ML19254E504
| ML19254E504 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 10/29/1979 |
| From: | Swart F PUBLIC SERVICE CO. OF COLORADO |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19254E505 | List: |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 P-79249, NUDOCS 7911010403 | |
| Download: ML19254E504 (47) | |
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\\/i October 29, 1979 Fort St. Vrain Unit No.1 P-79249 Mr. Domeni, B. Vassallo, Acting Director Division of Project Management Office of Nuclear Reactor Regulation U.S. Nuclear. Regulatory Commission Washington, D.C.
20555 Docket #50-267
Subject:
Followup Actions Resulting From the NRC Reviews Regar. ding the Three Mile Island Unit 2 Accident
References:
- 1) HRC Letter D.B. Vassallo to J.K. Fuller Dated September 13, 1979
- 2) HRC Letter D.B. Vassallo to J.K. Fuller Dated July 25, 1979
- 3) PSC Letter P-79239, F.E. Swart to D.B. Vassallo Dated October 17, 1979 Gentlemen:
Enclosed are the PSC replies to the above referenced NRC letters concerning the NRC followup actions on the Tl1I-2 Lessons Learned Task Force Report, NUREG-0578.
It should be noted that a meeting was held on Itay 2,1979 with the Division of Project Management, Advanced Reactors Branch, to discuss the Fort St. Vrain High Temperature Gas Caoled Reactor in light of the Tl11-2 incident. The purpose of this meeting was to discuss the unique design ari operating characteristics of the Fort St. Vrain HTGR and to provide other information as requested by the Staff to allow them to make a finding as to the susceptibility of the Fort St.
Vrain HTGR to a TltI-2 type event.
As a result of this meeting, PSC was led to believe that the Staff had reached a conclusion and had issued an internal meno to the eff ect that it was not possible for the Fort St. Vrain reactor to experience an incident similar to that experienced at TMI-2, and that other incidents postulated for LWR's as a result of TM:-2 were not applicable to the Fort St. Vrain reactor design.
1281 013 7911010405 1
Mr. Domanic B. Vassallo October 29, 1979 Page 2 Since the meeting of flay 2,1979, PSC has not been included in the orders ilUREG-0578, and was not a party to various flRC/0wner G Fort St. Vrain vere literally set aside from the !!RC proceedings following the PSC and TMI-2 incident.
Then on September 13, 1979, Learned Task Force with a rather short-tern inplementat These 11RC positions and the guidance and criteria given for their imple-mentation are based on LUR technology, most of which is not applicable to gas-cooled technology.
fleedless to say, PSC considers this is unreasonably short notification, allowing very little time for evaluation and response to subject matter that LUR's have been developing since July,1979.
does not feel it can acet the implementction schedules set forth.In this respect PSC In addition, PSC's review indicates that the recommendations set forth and the guidance givan by the ilRC were developed for LWR's without consideration of gas-cooled technology.
not applicable to Fort St. Vrain.On this basis many of the recomendations are clearly eration the inherent safety features of an HTGR.Other recommendations must take into co 11ajor differences occur during the development of an accident scenario, and there are significant differences in the consequences which result from accident scenarios in com gas-cooled reactors and water-cooled reactors.
