ML20010F747

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Forwards Assessments of SEP Topics XV-12,II-4.F & IX-4
ML20010F747
Person / Time
Site: Yankee Rowe
Issue date: 08/31/1981
From: Kay J
YANKEE ATOMIC ELECTRIC CO.
To: Crutchfield D
Office of Nuclear Reactor Regulation
References
TASK-02-04.F, TASK-09-04, TASK-15-12, TASK-2-4.F, TASK-9-4, TASK-RR FYR-81-130, NUDOCS 8109110280
Download: ML20010F747 (23)
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{{#Wiki_filter:l YANKEE ATOMIC ELECTRIC COMPANY ,Y a_ g h 1671 Worcester Road, framing %om, Massachuset's 01701 uxse 2.C.2.1 h* FYR 81-130 l'f[j/,) August 31, 1981 g 9 8' f F N. N 1j d United States Nuclear Regulatory Cammission g; e Washington, D. C. 20555 j S El' 10198) m Q \\; Ws m aamuma commenses Attention: Mr. Dennis M. Crutchfield, Chief qj>- // Operating Reactors Branch #5 Division of Licensing / Lt.fe rence : (a) License No. DPR-3 (Docket No. 50-19)

Subject:

SEP Topic Assessment Completion (Topics XV-12, II-4.F, and IX-4)

Dear Sir:

Enclosed please find our assessments of the following topics: XV-12; Radiological Consequences of the Rod Ejection Accident (PWR) II-4.F; Settlement of Foundations and Buried Equipment IX-4; Boron Addition System Two additional assessments are near completion and will be submitted shortly. We trust that you find this information satisfactory; however, if you have any questions, please contact us. Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY &L J. A. Kay Senior Engineer - Licensing JAK/kab Enclosure 0 06ik e10911o2so siOsai PDR ADOCK 05000029 P PDR

Topic XV Radiological Consequences of the Rod Ejection Accident (PWR) An analysis of the radiological consequences of a postulated control-rod-ejection accident have been performed following the assumptions and procedures indicated in Regulatory Guide 1.77 (Ref.1) and the Appendix to SRP 15.4.8 (Re f. 2). This analysis has been performed even though no fuel or clad fai:ures have been predicted for either past or present core loadings. Therefore, the analysis is based on conservative values for fuel and clad failure in an effort to pec ect maximum radiological consequences in the event u that future core loadings have the potential for clad or fuel failures as a result of a control-rod-ejection accident. The specific assumptions made regarding the plant conditions prior to the postulated accident and the expected response are listed in Table 1. Detailed assumptions and par:seters not stated in Table 1 are discussed below: a. It has beca conservatively assumed that the accident is followed by a complete loss of off-site power. There fore, the plant is cooled by releasing secondary steam to the environment through the safety and relie f valves. This condition is assumed to exist for 8 hours following the accident. After this time, the reactor coolant system pressure would be such that other means of plant cooldown would be us ed. b. Initial primary coolant activity prior to the accident corresponds to the Technical Specification limits. Specifically, primary coolant iodine concentration is at a value assuming a pre-existing iodine spike has occured prior to the accident and is equivalent to a value of 60uCi/gm dose equivalent I-131. Noble gases are at a leveltog00ive the Techneial Specification primary coolant activity limit of /E. Primary coolant activity levels by nuclide are listed in Table 2. c. Releases to the environment are calculated for two different cases; Case 1: The primary coolant system is breeched as a result of the accident and core and coolant activity is released to the l containment. Releases to the environment occur as a result of primary containment leakage. 25% of the iodine and 100% of the noble gases released to the coolant are assumed to be available for release from the containment. Equilibrium core and gap activities are listed by nuclida for iodines and noble gases in Table 2. Case 2: No breech of the primary coolant system occurs. Releases to the environment occur through primary to secondary coolant system leakage (assumed to be 1 gpm for the 8 hour period during which steam release to the atmosphere may be required for plant cooldown) and subsequent release to the atmosphere via safety and relief values. l l 1

d. Secondary coolant activity prior to the accident is equivalent to 0.1 uCi/gm dose equivalent I-131. e. Iodine decontamination factor of 10 between water and steam. The' do$cs calculated and presented in Table 6 are within 10CFR part 100 guidelines (reference 3) as specified in the acceptance criteria for SRP 15.4.8 (reference 2). Reference 1: USNRC Regulatory Guide 177, " Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors," Appendix B, " Radiological Assump tions. " Reference 2: Appendix to Standard Review Plant 15.4.8, " Radiological Consequences of Control Rod Ejection Accident (PWR), November 24, 1975. Reference 3: 10CFR Par +. 100, " Reactor Site Criteria."

