ML19257A568

Forwards Response to NUREG-0578 TMI Lessons Learned Task Force short-term Requirements Re Plant Shielding & Environ Qualification of Equipment Used in post-accident Operations

ML19257A568
Person / Time
Site:	Fort Saint Vrain
Issue date:	12/28/1979
From:	Swart F PUBLIC SERVICE CO. OF COLORADO
To:	Varga S Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-578 P-79312, NUDOCS 8001040640
Download: ML19257A568 (36)

Text

{{#Wiki_filter:.- Public service company oe omende December 28, 1979 Fort St. Vrain Unit No. 1 P-79312 Mr. Steven A. Varga Acting Assistant Director for Light Water Reactors Division of Project Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Docket #50-267

Subject:

Additional Information Regarding January 1,1980 Action Items Resulting From the Three Mile Island Unit 2 Accident

Reference:

PSC Letter P-79299, F.E. Swart to S.A. Varga dated December 12, 1979 Gentlemen: In the above referenced correspondence, PSC committed to supplying additional information to the NRC on the following NUREG-0578 sections by January 1,1980: Section 2.1.6.b - Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces / System Which May be Used in Post-Accident Operations Section 2.1.8.a - Improved Post-Accident Sampling Capability Section 2.1.8.b - Increased Range of Radiation Monitors Section 2.1.8.c - Improved In-Plant Iodine Instrumentation Under Accident Conditions g Section 2.2.2.b - Onsite Technical Support Center h I 5 g 0 gr, %cofff 0 o f (o 4'O 8001040 y g 1691 294

Mr. Steven A. Varga December 28, 1979 Page 2 In compliance with our commitment, enclosed are the January 1,1980 PSC submittals to the above sections. Should clarifications or additional information on this subject be required, please contact this office. I Very truly yours, /- g v' GM V y 7 Frederic E. Swart Nuclear Project Manager FES/MLP:pa Enclosures 1691 295

I , Section 2.1.6.b -- Design Review of Plant Shielding and Environmental Qualification of Ecuipment for Spaces / Systems Which May Be Used In Post-Accident Operations PSC December 12,1979 (P-79299) REPLY: "PSC will perform the radiation protection design reviews required by Section 2.1.6.b, utilizing the source terms recommended in Regulatory Guides 1.3,1.4, and 1.7, and will submit the results of the review to the NRC by January 1,1980. Where doses received are in excess of GDC 19 guidelines, PSC will take those steps necessary to permit post-accident operations in vital areas. Any required modifications will be completed by January 1,1981." PSC December 27,1979 (P-79312) SUBMITTAL: The assessment of post-accident operator actions in vital areas at Fort St. Vrain (FSV) indicates that doses received from a hypothetical FSV accident scenario will not be in excess of the GDC 19 guidelines for the duration of the accident, provided the FSV reactor plant exhaust filters are adequately shielded. PSC hereby commits to providing necessary shielding modifications to the FSV reactor plant exhaust filters by January 1,1981 to permit operator access to vital areas under accident conditions. The hypothetical Fort St. Vrain (FSV) accident scenario consists of the FSV Design Basis Accident (DBA) #1 combined with successive PCRV primary coolant leakage after depressurization. For clarification, the DBA #1 and PCRV leakage scenarios are explained below. DISCUSSION: To obtain a post-accident release of radioactivity equivalent to that described in Regulatory Guides 1.3,1.4 and 1.7 requires a permanent loss of all forced circulation for the FSV HTGR. This specific accident was identified as DBA #1 in FSAR Section 14.10 and Appendix D. These analyses performed by General Atomic Company at the time of licensing did not consider Regulatory Guides 1.3 and 1.4 source terms (i.e., the equivalent of the 50% of the core radiciodine and 100% of the core noble gas inventory for release to the primary coolant) appropriate for the HTGR. However, because of past precedence by the then Atomic Energy Commission (AEC) of using the above source terms, offsite doses resulting from the postulated accident were calculated and presented in the previously mentioned FSAR sections using both the General Atomic Company release assumptions and AEC TID-14844 release assumptions. In both cases the offsite doses are within 10CFR100 limits. 1691 296

DBA #1

Description:

A non-mechanistic loss of forced circulation is postulated from full power operation, where the reactor is scrammed by the plant protection system and all attempts to restore forced circulation using the multiple heat sinks, circulators and motive power for the circulators fail. Because of the large heat sink provided by the graphite core, considerable time is available to initiate primary coolant depressurization and to restore forced circulation. The FSV FSAR specifies the time available to initiate depressurization to be 5 hours, which was later amended by PSC letter P-77250 dated December 22,1977 to be 2 hours. The reduction in time was due to the capability of the helium purification system to process primary coolant during the planned blowdown of the clean primary coolant to the reactor building ventilation stack. Thus, the depressurization of the PCRV is initiated after 2 hours and completed 7 hours later (or 9 hours from the onset of the accident), at which time the PCRV has been depressurized to 5 psig. The fuel is slow to heat up due to the large heat sink provided by the core graphite. A peak average active core temperature of 5400*F is reached about 80 hours after the onset of the accident. At this temperature, the core structural integrity and geometry are not compromised since the vaporization temperature of graphite is 6900 F. Peak activity released to the primary coolant, considering decay, is reached about 24 hours into the accident. Heat removal is provided by the liner cooling system in the redistribute mode which maximizes cooling in the top head of the PCRV. Leakage of primary coolant from the PCRV is assumed to occur at a conservatively high leakage rate of 0.2% of the primary coolant inventory per day. Offsite doses were calculated for a 6 month duration of the accident, but most of the offsite dose occurs in the first 200 hours of the accident, due to fission product decay. The reactor building ventilation system maintains continuous venting of the reactor building environment at 1.5 volumes /hr during the entire period of the accident. Primary Coolant Leakace Rate During DBA #1: The FSV FSAR DBA #1 (Appendix D, page D.1-56) assumed an arbitrarily conservative and non-mechanistic estimate of PCRV leakage after the intentional depressurization by assuming that the liner has failed completely (or does not exist) and only concrete permeability controls the leakage. An internal 5 psi pressure differential was assumed which purportedly gave a PCRV leak rate of 8.33 x 10- fraction per hr (0.2%/ day). Reference was made to Question IX.7 of Amendment No. 9 of the FSV FSAR for the calcualtion of the permeation rate for the FSV PCRV concrete under these conditions. Examination of Question D.2 revealed simply the conclusion that a 5 psi positive differential pressure led to 0.2%/ day and 2 psi positive differential pressure led to 0.08%/ day. Question IX.7 also did not provide details of the calculation of the 0.2%/ day rate. However, considerable 1691 297

