. Dannitzel to Mr. Robert W.

Reid, USNRC, " Structural Modifications forthe General Electric Test Reactor," July 9, 1979.

29. General Electric Company, letter from Mr. R. W. Dannitzel to Mr. Robert W.

Reid, USNRC, " Response to Recent Requests for Additional Information - General Electric Test Reactor," September 5,1979.

30. General Electric Company, letter from Mr. R. W. Darmitzel to Mr. Robert W.

Reid, USNRC, " Response to Recent Request for Additional Information - General Electric Test Reactor," September 21, 1979.

31. General Electric Company, letter from Mr. D. L. Gilliland to Mr. Darrell G.

Eisenhut, USNRC, "Part I Response to NRC Questions, Structural Issues," April 23,1980.

32. General Electric Company, letter from Mr. R. W. Darmitzel to Mr. Darrell G.

Eisenhut, " Response to NRC Questions, Structural Issues - Part II," May 8,1980.

33. Engineering Decision Analysis Company, Inc., " Review of Seismic Design Criteria for the GETR Site," prepared for General Electric Company, EDAC 117.254.03, submitted May 8, 1980.

34. Engineering Decision Analysis Company, Inc., " Additional Investigations to Determine the Effects of Combined Vibratory Motions and Surface Rupture Offset Due to an Earthquake on the Postulated Verona Fault,"

prepared for General Electric Company, EDAC 117-253.01, Revision 1, submitted May 8,1980.

35. Engineering Decision Analysis Company, Inc., " Additional Investigations to Detennine Effects of Vibratory Motions Due to an Earthquake on the Calaveras Fault," prepared for General Electric Company, EDAC 117.253.02, submitted May 8,1980.

36. Engineering Decision Analysis Company, Inc., "Conservatisms in the Seismic Evaluations of the GETR Reactor Building," EDAC 117-254.02, submitted May 8, 1980.

37. Engineering Decision Analysis Company, Inc., " Expanded Description of Soil Pressure Analyses, Supplement I to Additional Investigations to Detennine the Effects of Combined Vibratory Motions and Surface Rupture Offset due to an Earthquake on the Postulated Verona Fault," prepared for General Electric Company, EDAC 117-253.01, Rev.1, Suppl.1, June 27,1980.

. 38. Engineering Decision Analysis Ccmpany, Inc., " Evaluations for 0.59 Ground Acceleration Case, Supplement 2 to Additional Investigations to Cetermine the Effects of Ccmoined Vibratcry Motions and Surface Rupture Offset due to an Earthquake en the Postulatad Verona Fault," prepared for General Electric Company, EDAC 117-2S3.01, Rev.1, Supplement 2, June 30,1980.

39. Engineering Decision Analysis Company, Inc., " Additional Investigation to Determine Effects of Vibratory Motions due to an Earthquake on the Calaveras Fault," prepared for General Electric Company, EDAC 117.253.02 Revision 1, June 30,1980.

JC. Engineering Decision Analysis Company, Inc., " Review of Seismic Adequacy of Piping and Equipment, General Electric Test Reactor," prepared for General Electric Company, EDAC 117-258.01, June 30, 1980.

41. Engineering Cecision Analysis Company, Inc., " Summary Report Structural Seismic Investigations of General Electric Test Reactor," prepared for General Electric Comoany, EDAC 117-258.02, July 8,1980.

42. General Electric Comoany, letter from Mr. R. W. Darnitzel to Mr. Darreil G. Eisenhut, USNRC, " Reliability and Respense Action Time for the Genera? Electric Test Reactor (GETR) Scraa System," August 14, 1980. 43. General Electric Ccmpany, letter frcm Mr. R. W. Darmit:el to Mr. Rcbert A. Clark, USNRC, submitting EDAC Report 117-258.03, " General Electric Test Reactor Response to Additional Informaticn Request Regarding Seismic Scram System," September 16, 1980. al. General Electric Cemaany, letter frem Mr. R.W. Darmitzel to Mr. Robert A. Clark, USNRC, " Responses to Questions Regarding to the General Electric Test Reacter (GETR) Fuel and Experimental Capsules," September 23, 1980. a l l

