ML20028G145

Forwards Addl Info Re Sodium Spills Analyzed in PSAR Chapters 6 & 15.Info Will Be Added to Future PSAR Amend

ML20028G145
Person / Time
Site:	Clinch River
Issue date:	02/04/1983
From:	Longenecker J ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT
To:	Grace J Office of Nuclear Reactor Regulation
References
HQ:S:83:205, NUDOCS 8302070451
Download: ML20028G145 (13)

Text

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Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:83:205 FEBC 33 Dr. J. Nelson Grace, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Dr. Grace:

ADDITIONAL INFORMATION FOR SODIUM SPILLS PERTAINING TO PRELIMINARY SAFETY ANALYSIS REPORT (PSAR) CHAPTERS 6 AND 15 Enclosed is additional information for the Clinch River Breeder Reactor Plant regarding sodium spills, specifically sodium spills in cell 102A and spills analyzed in PSAR Chapter 15.6. This will be added to the PSAR in a future amendment.

If you have any questions regarding the enclosure, please contact Mr. W. Pasko (FTS 626-6096) or Mr. D. Florek (FTS 626-6188) of the Project Office Oak Ridge staff.

Sincerely, John R. Longenecker v Acting Director, Office of Breeder Demonstration Projects Office of Nuclear Energy Enclosure cc: Service List Standard Distribution i Licensing Distribution 8302070451 830204 PDR AD?CK 05000537 A PDR

ENCLOSURE l 6 . 2 .1.3 Design Evaluation A spectrum of postulated in-containment sodium fires has been analyzed. Table 6.2-1 summarizes the results of the analysis for the most limiting fire investigated. This parametric analysis of postulated in-containment sodium fires has shown that the most limiting accident with respect - to the containment building

. temperature and pressure retaining capability is the postulated failure of the primary sodium storage tank during maintenance, assuming the tank is full of sodium and the tank cell de-inerted and open to the upper containment volume.

The primary sodium in-containment storage tank is located below the containment operating floor (Cell 102A)i.1 a cell. The floor area beneath the tank is 850 f tgormally .

inerted The cell walls are concrete, nominally 6 feet thick. The interior surfaces of the cell are protected with steel liners.

I In the event that major maintenance requires draining of one of the primary loops, the tank will be used to store the sodium coolant. The maximum volume of sodium stored in the tank will be 35,000 gallons and the sodium temperature will te maintained at approximately 4000F. The cell atmosphere will remain inerted.

In order to identify a scenario that could present a challenge to containment integrity, a hypothetical set of conditions must be postulated to exist where the primary sodium storage tank contains its maximum volume of 35,000 gallons of sodium with the tank cell de-inerted and interfacing with the atmosphere of the RCB. The occurrence of this set of conditions is highly unlikely for the following reasons:

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(1) Maintenance activities requiring draining of one or more PHTS loops will occur very few times over the life of the plant. Thus, Cell 102A will contain 35,000 gallons very few times over the life of the plant.

(2) If 35,000 gallons of sodium is present in Cell 102A, it will be at low temperature and low (atmospheric) pressure, and it will be pumped in and out of the cell at a relatively slow rate. These factors minimize the likelihood of any significant transient loads on the sodium containing boundary. The likelihood of any leak is minimized.

(3) De-inerting cell 102A when it contains more than about 4000 gallons of sodium will be prohibited by administrative procedure.

(4) Opening any door or hatch in cell 102A when it contains more than about 4000 gallons of sodium will be prohibited by administrative procedure. Violation of this administrative procedure is not likely because of the difficulty of opening the doors and hatches of the cell. In addition, there are only three scheduled entries into Cell 102A during the life 6.2-3a

of the plant. These entries are made to perform in-service inspection of.the reactor overflow vessel and the primary sodium storage vessel. Thus, operators will not be accustomed to opening the cell; it is unlikely they would do so inadvertantly. The " door" to be used for planned entries is a massive concrete structure that must be moved with a fork-lif t truck or equivalent. This door opens to Cell 105 which is below the operating floor. Other cell openings are closed even more permanently requiring extensive effort and use of heavy equipment (e.g., cranes) to open. Further, the number of doors and hatches will be minimized to those necessary for anticipated maintenance and repair activities.

All of the above factors assure that it is highly unlikely that the cell 102A would be in communication with the RCB at all. In the analysis, the door to be used for planned maintenance was assumed to be directly open to containment.