~ detail and a derivation was provided for the analysis of leakage rate tests at high pressures. The following equation was provided (egn.14 on page IX.7-8): W (lb/ day) = 1.13 x 10-5 in P)2 P2} A 2 (eqn. 14) Where 4P PCRV inside pressure in psig = AP = PCRV inside pressure in psig for which the net ~ compressive stress in concrete = 0 2 A Face area of concrete, ft = Concrete thickness, ft X = P) = Permeation or high side pressure, psia P2= Ambient or low side pressure, psia Numerical values were inserted for P)the following equation (egn.15 on same = 845 psig with the assumption that AP was approximately equal to P) in c Page): W = 1.13 x 10-5 9 0 5 + 9.1 x 10-7 9 0 ( 857.52 x in 12.52) = 0.043 + 602 = 600 lb/ day (eqn.15) The first item to note is that the coefficient for the second (laminar flow) term is in error which is most likely a single error 1n transcribing from equation 14 to 15 since equation 13 has the 9.1 x 10-7 coefficient. Equation 15 should read: W = 1.13 x 10-5 x _9 0 85 " + 2.2 x 10-6 9 0(857.52 2 in 12.5 ) 2. = 0.043 + 1445 = 1450 lb/ day (egn.15 revised) The second item is that the AP/4Pe term has been dropped in going from egn.14 to eqn.15, which is significant if it is assumed that these equations are appropriate for evaluating the leak rate at P) = 5 psig. 1691 298

LEAK RATE Pressure P) lb/ day %/ day (psig) Eqn 14 15 15 Revised 14 15 15 revised Given App D; Amend 9 Question D.2 5 .0019 .13 .30 .001 .07 .17 .20 Amend 9 Question D.2 2 .0003 .046 .107 .0001 .025 .059 .08 Since equation 14 is the appropriate equation, the 0.2%/ day leak rate is conservative by a factor of 200. Furthermore, the only equation that comes close to the values given in the SAR is 15 Revised, that is, AP/aP has been neglected which accounts for the factor of 200. o For purposes of plant shielding and equipment environmental evaluations, the historic 0.2%/ day is assumed to exist as an upper limit of all potential contaminated primary coolant leakage including permeability through the PCRV concrete. This is judged to be conservative since the primary coolant with any significant activity is contained within the PCRV or helium purification components contained in wells within the PCRV. Radionuclide Source Terms for DBA-1: As previously stated, the fuel within the graphite core is slow to he.atup during DBA#1. Once it has reached the FSAR fuel particle coating failure temperature of 1725 C (3137 F), the fission products are assumed, for purposes of this shielding evaluation, to be realeased per the TID-1 644 assumptions. For release to the pr.imary coolant within the PCRV, this is 100% of noble gases, 50% of the iodines and 1% others. The total activity in curies contained in the primary coolant, assuming no leakage from the PCRV, as a function of lapsed time, is given in Table 2.1.6.b-1. Consistent with TIC-14844 release assumptions, 50% of the iodines plateout within the primary coolant system resulting in a depletion of the iodine to 25% of core inventory in the reactor building air. Thus, the total activity in curies in the reactor building, assuming the upper limit of 0.2%/ day leakage (which is being purged by the reactor building ventilation system at the rate 1.5 volumes /hr), is given in Table 2.1.6.b-2. 1691 299

5 TA8LE 2.1.6.b-1 l'SV-HUREG-0578 STUDY TOTAL ACTIVITY (C1) PRESENT IN Tile PCRV PRIllARY COOLANT AT GIVE!I ELAPSED TIME (hours). PCRV PRESSU BOUNDARY REHAINS IHTACT. TID-14844 N0101ALIZAT10!! FRACTIONS USED,100% NOBLE CASES, 50% IODINE 1% OTilERS ELAPSED TIME (Hours) liUCLIDE 2 8 24 34 40 48 52 58 72 120 240 475 720 4320 Kr-88 1.05104 2.89105 2.80105 2.39104 5.89103 1.37-103 5.50102 1.76102 7.04101 0 0 0 0 0 Rb-88 8.57103 2.79105 2.80105 2.66104 6.51103 1.46403 6.07402 1.89102 7.08101 0 0 0 0 0 Zr-95 3.15101 6.66103 1.84105 2.57605 3.01105 3.59105 3.69105 3.84405 4.18105 4.12405 3.88105 3.43105 3.02405 4.60104 Ub-95 3.18101 6.74103 1.87105 2.63105 3.09105 3.69105 3.80105 3.97105 4.35105 4.37405 4.31405 4.12405 3.88605 8.90104 1-131 1.33103 3.50105 6.18106 6.91106 7.33806 ~7.88106 7.90106 7.93106 7.98106 7.57406 4.89106 2.07406 8.45105 0 1-132 1.44103 2.34405 1.79106 6.09:05 5.64105 5.61105 3.68105 2.96405 2.72105 1.76105 4.02404 4.99103 5.46602 0 1-133 2.48t03 5.30105 6.44106 5.25106 4.70106 4.12106 3.65106 3.05106 2.04106 4.84105 8.81403 0 0 0 Xc-133 5.25103 1.40 06 2.50107 2.78107 2.94407 3.14107 3.14 07 3.12107 3.09:07 2.73107 1.41407 3.86106 9.90105 0 1-135 1.98103 2.46105 1.40106 5.49105 3.31405 1.88105 1.25105 6.83104 1.78104 2.94102 0 0 0 0 Xe-135H 7.28102 8.34104 4.59605 1.72405 1.04405 5.97+04 3.91104 2.14104 5.58103 0 0 0 0 0 Xe-135 1.75103 5.43105 6.24106 3.86106 2.93106 2.11406 1.62106 1.08106 4.39105 1.01404 0' O O O Ba-140 5.44101 1.44104 2.58105 2.92105 3.13105 3.39105 3.42805 3.45105 3.54105 3.57+05 2.701051.56105 8.80104 0 1.a-140 3.34101 7.37403 2.01405 2.60105 2.93105 3.36105 3.43105 3.54105 3.75105 3.96105 3.10105 1.80t05 1.01405 0 %Q m CD CZ)