TABLE 1 (Sheet 1 of 2)

SUMMARY

OF PARAMETERS AND ASSUMPTIONS USED FOR THE ROD EJECTION ACCIDENT I. Data and assumptions used to estimate radioactive source from postulated accident. A. Power level 600 MWt B. Burnup Equilibrium Core Fission Product Inventory C. Percent of fuel melt 0.25% D. Percent of Clad Failure 10% E. Iodine fractions Elemental = 915 (elemental, organic Organic = 45 and particulate) Particulate = 5% F. Reactor coolant activity before the Table 2 accident G. Release of activity by nuclide Tables 3 and 4 II. Data and assumptions used to estimate activity released. A. Primary containment 0.25 per day 0 t 24hr leak rate 0.1% per day t 24hr i B. Secondary containment leak rate N/A C. Valve movement times N/A l D. Adsorption and filtration efficiency N/A E. Recirculation system parameters (flow rates versus time, N/A l missing factor, etc.) i -w

TABLE 1 (Sheet 2 v? 2) Containment spray F. parameters N/A G. Containment volumes 8.6x105 ft3 H. All other pertinent data and assumptions Main Text i III. Dispersion data A. Location of points Case 1 = ground Case II.= secondary of release level containment side steam safety leakage and relief valves - ground level B. Distances to appli-cable receptors EAB = 3100 ft. up and downstream (e.g., control room, LPZ = 2 miles upstream, 6 miles downstream exclusion boundary, and LPZ) C. X/Q's at exclusion 0-1 hour EAB = 2.84 x 10-4 sec/m3 area boundary and 1-2 hours EAB = 2.27 x 10-4 -sec/m3 LPZ (for time intervals 0-8 hours LPZ = 2.84 x 10-5 sec/m3 of 2 hrs., 8 hrs., 8-24 hours LPZ = 1 92 x 10-5 sec/m3 24 hrs., 4 days and 24-96 hours LPZ = 1.62 x 10-5 sec/m3 30 days) 96-720 hours LPZ = 1.04 x 10-5 sec/m3 IV. Dose data i A. Method of dose calculation Appendix A i B. Dose conversion assumptions Appendix A C. Peak (or f(t)) concentrations in Table 5 containment D. Doses Table 6 i .. _ _ ~, _ -

TABLE 2 SOURCE TERM FOR ROD EJECTION ACCIDENT - CONSERVATIVE a .. mary Coolant a Activity Core Gap b Prio' to Accident -Equilibruim Core Activity Radionuclide (Ci) Inventory (C1) (C1) I-131 2.85E+3* 1.69E+7 1.69E+6 1 32 9.94E+2 2.46E+7 2.46E+6 133 4 37Et3 3 38E+7 3 38E+6 134 6.22E+2 3 62E+7 3.62E+6 135 2 35E+3 3 14E+7 5 14E+6 Kr-83m 1 39E+2 1.85E+6 1.85E+5 85m 5.51E+2 4.02E+6 4.02E+5 85 4.22E+1 1.91E+5 5 73E+4 87 u.22E+2 7.26E+6 7.26E+5 88 1.10E+3 1.03E+7 1.03E+6 Xe-131m 2.17E+1 1.09E+5 1.09E+4 133m 1.84E+2 1.00E+6 1.00E+5 133 8.08E+3 3 40E+7 3 40E+6 135m 2.65E+2 7.08E+6 7.08E+5 135 1.01E+3 7 50E+6-7 50E+5 1 37 5.51E+1 2 96E+7 2.96E+6 138 2 30E+2 2.66E+7 2.66E+6