Should 30u have contact this office. questions regarding the enclosed PSC responses, please Very truly yours, l
l u f 5 (J i m v as e
Frederic E. Swart fluclear Project fianager FES/DUU:ler Enclosures 1 2 8 1 0' 1
~~~-
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n TABLE OF CONTENTS FOR PSC REPLIES TO ENCL.0SURE 6 OE NRC LETTER VASSALLO TO FULLER DATED SEPTE.'18ER 13, 1979 Section Title 2.1,1 Emer ?ncy Power Supply Requirements for the Pressurizer Heaters, Power-0perated Relief Valves and Block Valves, and Pressurizer Level Indicators in PWRs 2.1.2 Perfonnance Testing for BUR and PWR Rei cf and Safety Valves 2.1.3a Direct Indic? tion of Power-Operated Relief Valve and Safety Valve Position for PWRs and BURS 2.1.3b Instrumentation for Detection of Inadequate Core Cooling in PWRs and BWRs 2.1.4 Containment Isolation Provisions for PURs and BURS 2.1.5a Dedicated Penetrations for External Recombiners or Post-Accident Purge Syste:as 2.1.5c Capability to Install Hydrogen Recombiner at Each Light Water iluclear Power Plant 2.1.6a Integrity of Systems Outside Containment Likely to Contain Radioactive flaterials (Engineered Safety Syste:ns and Auxiliary Syste:ns) 2.1.6b Design deview of Plant Shielding and Environmental Qualifi-cation of Equipment for Spaces / Systems Which flay Be Used in Post-Accident Operations 2.1.7a Autoaatic Initiation of the Auxiliary feedwater System System for PWRs 2.1.7b Auxiliary Feedwater Flow Indication to Stea'a Generators for PURs 2.1.8a Improved Post-Accident Sampling Ca,nability 2.1.8b Increased Range of itadiation Ilonitors 2.1.8c Iuproved In-Plant Iodine Instrumentation 2.1.9 Analysis of Design and Off-flamal Transients and Accide,1ts 1281 015
e Page 2 S'ection Ti tle Instrumentation to Monitor Containment During the Course to NRC Letter of an Accident Vassallo to Fuller Dated
- 1) Containnent Pressure lionitoring Sept. 13, 1979
- 2) Containment Hydrogen lionitoring
- 3) Containment Water Level 11onitoring Installation of Renotely Operated High Point Vents in the to !<RC Letter Reactor Coolant System Vassallo to Fuller Dated Sept. 13, 1979 2.2.la Shift Supervisor's Responsibilities 2.2.lb Shift Technical Advisor 2.2.lc Shift and Relief Turnover t.ocedures 2.2.2a Control Room Access 2.2.2b Onsite Technical Support Center 2.2.2c Onsite Operational Support Center 1281 016
' Section 2.1.1 -- Emergency Power Supply Requirements for the Pressurizer Heaters, Power-0perated Relief Valves and Block Valves, and Pressurizer Level Indicators in PWRs.
NRC Position:
" Consistent with satisfying the requirements of General Design Criteria 10,14, 15,17, and 20 of Appendix A to 10 CFR Part 50 for the event of loss of offsite power, the following positions shall be implemented:
Pressurizer Heater Power Supply 1.
The pressurizer heater power supply design shall provide the capability to supply, from either the offsite power source or thu amergency power source (when offsite power is not available), a predetenained number of pressurizer heaters and associated controls necessary to establish and maintain natural circulation at hot standby conditions.
The required heaters and their controls shall be connected to the emergency buses in a manner that will provide redundant power supp?y capability.
2.
Procedures and training shall be established to make the operator aware of when and how the required pressurizer heaters shall be connected to the emergency buses.
If required, the procedures shall identify under what :onditions selected emergency loads can be shed from the emergency power source to provide sufficient capacity for the connection of the pressurizer heaters.
3.
The time required to acceaplish the connection of the preselected pressurizer heater to the emergency buses shall be consistent with the timely initiation and maintenance of natural circulation conditions.
4.
Pressurizer heater motive and control power interfaces with the emergency buses shall be accomplished through devices that have been qualified in accordance with safety-grade requirements.
Power Supply for Pressurizer Relief and Block Valves and Pressurizer Level Indicators 1.
Motive and control components of the power-operated relief valves (PORVs) shall be capable of being supplied from either the offsite power source of the emergency power source then the offsite power is not available.
2.
Motive and control components associated with the PORV block valves shall be capable of being supplied from either the offsite power source of the emergency power source when the offsite power is not available.
3.
Motive and control power connections to the emergency buses for the PORVs and their associated block valves shall be thrcugh devices that have been qualified in accordance with safety-gradr requirements.