6 TABLE 2.1.6.b-2 'SV-NUREG-0578 STUDY TOTAL ACTIVITY (C1) PRESENT IN 'IllE REACTOR BUILDING An!OSPIIERE AT GIVEN ELAPSED TIHE (hours). PCRV LEAK RATE TO BUILDING 0.2%/ DAY. REACTOR BUILDING VENTED AT 1.5 VOLUHES/IIR. TID-14844 NORMALIZED FRACTIONS USEll,100% NOBLE GASES, 25% IODINE,1% OTilERS ELAPSED TIME (Hours) HUCLIDE 2 8 24 34 40 48 _ 52 58 72 120 240 475 720 4320 Kr-88 3.77-01 1.31101 1.33101 1.32'100 3.22-01 7.10-02 3.00-02 9.23-03 3.38-03 0 0 0 0 0 Rb-88 3.55-01.1.37101 1.42101 1.48100 3.50-01 7.77-02 3.34-02 1.02-0. 3.61-03 0 0 0 0 0 Zr-95

1. 20-03 3. 29-01 9. 81100 1.40 601 1.64101 1. 97101 2.04101 2.12 : 01 2. 31101 2. 29101 2.16 t01 1. 914 01 1.68101 2.56100 Ub-95 1.21-03 3.33-01 9.98100'1.43101 1.69801 2.02101 2.10101 2.19101 2.41101 2.43101 2.39101 2.29101 2.16101 4.94400 2.52-02 8.6'4100 1.65102 1.90102 2.02102 2.17102 2.19102 2.20102 2.21102 2.10102 1.36102 5.76101 2.35401 I-131 0

1-132 2.57-02 5.24100 4.24101 1.58101 1.46101 1.46101 1.05101 8.46100 7.75100 5.14400 1.30100 1.61-01 1.77-02 0 I-133 4.68-02 1.30101 1.70102 1.44102 1.29102 1.13:02 1.01-102 8.45101 5.65101 1.34101 2.45-01 0 0 0 Xc-133 1.99-01 6.94101 1.34103 1.54103 1.63103 1.75103 1.75103 1.75103 1.75103 1.52103 7.85102 2.14-102 5.50101 0 t 1-135 3.68-02 5.89100 3.60101 1.51401 9.05100 5.09100 3.47100 1.89100 4.89-01 0 0 0 0 0 Xc-135H 3.14-02 7.71400 8.59101 7.38-101 5.53101 3.53101 2.75101 1.81401 6.20100 1.07-01 0 0 0 0 Xe-135 6.75-02 2.73101 3.40102 2.26102 1.72402 1.22402 9.56-101 6.40101 2.56101 1.01400 0 0 0 0 Ba-140 2.06-03 7.12-01 1. 38101 1.61101 1. 72101 1.874 01 1.89101 1.91101 1.96101 1.98101 1.50101 8.67400 4.89 t00 0-1.a-140 1.27-03 3.66-01 1.08101 1.43101 1.61101 1.85101 1.90f01 1.96101 2.08101 2.20101 1.72401 9.98100 5.62100' O 4 CD

Radiation Levels During DBA-1: Based upon TID-14844 source term release assumptions, the radiation levels were calculated in the reactor building and the control room to determine the operator accessibility. Details are described herein. Assumptions In addition to the assumptions used in deriving the source terms, the following assumptions were made for evaluating shielding adequacy: 1. Credit was taken for a 50". depletion of the iodines due to plateout in the primary coolant system prior to release to the reactor building atmosphere. 2. All fission products were assumed to remain gasborne. In other words, no plateout of fission products was contemplated. 3. All the activities were uniformly distributed throughout the free space of the reactor building or the PCRV. 4. The iodines and particulates removed by the reactor-building ventilation filters were deposited in any two of the three filters available. 5. Only major shielding such as concrete walls was considered. Reactor Building To determine the accessibility of the reactor building during the course of DBA-1, the gamma dose rate in the reactor building was calculated as a function of elapsed time. The contributing sources consist of the gasborne activity in the reactor building as a result of PCRV leakage, the primary coolant activity contained in the PCRV, and the buildup of iodines and particulates on the reactor building ventilation HEPA and charcoal adsorbers. The contribution from the ventilation filters was not considered, as the filters will be properly shielded. Two dose points were selected for the dose-rate calculation. The first point is located at the center of the space above the refueling floor (= 40 ft from the floor), and the second point is on the refueling floor directly above the refueling penetration. The PATH code described in FSAR Section 11.2.2.4 was utilized to perform the calculation. Figure 1 shows the dose rate at the first dose point. Essentially all the contributions come from the gasborne activity in the reactor building. The activity in the PCRV is relatively insignificant to the first dose point, because of a large separation distance between the source and dose point. Short-term access to the reactor building is possible. 1691 302