  • 2.85E+3 = 2.85 x 103 a Based on Technical Specification Limit with pre-existing iodine spike b 10% of equilibrium core iodines and noble gases, except Kr-85 = 305 l

I l i

TABLE 3 ACTIVITY RELEASED TO THE ENVIRONMENT FROM CONTAINMENT LEAKAGE p ROD EJECTION ACCIDENT Radionuclide 0-2 Hours Release (Ci) 0-30 Days Release (Ci) I-131 7.2E+1' 5.0E+3 132 7 9E+1 1.8E+2 133 1.4E+2 1.7E+3 134 7.8E+1 9.8E+1 135 1.2E+2 6.2E+2 Kr-83m 3.lE+1 1.lE+2 85m 5.8E+1 2.2E+2 85 3 2E+0 6.0E+2 87 7 5E+1 1.2E+2 88 1.4E+2 3.6E+2 89 8.0E+0 8.2E+0 Xe-131m 1.8E+0 3 4E+2 133m 1.7E+1 5.7E+2 133 5 7E+2 3.4E+4 135m 9 5E+1 4.8E+2 135 1.6E+2 3 3E+3 1 37 2 3E+1 2.3E+1 138 7 5E+1 7.7E+1 7.2E+1=Y.2x10 1 72 = l l l l

TABLE 4 ACTIVITY RELEASED TO THE ENVIRONMENT FROM STEAM GENERATOR ROD EJECTION ACCIDENT o 0-2 Hours 0-8 Hours Radionuclide Release (C1) Release (C1) I-131 5.9E+28 2.1E+3 132 6.5E+2 1.2E+3 133 1.1E+3 '3 7E+3 134 6.4E+2 8.0E+2 135 9.8E+2 2 7E+3 Kr-83m 1.3E+3 3.2E+3 85m 2.4E+3 5.8E+3 85 1 3E+2 4.6E+2 87 3 1E+3 4.4E+3 88 5.6E+3 1.1E+4 89 3.4E+2 3 4E+2 Xe-131m 7.4E+1 2.6E+2 133m 6.8E+2 2.4E+3 133 2 3E+4 8.2E+4 135m 3.6E+3 1.0E+4 135 6.3E+3 2 9E+4 137 9.6E+2 9.6E+2 138 3.2E+3 3 2E+3 T.9E+2 = 5 9 x 104 590 5 = --v~v ve n. <,v-..- --.n.---,_----n-- ... a

TABLE 5 AIRB0hNE RADIOACTIVITY INSIDE THE CONTAINMENT ROD EJECTION ACCIDENT L Maximum Containment Radionuclide Airborne Concentration (Ci/cm3) I-131 1.8E+1' 1 32 2.6E+1 133 3.6E+1 134 3.8E+1 135 3 3E+1 Kr-83m 7.8E+0 85m 1.7E+1 85 8.0E-1 87 3.lE+1 88 4.4E+1 89 5.3E+1 Xe-131m 4.6E-1 133m 4.2E+0 133 1.4E+2 135m 3.0E+1 135 3 2E+1 1 37 1 3E+2 138 1.lE+2 1 18

  • 1.8E+1 = 1.5 x 10

=

TABLE 6 OFFSITE DOSES AT GIVEN SITES DUE TO ROD EJECTION ACCIDENT e Site' Thyroid (Rem) Whole Body (Rem) Skin (R3m) a. Case 1 (Containoent Leakage) EAB (0-2 Hours) 1.4E+1** 7.8P-2 1.2E-1 LPZ (0-30 Days) 2.7E+1 3.72-2 6.6E-2 b. Case 2 (Steam Generator Release) EAB (0-2 Hours) 1.1E+2 1 7E+0 3 0E+0 LPZ (0-30 Days) 4.2E+1 4.lE+1 6.8E-1

  • EAB: Exclusion Area Boundary; LPZ: Low Population Zone
    • 1.4E+1 = 1.4x101 ig

APPENDIX A

SUMMARY

OF PARAMETERS USED FOR DOSE CALCULATION O A. Method of Dose Calculations 1. Conservative Case (a) Thyroid inhalation dose DThy = (X/Q) *B*1Q1