Though this assumption was unrealistic, it provided an upper bound on the amount of direct communication of Cell 102A with the upper containment and conservatively enveloped conditions that would occur if there was a large sodium inventory in Cell 102A. Any more direct communication '

between Cell 102A and containment is incredible.

(5) A non-mechanistic instantaneous failure of the primary sodium storage tank must be hypothesized whereby the total 35,000 gallons'of sodium are spilled onto the floor of the tank cell with immediate commencement of sodium pool burning.

For purposes of evaluation, it is assumed that the hypothetical accident occurs near the end of plant lif e thereby maximizing the primary sodium coolant radiological activity. The radioisotope concentrations in the sodium coolant under these conditions are summarized in Table 15.6.1.4-4. The models and assumptions used in computing the coolant activity levels are discussed in Section 12.1.3. In addition the evaluation assumes that the tank cell environment interfaces directly with RCB environment via a hypothetical tank cell doorpassageway)

(21 Ft. .

equivalent in cross-sectional area to a The rate of sodium combustion with resultant temperature and pressure histories in the containment were computed using the G ESOFIRE (Reference 1) computer code. The time behavior of the aerosol generated as a result of sodium combustion was computed with HAA-3 computer code (Ref erences 2 & 3) . Descriptions of these codes are provided in Appendix A.  ;

Sodium fire-burning rate with resulting containment pressures, temperatures and aerosol concentration are shown as a function of time in Figures 6.2-1 through 6.2-5. Table 6.2-1 summarizes the important design values of the containment and the significant results of the analyses. Table 6.2-3 provides an itemized listing of the radioactive constituents of the aerosol resulting l from sodium burning.

6.2-4

15.6 SODIUM SPlLLS - INTRODUCTION Postulated sodium fires could possibly result in the dispersion of some radioactive material to the ahnosphere. Fires involving primary sodium coolant are of most concern since this sodium circulatas through the reactor core and accumulates radioactivity due to neutron activation and entrainment of fission products leaking f rom def active f uel . Postulated fires involving sodium used in the Ex-Vessel Storage Tank (EVST) Cooling System could also result in radiological releases. The EVST sodium is essentially non-radioactive at the beginning of plant life. However, during refueling a small quantity of primary sodium is tranf erred to the EVST along with each '

irradiated assembly, resulting in a slow buildup of radioactivity in the EVST sod i um.

Besides the potential radiological impact of postulated sodium fires, these fires can result in pressure / temperature transients. Theref ore, for each f ire the consequences are evaluated in terms of: 1) the potential individual whole body and organ doses at the site boundary and low population zone and 2) the pressure / temperature transient in the af fected cell / building. The possibility of occurrence of any of the fires considered in this section is extremely unlikely. As such, it will be shown: 1) that the potential of f-site doses I are well within the guideline limits of 10CFR100, and 2) that the pressure /

temperature transient does not exceed the design capability of the af fected cell / building.

These fires can also result In pressure, temperature and aerosol challenge to equipment contained in the cell where the fire occurs and any connected cells.

These challenges are generally mitigated by providing redundant equipment in a cell which is separate and Isolated from the cell where the fire is postulated to occur. For those cases where such separation of redundant equipment is not possible, the environments resalting f rom sodium fires have been explicitly identified as challenges to be considered in the environmental qualification program. This includes both (1) the environment inside the cell or building in which the fire occurs and (2) the environment resulting from ingestion of I

the combustion products into other buildings af ter initial release from the pl ant.

The computer codes utilized in the analysis of sodium spills and fires are SPRAY-3B, GESOFIRE, SOFIRE-I I, SPCA, and HAA-38. These codes are described in Appendix A with Identification of supporting references.

Sodium spills at potential locations other than those discussed in this section have been examined. However the resul ts of these spills were considered to be less severe in terms of radiological consequences and cell temperature / pressure transients and for this reason are not presented.

Since cells containing either primary or EVST sodium are normally closed and inerted, the potential for large postulated radioactive sodium fires exists only during maintenance, when these cells are opened and de-Inerted, and suf ficient oxygen is available to sustain combustion. A spectrum of fires, both in inerted and de-Inerted atmospheres, is investigated in this section.

15.6-1 Amend. 75 Jan. 1983 RT-nnRA

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Tne consistent application of conservative assunptions throughout the analyses presented in this section provides confidence that the consequences of the fires are within the predicted results. A number of these assunptions are generic to all the fires evaluated in this section, and are sunmarized below:

1. The radioactive content of the sodium is based on continuous plant operation f or 30 years. The design basis radioisotope concentrations were assumed present in the sodium for the accident analyses. Included in the basis and discussed in PSAR Section 11.1.5 is a design limit of 100 ppb (parts per billion) for plutonium content of the primary coolant.