The dose rate at the second dose point (i.e., on the refueling floor) is given in Figure 2. The contributions from the reactor building and from the PCRV are individually represented, along with the total dose rate. The contribution from the PCRV is due to the primary coolant activity present in the interspace below the primary closure for the control rod drive. The maximum dose rate on the floor is 1.0 rem /hr, which is less than the peak dose rate of 1.4 rem /hr at the first dose point. Therefore, the refueling floor is accessible on a short-term basis. Control Room The dose rates in the control room include the contributions from the airborne activity in the reactor building atmosphere, and from the iodine and particulate activity accumulated in the plant ventilation filters. The PATH code was used to determine the contribution from each source as a function of time into accident. The dose point was located in the reactor engineer's office, as shown in Figure 3. The results of the PATH calculations are shown in Figure 3 as a function of elapsed time. It is apparent that the contribution from the airborne activity in the reactor building is relatively small or negligible as compared with that from the reactor building ventilation filters. The dose rate reaches a peak of 700 mrem /hr about one month into accident. The important nuclides are Zr95, Nb95 and La140 accumulated in the filters. The dose rate in the control room appears to be excessive for continuous manned access. Adequate shielding will be provided for the ventilation filters so that the dose rate from the filters can be reduced to an acceptable level. Summary Results The peak dose rates in the reactor building and control room are summarized below. Also indicated are the time at which the peak dose rate occurs following an accident, and the total dose accumulated over a period of 180 days from the initiation of the accident. 180 Day Peak Gamma Accumulated Location & Condition Dose Rate Time of Peak Dose (rem) Reactor Building (above 1.4 R/hr 24 hrs. 400 refueling floor) Control Room From Vent Filters (Unshielded) 700 mR/hr

= 720 hrs.

2400 From Reactor Building 3 mR/hr 24 hrs. 0.9 1691 303

Conclusion The following conclusions are reached from the review of shielding design adequacy for DBA-1 conditions and TID-14844 source term release assumptions: 1. The reactor building ventilation filters will be adequately shielded to reduce the dosage contribution from the filters. 2. Areas immediately outside the reactor building should be accessible only on a restricted basis, because of direct radiation from the activity in the reactor building. 1691 304

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Early Doerator Actions Required During DBA-1: The time frame of interest is the first 9 hours of the accident. In the first two hours, operator access to the reactor building is required to connect the high temperature filter absorber cooling coil to the reactor plant cooling water system and to open the 1" vent line to the reactor building ventilation stack. The 1" valve is V-23279 shown in Figure 4. At this point there has been no significant fuel failure and thus no release to the reactor building environment. All other required operator actior.s are accomplished remotely from the control room. These include operation of the reserve shutdown system, placing the liner cooling system in the redistribute mode, and increasing the liner cooling water pressure by increasing the cover gas pressure from the normal 2 peig to 30 psig. Between 3-1/4 and 3-3/4 hours into the accident, operator access to the reactor building is required to open the 2 inch valve (V-23271 in Figure 4) in the primary coolant flow path to the reactor building vent system. About 9 hours from the start of the accident, operator access is again required to close the two valves (V-23279 and V-23271) at the completion of the PCRV depressuri:ation to 5 psig. It is estimated the transient time and time to operate the valves will require at most 5 minutes in the reactor building per event. Post Accident Operator Actions Recuired Following DBA-1: There are no operator actions for maintenance of core cooling which require access to the reactor building after 9 hours from the onset of the accident. Following are evaluations of plant systems regarding required post accident operation capability from the control room. 1. System 46, Reactor Plant Cooling Water System This system is operating and must continue to operate for maintenance of liner cooling water. To remain operable, the following equipment must remain cperable or the following actions must be taken: a) Cooling water pumps, one per loop, must opccate. Since one pump per loop is on at the time of the accident and since they are operated remote-manual (R-M) from the Control Room (CR), this requirement is satisfied, b) The operator must be capable of increasing the cover gas pressure to the surge tanks to maintain subcooling allowance. This is R-M from the CR. c) The operator must maintain water level in the surge tanks. The tanks are level alarmed and recorded in the CR. Condensate is added R-M from the CR. 1691 308

d) The operator should know if there is flow in the system. There is a flow alarm and automatic standby pump start capability in the CR. e) The operator should be able to redistribute flow to the top head and upper core barrel regions to prevent localized boiling. This is done R-M from the CR. f) The operator must maintain service water flow to the cooling water heat exchangers. The cooling water temperature is recorded and service water flow controlled in the CR. 2. System 21, Helium Circulator Auxiliary Systy This system has been shutdown and it does not have to operate. In the non-operating mode, it is expected that a) The auxiliary system is shutdown b) The circulator isolation valves are closed c) The brake and seal are set d) Should the seal malfunction, the static seal-backup isolation system can be employed to establish a water seal in the circulator and prevent leakage of radioactive helium. This is accomplished R-M from the CR. 3. System 22, Secondary Coolant System This system has been shutdown and it does not have to be operated. In the non-operating mode, it is expected that a) Tne feedwater valves are closed b) The valves to desuperheaters, preflash tanks, and main condenser are open so that pressure cannot icild up in the steem generators c) The steam generators eventually boil dry, at which tlme the valves in 3.b) above are closed R-M from the CR. 4. System 47, Purification Coolina Water System This system provided cooling to the high tcunperature filf er absorber (HTFA) cooling coils during the initial PCRV depressurizittion. It war then shutdown and does not have to operate in any furthe? post accidert operations. 1691 309