  • DCFi (b) Whole body gamma dose (semi-infinite cloud model)

C(X/Q) *iQi DFBi D = (c) Skin dose (beta plus gamma; semi-infinite cloud model) C(X/Q) *iQi

  • DFSi + 1.11 C(X/Q) *iQi
  • DFi D =

3 Where: DThy : Thyroid dose received during time interval of interest (rem) Skin dose (equal to the sum of the beta dose and 1.11 times D = 3 the gama dose to air, per Regulatory Guide 1.109 (rem)) C = Conversion factor, equal to 1/3600 hours sec ~1 Qi = Quantity of isotope i released during the time interval of interest (curies) Atmospheric dispersion parameter used for the (X/Q) = determination of the ground-level concentration (sec meter-3) B = Breathing rate (meter 3 _ 3ee-1) f I DCFi = Thyroid dose conversion factor for iodine isotope 1 (rem -curie-1) inhaled DFBi = Whole body gama dose conversion factor for isotope 1 (rem - hour-1/ curie - meter-5) = Gamma dose-to-air conversion factor DFi l (rem - hour-1 curie - meter-3) for isotope i i l DFSi = Beta dose-to-skin conversion factor for isotope 1 l (ren. - hour-1/ curie - meter-3) t l 1 l

The breathing rates assumed are as follows: BREATHING RATES

  • o

.. Time Interval (hours) Breathing Rate (m3/sec) 0-8 3 47 x 10-4 8-24 1.75 x 10-4 24-720 2 32 x 10-4

  • Regulatory Guide 1.4 B.

Data for dose conversion-factors are presented in Table 15A-1. C. Doses discussed in individual accident analysis.

TABLE A-1 DOSE CONVERSION FACTORSS o No. Isotope DCF DFS DF DFB 1 Br-83 0. 1.470E+02 1.667E+01 6.183E+00 2 Br-84 0. 8.390E+02 1.651E+03 1.609E+03 3 Br-85 0. 9 250E+02 0. O. 4 Br-87 0. 1.600E,03 3 638E+03 3 582E+03 5 I-129 0. 5.700E+01 2.226E+01 1.150E+01 6 I-131 1.490E+06 1.710E+02 3 359E+02 3.168E+02 7 I-132 1.430E+04 4.440E+02 2.030E+03 1 92SE+03 8 I-133 2.690E+05 3.660E+02 5.238E+02 4.947E+02 9 I-134 3.730E+03 5.230E+02 2.119E+03 2.014E+03 10 I-135 5.600E+04 2 720E+02 1.530E+03 1.478E+03 11 I-136 0. 1.260E+03 2.460E+03 2 387E+03 12 Kr-83M 0. O. 2.206E+00 8.303E-03 13 Kr-85M 0. 1.667E+02 1.410E+02 1 349E+02 14 Kr-85 0. 1.530E+02 1.965E+00 1.852E+00 15 Kr-87 0. 1.111E+03 7.048E+02 6.821E+02 16 Kr-88 0. 2 706E+02 1.734E+03 1.689E+03 17 Kr-89 0. 1.153E+03 1.836E+03 1.Tt 7E+03 18 Xe-13nt O. 5.430E+01 1.783E+01 1.040E4 01 19 Xe-133M 0. 1.135E+02 3.695E+01 2.830E+01 20 Xe-133 0. 3.490E+01 4.036E+01 3 362E+01 21 Xe-135M 0. 8.120E+01 3.839E+02 3.602E+02 22 Xe-135 0. 2.123E+02 2.196E+02 2.084E+02 23 Xe-137 0. 1 393E+03 1.722E+02 1.632E+02 24 XE-138 0. 4.715E+02 1.052E+03 1.017E+03

  • Based on Regulatory Guide 1.109, and Yankee Internal memorandum by J.

Hamawi, " Gamma Dose Correction Factors for Finite Hemispherical Clouds," Radiological Engineering Group, May 31, 1977.