2. Retention, f allout, plateout, and agglomeration of sodium aerosol in cells of buildings, whose design does not include specific safetl features to accomplish that f unction are not accounted for in the analysis. Neglecting these f actors (an assumption that all of the aerosol is available f or release to the ahnosphere) leads to over-prediction of potential of f-site exposure.

3. No credit f or non-saf ety-related f ire protection systems is taken.

4. Dispersion of aerosol released to the atmosphere was calculated utilizing the conservative ahnosphere dilution f actors (X/Q) applicable to discrete time intervals provided in Table 2.3-38 (the 95th Percentile Vaiues). Guidance provided in NRC Regulatory Guide 1.145 was followed in calculating the X/Q values. Detailed descriptions of the atmospheric dilution factors estimates are provided in Section 2.3.4.

5. Fallout of the aerosol during transit downwind was neglected.

6. The cells will be structurally designed to maintain their .

Integrity under the accident temperatures and pressures and the weight of the spilled sodium. For radiological calculations, no credit is taken for cell ahnosphere leak tightness.

7. The cell liners, catch pans, and catch pan fire suppression decks are designated as Engineered Safety Features and will have design temperatures equal to or greater than the sodium spill temperature, thus conf ining the sodium spil l.

8. Both inerted and air-filled cells will be designed to accommodate IIquid metal spills resulting from a leak in a sodium or NaK pipe / component in the cell producing the worst case spill /

temperature condition. The leak is based on a Moderate Energy Fluid System break (1/4 x pipe disneter x pipe thickness) as defined in branch technical positionn MEB3-1 with the sodium or NaK system operating at its maximum normal operating temperature and pressure.

15.6-2 Amend. 75 Jan. 1983

9. The only credit for operator action in mitigation of postulated sodlun spills is shutdown of the Na overflow system makeup pumps 30 minutes af ter plant scram for a postulated leak in the Primary Heat Transport System (see Section 15.6.1.4).

10. Analyses of liquid metal burning in Inerted cells assumes burning of all oxygen in the cell in which the liquid metal is postulated to leak and burning of all the crygen contained in cells which are environmentally connected to the cell the liquid metal leak.

11. The analysis of postulated liquid metal fires in air-filled cells does not include reaction of the iIquid metal w!th postuiated water released f rom concrote. The vaildity of. this approach is presently being vc-ifled in conjunction with the large scale sodium fires test progran discussed in Section 1.5.2.8 of the PSAR. If the test progran does not support the present analysis approach, the appropriate of fects of water release f rom concrete will be included in subsequent analyses.

Table 15.6-1 provides a summary of the initial conditions for each fire considered and the maximum of f-site dose as a percentage of the 10CFR100 guidel ine l imits. As the table Indicates, a large margin exists between the potential off-site doses and 10CFR100. A o!scussion of the pressure /

temperature transient for each event is provided in the following sections; in no case do the fires result in conditions beyond the design capabilty of the celi /butIding.

The Project is assessing the impacts of NaK spills in the Reactor Service Building and will provide the results of aerosol released from the Reactor Service Building when the assessments are completed. The aerosols released f rcrn the RSB as a result of NaK spill will be controlled so as not to af fect saf ety-related equipment.

l 15.6-2a Amend. 75 Jan. 1983 l

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( V T ABLE 15.6-1 SODIUM SPILL EVENTS Max. Off-Site Sodium Spill Location

• Dose Max. Cell Gas Section Atmosphw e Blog.

ho. Events Gallons Temp (F) Cell 5 of 10CFRIDO Press / Temp 15.6 Sodlum spills 15.6.1 Extranely Unlikely 15.6.l.1 Primary sodium in 35,000 400 Normal RCB Overflos 0.8 psig containment, stor- Air Tank Cell 0.19 1380Fee age tank failure -

durina maintenance _ Design Press 10 psig,_

/ 2 50 M

W Inerted RSB Ex-Vessel 3.8 psig 15.6.1.2 Fallure of ex-vessel ,

sodium cooling sys- Sodlum Tank 0.I 'l 254orees

- tem durIng operation CeiI Design Press 12 psig i - - - "::I;L-G OT . -

g geo SGB/ Storage Tank 3.5 psig Fallure of ex-con- J H895 450 Inerted 2.13 260 0F * *