~ Appendik A NATHAN M. NEWMARK OONSULTING ENGINEERING SERVICES 1114 CIVIL ENGINEERING BUILDING URBANA. ILLINCIS 61801 29 September 1980 4 Mr. Chris Nelson Operating Reactors, Branch 3 Division of Licensing US Nuclear Regulatory Comission Washington, DC 20555 Re: Final Reports on GETR Contract NRC-03-78-150

Dear Mr. Ne; son:

We are enclosing two copies of our Final report on the structures and equipment for the GETR facility. Also the final report entitled " Seismic Evaluation of Vallecitos Site -- Basis of Earthquake Ground Motion Design Criteria" is enclosed. The latter report is a supplement to our 14 April 1980 report entitled " Seismic Evaluation of Vallecitos Site" and was prepared to supply further background in partial response to comments made at the ACRS Subcomittee meeting in June 1980. Sincerely yours, l 5d h-NM i W. J. Hall P9 Enclosures (2) cc: N. M. Newmark J. Martore 'Aco3 t l i i l 8010070 4 1 2

e NATHAN M. NEWMARK i CONSUt T;NG ENGINEERING SERVICES 1114 CIVIL ENGINEERING BUILDING URBANA. !LLINCIS 61801 29 September 1980 SEISMIC EVALUATION OF STRUCTURES AND EQUIPMENT OF THE GENERAL ELECTRIC TEST REACTOR AT THE VALLECITOS NUCLEAR CENTER l by W. J. Hall and N. M. Newmark Nathan M. Newmark Consulting Engineering Services 1211 Civil Engineering Building i Urbana, Illinois 61801 I. BACXGROUND i The purpose of this report is to provide an evaluation of the seismic resistance of structures and equipment of the Ganeral Electric Test Reactor at the Vallecitos Nuclear Center, located near Pleasanton and Sunol, California. The reactor and associated critical equipment are housed in a domed, cylindrical steel containment building; the reactor, pool and critical equipment are located within the containment building in a massive concrete core structure. Complete descriptions of the containment building and associated facilities can be found in Refs.1-4. The writers of this report have participated, either individually j or jointly, in numerous technical meetings attended by personnel of the Nuclear Regulatory Commission and/or the staff and consultants of the owners of the facility; these meetings first occurred i.i 1977 and the latest occurred on 30 July 1980. In addition there have been numerous telephone conversations wa' ,1 Ha 1 h ' cited the site twice, once +4 + - ities. In addition to the DUPLICATE DOCUMENT dditional preliminary Entire dccument previously entered into system under: M) g@dt {g._n ANO No. of pages: 3

NATHAN M. NEWMARK ! ) CONSULTING ENGINEERING SERVICES : 1211 CIVIL ENGINEERING BUILDING l URSANA. ILLINCIS 61801 29 September 1980 SEISMIC EVALUATION OF iALLECITOS SITE -- BASIS OF EARTHQUAKE GROUND MOTION DESIGN CRITERIA \\ by W. J. Hall and N. M. Newmark l Nathan M. Newmark Consulting Engineering Services ) 1211 Civil Engineering Suilding Urbana, Illinoit 61801 INTRODUCTION The purpose f this background Smarandum is to describe in more detail the basis of cur reasoning employed in arriving at the recommended seismic review criteria for the Vallecitos site as presented in our memorandum of J 14 April 1980. Our recommendations remain u 1 changed. It is important to note that the earthcuake ground motion design criteria constitute but one part of the design criteria employed in the design or review of a nuclear facil i ty. The most important seismic effects for a nuclear plant arising from earthquake-induced motions at the site normally sould be expected to occur from earthquakes having a source close to the site rather than from more distant earthquakes on the various fault systems. This observation has been taken into account in arriving at the recommendations for the Vallecitos facility since the sources are considered to be possibly very close to, or under, the site. thquake motions is DUPLICATE DOCUMENT Ifinite probabilistic Entire document previouslY ng such effects, the entered into system under: 9Q/WQ d ty$%M / ANO No. of pages: h

Appendix B EVALUATION OF SOIL PROPERTIES AND FOUNDATION BEARING CAPACITY FOR SEISMIC ANALYSIS i (to be provided by separate letter)}}