  • Units: DCF (rem /Ci inhaled); DFS, DF, DFB (rem /hr per C1/m3),

l l l l

SEP Topic II-4.F Settlement of Foundations and Buried Equipment Geological Features b The Yankee site is situated on the eastern edge of the Deerfield River valley on very dense glacial till of Late Pleistocene age at an elevation of 1,12C to 1,140 feet above mean sea level (Figure 1). The till, which blankets crystalline gneiss bedrock of Cambrian age, ranges from 0 to about 80 feet in thickness beneath the site area, and from about 70 to 80 feet in thickness immediately beneath the reactor containment vessel. Underlying the till above bedrock in the southwestern part of the centainment is a 0 to 40 feet sequence of interbedded till, compact clay-silt and very compact sand. Bedrock beneath the glacial sediments is composed of hard, medium-grained, quartz-albite-biotite gneiss with an evenly-layered foliation structure which dips 30 -35 to the southeast. No cavernous lithologies or throughgoing fault structures have been detected in the site area. Hillsides in the site locale above Elevation 1,200 feet are commonly characterized by a thin cover of stony ground moraine and numerous bedrock expo ures. The valley of tr.e Deerfield River below Elevation 1,200 feet contains scattered deposits of late-glacial sand and gravel outwash and alluvium, overlying dense Late Pleistocene logement till. The till is olive gray in color, fine-grained, and contains numerous pebbles and few large fragments. The surricial stratigraphy in the immediate site area prior to plant construction comprised an upper layer of up to 30 feet of sand and gravel outwash overlying up to about 80 feet of dense lodgment till. In the southwestern part of the site area, where the bedrock surface lies at low elevations, the lodgment till is in turn underlain by an interbedded sequence of compact varved (lacustrine) clay-silt, till and very compact fine-to-coarse sand, ranging to about 80 feet in combined thickness. A series of five backhoe test pits in the area ranging to 10-foot depths, encountered fairly compact lodgment till with a d;; eloped A and B soil horizon averaging 30-40 percent gravel and cobbles by volume, with a silty sand t'o sand-silt matrix and a faintly fissile fabric oriented parallel to the ground surface. The bedrock in the site locale is comprised of a succession of Lower Cambrian gneiss, schist, and dolomitic marble and Lower Cambrian gneiss which form a local south-plunging anticlinal structure along the axis of Sherman Reservoir. The site is situated over the uniformly southeast-dipping flank of l this anticline, on quartz-albite-biotite gneiss of the Hoosac Formation. With the possible (but not apparent) exception of the dolomitic marble member of l the Cavendish Formation, all bedrock formations in the site locale are hard, ( internally welded, and not notably subject to degradation by groundwater solution effects. Foundation Material Figure 2 shows the site to be situated on glacial sediments in the lower elevation of a broad bedrock valley, with bedrock layering beneath the site dipping 30 -500 to the southeast. A blanket of till overlies bedrock for about two-thirds of the distance up the valley slope to the southeast. To the nortnwest, above the west abutaant of Sherman Dam, the hillside is supported l by bedrock whose layering nearly parallels the hillside surface. The glacial stratigraphy beneath the site, as estimated from seismic refraction profiles 1

obtained in 1956, eight borings put down at the site in 1956 and 1977, and six borings pnt down in 1978 is also shown on Figure 2. Three-dimensional depictions of the soil layering at the site are presented on Figure 3 showing A., estimated thickness of the upper sand and gravel layer outwash; B., estimated thickness of tne lower lodgment till and varved clay-silt sequence; o C., esti=ated combined total thickness of the soils overburden column; and D., ustimated topography of the underlying bedrock surface. The estimated thickness of the upper sand and gravel outwash layer is shown in its original, pre-construction configuration. Seismic survey results show that overburden cover becomes thinner at high elevations. Near the base of the valley adjacent to the power plant, the overburden thickness is as great as 200 feet. Almost all of the overburden material has seismic volocities' ranging from 5,500 rps to 7,000 fps, indicative of dense glacial till. A maximum thickness of 25 to 35 feet of lower velocity ove-burden materials (1,500 to 3,000 fps) exist at some locations. This velocity range indicates loosely-consolidated sur*1cial deposits. Bedrock velocity values range from 12,000 to 16,000 rps, indicative of fresh competent bedrock. Liquiraction of the dense glacial till founaation material is not considered possible under conceivable loading conditions, due in part to the dilatant behavior of the till in shear. Observed Settlement Settlement data on the containment structure was collected for about a year after completion of construction. Figure 4 shows the settlement reference points and Table 1 gives the observed settlement at each point. These data show that the containment settled less than 0.04 feet during the 1 year of observation. Additionally, during the recent excavation required for installation of the reactor support structure collars, a visual survey of the condition of the 23 year old concrete ring foundation was conducted. No evidence of any cracking caused by differential settlement was observed. Because all structures at Yankee are founded on a very dense glacial till and because no detrimental effects of diffential settlement have been observed in the 23 years since construction of Yankee first began, it is concluded that settlement of foundations and buried equipment will not be a safety problem at Yankee.