• 15.6.1.3 fB Cell tainment primary g- sodium storage tank Design Press 4 psig h

,....._-w 35,100 1015 Inerted RCB PHTS Cell <10-4 14.4 psig 15.6.1.4 Primary Heat S80er m Transport Systern (PHTS Cell) Design Press 30 psig, piping leak -----.c -, ...-

29,200 RCB Reactor Cavity (Reactor Cavity) 750 Inerted Design Press 35 psig 10 3 g 39,000 8000F Normal SGB/ IB 0.4 psig 15.6.1.5 Intermediate Heat Alr 3

Transport Systen IB 6300F m ,

Piping leak Design Press 3 psig _ _

• RCB - Reactor Contaltwent Building R$8 - Reactor Service ButIding SGB/lB - Steam Generator Bldg / intermediate Bay PHTS - Primary Heat Transport System gg r0

o. 3 g ,a se in Containment c

• In Affected Cell (a cn N .c=

15.6.1 Extremotv Unlikelv Events 15.6.1.1 Primarv' Sodium In-Containment Storace Tank Failure Durina Maintenance 15.6.1.1.1 Identification of Causes and Accident Descriotion A detailed description of this postulated event is provided in Section 6.2.

Section 6.2 includes a complete discussion of the analysis methods and the calculated consequences for this event.

15.6.1.2 Failure of the Ex-Vessel Storace Tank Sodlum Coolina System Durina l Ooeration l

l 15.6.1.2.1 Identification of Causes and Accident Descriotion l I l There are three Ex-Vessel Storage Tank (EVST) sodlum cooling circuits, two forced convection circuits normally used (alternately) to cool sodium circulated to and from the EVST, and one backup natural convection circuit used in the event the normal circuits are unavailable. Each normal circuit is located talow grade in the Reactor Service Building (RSB); the backup loop is located sbove grade. Each cooling circuit is located in separate cells. The pump sur.tlon iIne for each circuit exits from the EVST at an elevation above the normt sodium level In the tank. There are internal downcomers within the

i. EVST which extend down below the sodlum level. A r motely operated isolation valve in the pump suction line for the normal cooling circuits is located slightly above the tank outlet elevation.

During operation, all the sodium cooling circuit cells are closed and Inerted.

The Interior surf aces of the cells are protected with a steel lincr, 3/8-in.

thick. The cell walls are nominall 4-ft. thick concrete. The free volume of the cell is approximetely 14,950 l i f l oor area i s 680 f t2 The postulated accident is a leak in g t -g 9 line in the operating normal mp cooling circuit in cell 337. In the event of this postulated accident, the other normal or backup cooling circuit would be brought on line to permit g continued EVST cooling. \lhu . up wre Is a sumed 1o occur ev-w&fow point-or 3 t.h,9_

h pump _Sec44en l inep thtriting- In -th siphoning of 'toditrm'down to the botton of ttds.h49h-point-stIction+1ne within the-EVST. This postulated r@rc hscd d results-in the maximum'*possibib"q6antity of ~sodlunt" discharged f ron-the-system Avr.Jng.cporatJ6n.-ApproximatM 7500 gal ( 57%00'Ib) of 4750F sodium would peaDilled -[nto the cel l~. -f The maximum spill postulated would require a simultaneous major piping failure plus f ailure of the ranotely operated Isolation valve (which is located in a separate environment from the spill). As such, the accident is extremely unlikely and is not expected to occur over the life of the plant.

The EVST sodlum is essentially non-radioactive at the beginning of plant life.

However, during ref uellog a small quantity of primary sodium is transferred to the EVST along with each irradlated assembly, resulting in a slow bulldup of the rodloactivity in the EVST sodium. For conservatism, it is assumed that the accident occurs at the end of plant life (30 years) and immediately following a ref ueling operation when the EVST sodium activity has reached its peak. The design basis radioisotope concentrations in the EVST sodlum under 15.6-4 Amend. 64 Jan. 193-

n INSERT A The spill volume in cell 337 assumes a leak in the 4-in. EVST return line from sodium cooling Loop 1, with Loop 1 in operation.

The spill volume is based on the loss of the loop inventory and pump out of EVST sodium down to the inlet of the suction piping within the EVST. The leak is essentially constant at an MEFS rate of 6 gpm; sodium temperature is assumed to be 600 degrees F.