5. System 45, Fire Protection System The Fire Protection System in the Fort St. Vrain plant design serves as a backup source of cooling water to the PCRV liner cooling system. In the unanticipated event that either of the two loops of the reactor plant cooling water system are not operable, fire water would be used to perform the PCRV liner cooling function. Local manual valve operation in the reactor building is required to use this mode of liner cooling. Operator access to these valves is possible at any time during the course of DBA-1 since the accumulated dosage would be within GDC-19 limits. 6. System 48, Alternate Cooling Method The Alternate Cooling Method (ACM) provides an alternate means of providing electrical power and control for cooling the reactor in the event of the occurrence of disruptive faults or events, such as a major fire in congested cable areas or the Three Room Control Complex. The system is provided to ensure public health and safety consequences, analyzed and presented in DBA-1 in the FSAR, are not exceeded. This system represents additional plant capability to maintain reactor cooling under DBA-1 conditions for the events stated above. It would not be used when normal control functions are maintained from the Three Room Control Complex. Should the ACM be required, its use would be initiated within the first two hours of the accident. Specifically, the following equipment in the reactor building would require local manual operation of the ACM transfer switches. a) PCRV Cooling Water Pump 1 A-(P-4601) b) PCRV Cooling Water Pump 1B-(P-46015) c) PCRV Cooling Water Pump 1C-(P-4602) d) PCRV Cooling Water Pump 10-(P-46025) e) Purification Cooling Water Pump (P-4701) f) Purification Cooling Water Pump (P-4702) g) Valve HV-2301 h) Valve HV-2302 Adequacy of Equipment and Instrumentation for DBA-1 Radiation Levels: The 400 rem accumulated gamma dose in the reactor building for the 180 day duration of DBA-1 pose no hazard to instrumentation or equipment contained therein. The same is true for 2700 rem accumulated dose in the vicinity of the reactor building ventilation filters. 1691 310

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Section 2.1.8a -- Improved Post-Accident Sampling Capability PSC December 12,1979 (P-79299) REPLY: "In the Reference 1 letter to the NRC, the first paragraph in the PSC reply to the NRC Position on Section 2.1.4 indicated that the Fort St. Vrain plant does not utilize a reactor containment similar to PWR and BWR plants. The FSV primary coolant system is completely contained within the Prestressed Concrete Reactor Vessel (PCRV) with the vessel's steel liner, steam generator tubing and PCRV penetrations and primary closures constituting the primary containment. Secondary closures on the PCRV penetrations and the PCRV concrete structure constitute the secondary containment. Thus, the NRC requirement for obtaining primary coolant and " containment" atmosphere samples is applicable solely to obtaining a primary coolant sample from within the PCRV at Fort St. Vrain. In this light, PSC commits to 1) performing a design review of the existing primary coolant sampling system, 2) performing necessary required interim minor modifications for primary coolant sample collection, in accordance with the requirements of Section 2.1.3.a by January 1, 1980... " PSC December 27,1979 (P-79312) SUBMITTAL: A design review of Fort St. Vrain post-accident sampling capability was performed using the accident scenario given in Section 2.1.6.b, and the Regulatory Guide 1.3 and 1.4 fission product release assumptions. The total activity present in the PCRV primary coolant, as a function of time, using TID-14844 assumptions is given in Table 2.1.6.b-1. Table 2.1.6.b-2 shows the resulting activity in the reactor building assuming a primary coolant leak rate to the building of 0.2%/ day with the ventilation system in operation. The analytical instrumentation facility for sampling the primary coolant, located within the reactor building, will require access under accident conditions. The activities given in Table 2.1.6.b-2 for 2 hours after initiation of the accident correspond to a 2.5 mrad / hour gamma radiation field in the reactor building as shown in Figure 1 of Section 2.1.6.b. Based upon these findings, it will be possible to sample the primary coolant throughout the postulated accident without incurring excessive personnel radiation exposures. The highest radiation levels expected in the reactor building would occur at 24 hours into the accident and correspond to approximately 1.4 Rad / hour radiation field. A procedure for obtaining primary coolant samples has existed at FSV for several years (lef.1). The procedure, prepared with the objective of minimizing personnel radiation exposures, was previously reviewed by the I&E office of the NRC. Minor modifications to the sampling procedure have been incorporated which will allow sampling the primary coolant throughout an accider.t without exceeding GDC19 exposure limits. Verification of the applicability of the procedural changes will be completed once the reactor returns to power. 1691 312

18 In addition to the ability to obtain samples of the primary coolant, Fort St. Vrain has a continuous on-line sampler (RT 9301) that monitors primary coolant activity and provides a continuous indication of fuel degradation. Remote control room readout for this system provides a continuous indication of fuel integrity without the necessity of entering the reactor building. The equipment is calibrated annually (Ref. 2) and checked weekly against primary coolant grab samples prepared and analyzed as described in Refs.1, 3 and 4. Boron and chloride analyses during the accident are not pertinent to the HTGR. Negative reactivity is assured by the control rods and by insertion of the reserve shut down balls which are solid boronated graphite materials. Chloride analysis is not required since the steam generators are essentially removed from service. The chemical impurities CO, C0, H, and CH are indicators of the 7 7 4 condition of the reactor core. Modest Tncreases in the gaseous species H ' 2 CO, CO, CH and N in the primary coolant gas would indicate that core 2 4 2 heatup has occurrec causing outgassing of graphite components. Heatup alone would not damage the graphite but would cause fuel particle failure. Large increases in H3, H 0 and C0 could indicate oxidation of core components due to water ingress. 2 indicative of air ingress and graphite oxidation. 2 (but not H ) could be Large increases in CO, N and 0 7 2 It is te be noted that small amounts of graphite oxidation would produce relatively large levels of gaseous product CO. For example, a C0 pressure of 1 atm in the primary circuit would be indicative of only about 300 lbs. of graphite oxidized or about 0.1% of the fuel element graphite. Continuous on-line monitoring equipment exist at the FSV facility for measuring the behavior of C0 and moisture. All other impurities are measured with a gas chromatograph located in the analytical instrumentation laboratory. Access to the facility will be necessary during an accident to provide additional information concerning these impurities. Chemical impurity sampling requires approximately 1 hour to complete. As shown earlier in this section, the maximum expected gamma dose will be approximately 1.5 Rem for a 1-hour exposure. Therefore, it will be possible to sample the primary coolant for chemical impurities without exceeding the GDC19 exposure limits. 1691 313