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SYSTEMATIC EVALUATION PROGRAM TOPIC IX-4, BORON ADDITION SYSTEM YANKEE ROWE o I~. INTRODUCTION Following a loss-of-coolant accident (LOCA), boric acid solution is injected into the reactor vessel during both modes of safct,...j e c t i on. In the initial injection mode, borated water is provided from the accumulator and the safety injection tanks. Af ter this initial period, which may last somewhere between 20-60 minutes, the Emergency Core Cooling System (ECCS) is realigned for the recirculation mode. In thic mode, borated water is recirculated from the containment sump to the reactor vessel and back to the sump through the break. A portion of the water introduced into the raactor vessel is converted into steam by the decay heat generated in the core. Since the steam contains virtually no impurities, the boric acid content in the water that was vaporized remains in the vessel. The concentration of boric acid in the core [ region will therefore continuously increase, unless a dilution flow is l provided through the core. Without the dilution flow, the concentration of boric acid will eventually reach the saturation limit and any further increase in boric acid inventory will cause its precipitation. Boric acid deposited in f the core may clog flow passages and seriously compromise the performance of l the ECCS. Topic IX-4 is intended to review the boron addition system, in f l particular with respect to boron precipation during the long term cooling mode of operation following a loss of coolant accident, to assure that the ECCS is designed and operated in such a manner that a sufficient throughflow is provided before the concentration of boric acid will reach its saturation l l l

limit. II. REVIEW GUIDELINES o There are no unique SRP sections that deal with this issue. The primary criterion used for review of this system was discussed in a memo dated January 21, 1976 entitled, " Concentration of Boric Acid in Reactor Vessel During Long Term Cooling - Method for Reviewing Appendix K Submittals." (Reference (1)). III. EVALUATION Yankee has reviewed the analytical methods of Reference (1) used to calculate the switchover time from cold leg to simultaneous hot and cold leg injection, and the minimum hot leg and cold leg injection flow rates. These have been compared to the methods used to calculate to same information which was supplied to the "RC by Reference (2). Based on this review, the two methods were determined to be comparable, and no reanalysis is required. Based on the calculations of Reference (2), the maximum allowable boron { concentration would be obtained at 26 hours af ter the initiation of safety I injection. Also, xn order to prevent boron precipitation, an injection flow i rate of approximately 25 gpm at this time is required, l I l I l i

- To prevent boron precipitation, Yankee utilizes simultaneous hot leg and cold leg injection. Plant emergency procedures require the initiation of simultaneous hot leg and cold leg injection between 20 and 24 hours following the initiation of safety injection. Both injection flow rates are identical, and are greater than the required 25 gpm. The details of the system configuration to fulfill these requirements were submitted to the NRC in Reference (3)'. IV. CONCLUSION t i The Yankee method of preventing boron precipitation is to simultaneously inject into the hot and cold legs. Based on this review, using the review criteria of Reference (1), the Yankee method for prevention of boron precipitation is acceptable. V. REFERENCES 1. Memorandum for Thomas M. Novck, Chief, Reactor Systems Branch from K. I. Parczewski, Reactor Safety Branch, dated January 21, 1076. 2. Letter from YAEC to USNRC, WYR 75/70, dated July 8,1975. 3. Letter from YAEC to USNRC, WYR 78-48, dated June 6,1978.}}

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