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I these conditions are summarized in Table 15.6.1.2-1. Only those isotopes (O

) which make a significant contribution to the radiological content of the EVST sodium are inciuded In the Table. The models and assumptions used in computing the radionuclide concentrations are included in Sections 11.1 and 12.1.3.

b'Mo k The Design Basis spill temperature is 4dt98F4. The potential radiological consequences of this event are controlled by the extent of radioactive sodium aerosol formation. The aerosol formation is controlled by the limited amount of oxygen available in the inerted (2% 0 )2 EVST cooling equipment cell. Thus, the radiological consequences are rather Insensitive to a wide range of initial sodium release (spray or pool) conditions. This is especially true because no credit was taken for retention, plate-out, or settling of the ,

aerosol in either the EVST cooling equipment cell or the Reactor Service Building, it was conservatively assumed that all the aerosol generated during combustion was released directly to the environment.

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Amend. 64 15.6-5 Jan. 1982

TABLE 15.6.1. 2-1 i DES!GN BASIS RADIOACTIVE CONTENT OF EVST S001UM 30 YEARS REACTOR OPERATION lsotooe uCl/cm Sodium Isotooe uCl/cm Sodlum Na-24 1. 47 E+1

• l-132 1.50E-1 Na-22 5.80E-1 Sb-125 9,04E-3 Cs-137 7.10E+0 Sr-90  ?,87E-3 Cs-136 4.39E-1 Am-241 5.39E-4 Cs-134 7.10E-1 Am-242m 2.60E-5 l-131 8.90E-1 Cm-244 1. 22E-4 Pu-238 6.90E-3 Pu-239 1.85E-3 Pu-240 2.42E-3 Pu-241 1.63E-1 Pu-242 5.18E-6 H-3 1.40E-2

• Peak activity during the f uel handling cycle.

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O Amend. 64 15.6-5a Jan. 1982 l

15.6.1.2.2 Analysis of Effects and Consecuences iV) The consequences of this postulated event were determined as follows:

a. The sodlum reacts with all the available oxygen in the Inerted cel l (*2%02 ). The burning releases Na20 as aerosol.

b. The radioisotope concentrations in the aerosol are the same as the initial concentrations in the sodlum.

c. Radioactive decay during the accident is neglected,

d. No credit for retention, plate-out, or settling of the Na 90 aerosol in either the EVST cooling equipment cell or the RSB was taken, it was conservatively assumed that all the aerosol generated during combustion was released directly to the atmosphere, e Fellout of the aerosol during transit downwind was neglected.

ISPMY-3B.

. " analysis of the fire in the in eted cell Indicates that combustion is compt ted (0 2 depleted) in less 1han ours. A total of 45.4 kg of Na20 ,

containing 33.8 kg of Na, is released to the atmosphere as a result of the postulated accident. Release during specific time intervals is as follows:

Time (ttt). Mass Na Released (kg),

0-2 15.8 2-8 17.8

>B

[$ 0.2 Even though no credit for aerosol etention in the Ex-Vessel Sodium Tank Celi was taken in the analysis, the cell pressure / temperature history was computed for en evaluation of cell Inte r The results of the analysis indicate a peak cell pressure of only psig. This peak occurs ast hours following the postulated spill. The cell gas pressure decreases to less than.eI % sig after 3

$ M hours. The celi temperature increases from nominally M81@F to D

inhyL hours and then decreases gradually to approximately.Me0F hours afte the postulated spill. ggo g g ISS*!=

0 The results of the radiological assessment are provided in Table 15.6.1. 2-2. The radiological assessment was performed utilizing atmospheric dispersion factors (X/Q) in Chapter 2 of the PSAR.

15.6.1.2.3 Conclusions The calculeted transient cell nressures and temperatures are within the design )

pressure and temperatures. The offsite radiological consequences are small I f ractions of the 10 CFR 100 guidelines.

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! / mend. 64 15.6-6 Jan. 1932

.m TABL E 15.6.1.2-2 POTENTIAL OFF-SITE DOSES FOLLOWING FAILURE OF THE EVST COOLING SYSTEM Dose (Rem) l SB (0.42 mi) LPZ (2.5 ml) l Organ 10CFR100 2-hours r 30-days l Whole Body"* 25 2.59 E-2 5.31 E-3 Thyroid 300 2.20 E-2 4.52 E-3 Bone 150+ 7.13 E-1 1.46 E-1 Lung 75+ 3.51 E-2 7.20 E-3

• 2. 59 E-2 = 2. 59 x 10-2

+Not covered in 10CFR100; used as guideline values.

• • Includes both inhalation and external gamma cloud exposure.

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. O Amend. 64 15.6-7 Jan. 1982