19 Section 2.1.8.b - Increased Rance of Radiation Monitors PSC December 12,1979 (P-79299) REPLY: "In our October 29, 1979 response to 2.1.8.b, we neglected to indicate that actions that are indicated by our evaluation of the noble gas effluent monitors would be implemented by January 1,1981. Our response is modified accordingly. As indicated in our response to 2.1.8.a. we do not have any postulated accident condition that will result in our present radiation monitors going off scale. We are, however, evaluating the use of temporary portable high range monitoring equipment similar to that indicated in your October 30, 1979 letter and will make every effort to make provisions to install a portable high range monitor by January 1,1980 along with necessary interim procedures for determining high level releases should our existing monitors go off scale." PSC December 27,1979 (P-79312) SUBMITTAL: 1. Nobel Gas Effluent Monitors As indicated in our response to 2.1.6.b contained in this letter, the postulated accident condition will result in the reactor building activities as given in Table 2.1.6.b-2. These values were used to calculate the average stack effluent activities as given in Table 2.1.8.b-1. The average activity of noble gases and iodines were calculated assuming a reactor building ventilation rate of 1.5 building volumes per hour. 1.1 x 10'gV noble gas effluent monitor, RT-7324, has a range of The F to 3.7 x 10 p.Ci/cc with remote readout in the control room. The instrument is calibrated quarterly in accordance with Ref. 5. The noble gas monitor is on the esseittial power bus which provides essential power supply off the emergency diesel generators upon loss of normal power. Referring to Table 2.1.8.b-1, the maximum noble gas activity expected in the exhaust stack gas occurs at 24 hours into the accid The value determined is 5 x10~gntpresentedinSection2.1.6.b. p.C1/cc, nearly an order of magnitude below the upper range capacity of radiation monigor RT-7324. Therefore, the noble gas upper range capacity of 10 ACi/cc deemed necessary for water reactors is totally inapplicable to Fort St. Vrain. 1691 314

2. Iodine Gaseous Radiation Monitors FSV currently has the capability of continuously monitoring gaseous iodine release from the reactor building exhaust. The system RT-7325 located in the turbine building, consists of a charcoal adsorption cartridge through which reactor building exhaust is pumped, a scintillation detector and a single channel pulse height analyzer. The radioiodine monitor like the noble gas monitor is on the essential power bus. The analyzer has its energy window set for 354 key, the major ggama peak for 1 dine-131. The 2 monitor has a detection range of 1.7 x 10 to 5.5 x 10 A.Ci of I-131 deposited on the charcoal cartridge and has remote readout in the control room. The instrument is calibrated quarterly in accordance with Ref.5. Several factors effect the output of the iodine monito. Noble gas activity, also contained in the building exhaust, has been shown to increase both the background and I-131 readings. Also the charcoal cartridge provides sufficient delay time to allow xenon gases to decay to their metallic daughter products which are retained in the charcoal. The noble gas and daughter product activities artificially increase the apparent iodine plateoutactivityigthemonitor. Finally, the nstruments upper limit of cpm, equivalent to 5.5 x 10{p Ci of I-131, will not be detection is 1 x 16 adequate for peak iodine release values. Table 2.1.8.b-1 shows the expected noble gas and iodine activities in the effluent gas as a function of time. The iodine activity in the exhaust gas was calculated assuming the main stack filters were 90% efficient for theI-131concentrationsgown(2.4x10gostulatedaccident(0-2 hours), iodine. During the early stages of the p,Ci/cc) will result in a total I-131 activity of 8.6 x 10 p Ci deposited on the charcoal cartridge in 1 hour. This activity is well within the capabilities of the equipment. However, during peak releases (24-34 hours) the iodine activity collected in 1 hour will be 4 orders of magnitude higher which will exceed the instruments upper limit of detection. Due to these obvious limitations, the continuous iodine monitor should be used only as a gross indicator of conditions in the exhaust stack. In order to quantify the iodine release, the charcoal cartridge will be removed at preselected times during the accident and subjected to gamma spectroscopic analysis. Sampling and calibration procedures for spectrochemically analyzing the charcoal cartridge exist and are currently in use at FSV (Refs. 3 and 4). These methods and procedures are applicable for accident situations, and therefore meet the intent of the NRC's position in the October 30, 1979 letter. 3. Hiah Range Containment Radiation Monitors fherequirementofinstallingradiationmonitorswithamaximumrange of 10 rad /hr is unjustified when applied to FSV. Figure 1 in Section 2.1.6.b of this letter shows a peak gamma dose of 1.4 rad /hr following an accident. Existing area radiation monitors (RT-93250, 93251 and 93252) are capable of accurately reading up to 10 rad /hr gamma doses. With the exception of RT-93251-1, the area radiation monitors appear to be adequate for monitoring post-accident radiation levels.

21 RT-93251-1 is positioned near the main stack filters at the elevation 4864 of the turbine building. The expected dose rates from the filter are shown in Figure 3 of Section 2.1.6.b. The maximum dose rate of 600 rad /hr calculated to be present at 1000 hours into the accident will necessitate upgrading RT-93251-1. FSV currently has a high-range area radiation monitor (RT-39250-14) located on the refueling floor east wall at elevation 4881. The monitor has a maximum range of 10 rad /hr, well above the expected dose rates from the filters. Replacement of RT-93251-1 with a detector-readout system similar to monitor RT-39250-14 will be completed by January 1,1981. 1691 316

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23 Section 2.1.8.c -- Improved In-Plant Iodine Instrumentation PSC December 12,1979 (P-79299) REPLY: "PSC will identify the areas at Fort St. Vrain which require continuous occupancy to mitigate the consequences of an accident. Acceptable portable or cart mounted iodine samplers with attached single channel analyzers are not on the market today to the best of our knowledge. To meet the requirements of the Section, PSC proposes to take air samples, utilizing charcoal filter adsorbers, from those icoations that require continuous habitability during accident conditions. Sampling will be analyzed utilizing a multi-channel analyzer, which will be located external to the reactor building to assure analytical capability in a timely fashion under accident conditions. These procedures will be in effect by January 1,1980. We will continue to evaluate the development of portable iodine monitors and will purchase such equipment if and when reliable equipment becomes available. By January 1,1981, PSC will have the capability to remove an iodine sampling cartridge to a permanent, low background, low contamination area where accident condition iodine concentrations can be accurately measured." PSC December 27,1979 (P-79312) SUBMITTAL: Due to the complex geometry of the FSV reactor building, installation of continuous on-line radioiodine monitors within the building does not appear feasible. As mentioned in the P-79299 PSC reply (above) acceptable portable iodine samplers with attached single channel analyzers are not currently available on the market today to the best of our knowledge. To meet the requirements of this section, PSC will use portable high volume air samplers utilizing charcoal filter adsorbers to obtain samples from those locations that require continuous or infrequent habitability during accident conditions. It should be noted from Section 2.1.6.b of this letter that access to the reactor building should not normally be required past 9 hours into the accident. These portable high volume air samplers exist at FSV and procedures are available for conducting air sample surveys throughout the postulated accident (Ref. 6). The objective will be to assure personnel exposures during sampling to as low as practicable. The portable air samplers have been modified to incorporate a charcoal cartridge of the same size used in the radioiodine effluent monitor. Gamma spectral analysis of the cartridge will be performed external to the reactor building to assure analytical capability in a timely fashion under accident conditions. Procedures exist for gamma counting the cartridge and providing specific activities for all iodine radionuclides expected (Refs. 3 and 4). 1691 318

4* REFERENCES 1. Fort St. Vrain Nuclear Generating Station, Health Physics Procedure HPP-14, " Analytical Instrumentation Room", May 5,1975. 2. Fort St. Vrain Nuclear Generating Station, Surveillance Requirement SR 5.4.9. A-2, " Beta Process Monitor Calibration", March 5,1974. 3. Fort St. Vrain Nuclear Generating Station, Radiochemistry Procedure RCP-9, " Sample Preparation for Gamma Spectral Analysis", July 8,1975. 4. Fort St. Vrain Nuclear Generating Station, Radiochemistry Procedure RCP-26, " Operating Procedure for Canberra Quanta System", April 26, 1976. 5. Fort St. Vrain Nuclear Generating Station, Surveillance Requirement SR 5.8.1 cd-Q", Radioactive Gaseous Effluent System Calibration", February 11, 1974. 6. Fort St. Vrain Nuclear Generating Station, Health Physics Procedure HPP-12. "High and Low Volume Air Particulate Sample Collection", April 29, 1975. 1691 319

Section 2.2.2.b - Onsite Technical Support Center October 30, 1979 NRC CLARIFICATION: 1. "By January 1,1980, each licensee should meet items A-G that follow. Each licensee is encoura upgrading of the TSC (items 2-10)ged to provide additional as soon as practical, but no later than January 1,1981..." PSC December 27,1979 (P-79312) SUBMITTAL: The PSC submittal to items 1. A-G was provided in PSC letter P-79249, dated October 29, 1979 from F.E. Swart to D.B. Vassallo. Additional information on Technical Support Center (TSC) items 2-10 is provided as follows: 2. Location The TSC that PSC plans to install at FSV is in the preliminary design phase. However, present plans are to build a reinforced concrete structure immediately East of the reactor building, inside the security boundary. The TSC will occupy a portion of the second floor of this building with access into the plant provided one level below the control room. Plan and elevation sketches of the preliminary design can be found in Figures 2.2.2.b-1, 2, 3 and 4. 3. Physical Size & Staffina The TSC will be large enough to accommodate 25 people and the necessary engineering data and information displays. 4. Activation Instrumentation will be powered from two sources. The portion provided from plant equipment will be in service at all times. The other equipment can be activated by the first individual manning the TSC by simply turning on the power. Procedures to activate the center will be prepared prior to it being placed in service. 5. Instrumentation Instrumentation in the TSC will be, after it is activated, continuously monitoring. It will not be safety grade, however, it will be reliable equipment of approximately the same quality level as the equipment in the control room. Control room action will not effect the TSC instrumentation nor will the TSC equipment degrade the contr71 room equipment. 1691 320

26 6. Instrumentation Power Supply The primary power supply to the TSC will be from an essential plant bus. Emergency power will be provided from the emergency diesel generator. The equipment in the TSC will not be interfaced with any computer systems. 7. Technical Data Plant parameters necessary for assessment of Plant conditions will be displayed by means of indicating meters, position lights and trend recorders as required. The preliminary list of indications to be monitored in the TSC is as follows: Analog Monitorina Devices Description Circulator Inlet Temperature-Loop I Circulator Inlet Temperature-Loop II Circulator l A Speed Circulator 1B Speed Circulator 1C Speed Circulator 1D Speed Feedwater Flow-Loop I Feedwater Temperature-Loop I Feedwater Flow-Loop II Feedwater Temperature-Loop II Feedwater Pressure Emergency Feedwater Pressure Main Steam Pressure-Loop I liain Steam Pressure-Loop II flain Steam Temperature-Loop I Main Steen Temperature-Loop II Hot Reheat Pressure-Loop I Hot Reheat Pressure-Loop II Hot Reheat Temperature-Loop I Hot Reheat Temperature-Loop II Cold Reheat Temperature-Loop I Cold Reheat Temperature-Loop II Emergency Condensate Pressure PCRV Cooling Water Temperature-Loop I PCRV Cooling Water Temperature-Loop II PCRV Cooling Water Flow Purification System Cooling Water Outlet Temperature-Train A Purification System Cooling Water Outlet Temperature-Train B 1691 321

27 Primary Coolant Pressure Startup Channel Wide Range Channels Moisture Monitor-Loop I Moisture Monitor-Loop II Primary Coolant GSS Gas C0 Analyzer Description Region 1 Core Outlet Gas Temperature Region 2 Core Outlet Gas Temperature Region 3 Core Outlet Gas Temperature Region 4 Core Outlet Gas Temperature Region 5 Core Outlet Gas Temperature Region 6 Core Outlet Gas Temperature Region 7 Core Outlet Gas Temperature Region 8 Core Outlet Gas Temperature Region 9 Core Outlet Gas Temperature Region 10 Core Outlet Gas Temperature Region 11 Core Outlet Gas Temperature Region 12 Core Outlet Gas Temperature Region 13 Core Outlet Gas Temperature Region 14 Core Outlet Gas Temperature Region 15 Core Outlet Gas Temperature Region 16 Core Outlet Gas Temperature Region 17 Core Outlet Gas Temperature Region 18 Core Outlet Gas Temperature Region 19 Core Outlet Gas Temperature Region 20 Core Outlet Gas Temperature Region 21 Core Outlet Gas Temperature Region 22 Core Outlet Gas Temperature Region 23 Core Outlet Gas Temperature Region 24 Core Outlet Gas Temperature Region 25 Core Outlet Gas Temperature Region 26 Core Outlet Gas Temperature Region 27 Core Outlet Gas Temperature Region 28 Core Outlet Gas Temperature Region 29 Core Outlet Gas Temperature Region 30 Core Outlet Gas Temperature Region 31 Core Outlet Gas Temperature Region 32 Core Outlet Gas Temperature Region 33 Core Outlet Gas Temperature Region 34 Core Outlet Gas Temperature Region 35 Core Outlet Gas Temperature Region 36 Core Outlet Gas Temperature Region 37 Core Outlet Gas Temperature 1691 322

28 Description Plant Gaseous Radioactivity Reactor Building Exhaust PCRV Relief Valve Piping Air Ejector Discharge Gas Waste Monitor Reactor Plant Exhaust Reactor Building To Ambient Differential Pressure Reactor Building Exhaust Flow Wind Speed Wind Direction Outside Air Temperature 480 Volt Bus i Voltage 480 Volt Bus 2 Voltage 480 Volt Bus 3 Voltage 480 Volt Bus ACM Voltage 120 '!olt Instrument Bus 1 Voltage 120 Volt Instrument Bus 2 Voltage 120 Volt Instrument Bus 3 Voltage 120 Volt Instrument Bus 4 Voltage 125 Volt DC Bus 1 Voltage 125 Volt DC Bus 2 Voltage 125 Volt DC Bus 3 Voltage Valve Position Indications LOOP I Helium Circulator l A Water Turbine Inlet Helium Circulator l A Water Turbine Outl et Helium Circulator 18 Water Turbine Inlet Helium Circulator 1B Water Turbine Outlet Feedwater Inlet Emergency Feedwater Inlet Feedwater Dump Main Steam Stop Check Emergency Condensate Steam Turbine Bypass Helium Circulator l A Steam Turbine Helium Circulator 1B Steam Turbine Helium

• Circulator l A Steam Turbine Trip Valve Helium Circulator 1B Steam Turbine Trip Valve Hot Reheat Stop Check 1691 323

29 LOOP II Helium Circulator 1C Water Turbine Inlet Helium Circulator 1C Water Turbine Ou tl e t Helium Circulator 1D Water Turbine Inlet Helium Circulator 1D Water Turbine Outlet Feedwater Inlat Emergency Feeawater Inlet Feedwater Dump Main Steam Stop Check Emergency Condensate Steam Turbine Bypass Helium Circulator 1C Steam Turbine Helium Circulator 10 Steam Turbine Helium Circulator 1C Steam Turbine Trip Valve Helium Circulator ID Steam Turbine Trip Valve Hot Reheat Stop Check LOOP I - LOOP II COMMON Purification Train A Inlet Purification Train B Inlet PCRV Relief V-11702 PCRV Relief V-ll710 Technical data such as P & I diagrams, electrical schematics, FSAR, Technical Specifications etc., to assess the Plant Safety System Parameters, In-Plant Radiological Parameters for FSV will also be available in the TSC. 8. Data Transmission PSC will, in the design of the data links between the control room and the TSC, investigate ways in which the data can be transmitted to other locations. 9. Structural Intearity A. Refer to item 2 above for building information. 10. Habitability Monitoring equipment for direct radiation and airborne radioactive contaminants will be installed in the TSC with alarms to indicate radiation levels in the TSC are reaching potentially dangerous levels. Action levels and procedures will be established to define when and which protective measures are to be taken. 1691 324

30 The TSC will be constructed to ensure radiation levels in the TSC will meet GDC19 requirements. The ventilation system will be designed utilizing high efficiency particulate air filters (HEPAs) and charcoal adsorber filters. The ventilation power supply will be provided with a back-up source from the ACM Diesel Generator in the event the primary electrical source fails. 1691 325

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