1281 017
/
Page 2 4.
The pressurizer level indication instrument channels shall be powered from the vital instrument buses. These buses shall have the capability of being supplied from either the offsite power source or the emergency power source when offsite power is not availabe."
PSC Reply to Specific NRC Concerns:
Redundlnt emergency power for '.) pressurizer heat" s and 2) control and motive power systems for power operated relief valves anc. associated block valves 1s not required at Fort St. Vrain (FSV) for the following reasons:
A.
A pressurizer and power operated relief valves are not incorporated into the FSV primary system design, as pressurizers are not utilized to establish and maintain natural circulation at hot standay conditions, and a " feed and bleed" mode of reactor coolant systcm operation is not used at FSV for decay heat removal.
B.
As no pressurizer is incorporated into the FSV design, it follows that the requirement for powering the pressurizer level indication instrument channels from the vital instrument buses is not applicable.
Additional PSC Reply to General NRC Concerns:
The NRC has indicated concern that unavailability of the pressurizer heaters in PWR designs could necessitate the operation of high pressure emergency core systems to maintain reactor coolant pressure during operating transie *.
The frequency with which the high pressure emergency core cooling system operates (or in.the general case, the frequency with which any safety system operates) may then exceed the previously understood and accepted design basis.
A.
At FSV most safety systems are normally furctioning during routine plant operations and are not challenged to function from a standby condition. These normally operating systems are designed and intended to operate for the life of the plant with only routine maintenance and repair.
B.
The FSV safety systems and equipment that are not in service during nonnal plant operation, such as the diesel driven firewater pump and the standy diesel generators, are tested on a periodic basis per fechnical Specification surveillance requirements.
Based on actual plant experience to date, these standby safety systems and equipment have not experienced an excessive usage or a unanticipated number of operating cycles.
1281 013
'Section 2.1.2 -- Perfonnance Testing for BWR and PWR Relief and Safety Valves NRC Position:
" Pressurized water reactor and boiling water reactor licensees and appli-cants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for design basis transients and accidents.
The licensees and applicdiits shall detennine the expected valve operating conditions through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Revision 2.
The single failures applied to these analyses shall be chosen so that the dynamic forces on the safety and relief valves are maximized.
Test pressures shall be the highest predicted by conventional safety analysis procedures.
Reactor coolant system relief and safety valve qualification shall include qualifica-tion of associated cont'ol circuitry piping and supports as well as the valves themsel ves. ",
PSC Reply:
The Prestressed Cencrete Reactor Vessel (PCRV) safety valves at Fo t St.
Vrain (FSV) provide the ultimate protection against the pr.imary coclant system pressure exceeding the PCRV reference pressure of 845 psig.
Two safety valves are provided, either of which has adequate capacity to prevent the pressure in the PCRV from exceeding the PCRV reference pressure under the design basis accident conditions. The design basis accident for the PCRV safety valves, described in the FSV Final Safety Analysis Report (FSAR), Section 6.8, assumes the complete offset rupture of a steam generator subheader at the feedwater inlet end, which is the only credible means of substantially increasing the primary coolant pressure.
For the assumed accident to increase primary coolant pressure to the point where the safety valves are required to operate, it is necessary to postulate all of the following additional failures:
1.
The primary coolant moisture monitors fan o identify, shut off the feedwater to, and dump the failed steam generator, scram tno reactor and 2.
the plant pro.ective system fails to shut off the feedwater to and dump a preselected steam generator, scram the reactor on high primary coolant pressure, and 3.
the operator fails to manually scram the reactor and dump and shut off feedwater to the failed steam generator.
For the assumed accident, the only two fluids present in the PCRV are helium and steam as the result of the steam generator subheader rupture. No liquid is present in the primary system.
The PCRV relief valves are therefore required to operate passing 100% helium,100% steam or a mixture of the two gases.
Conditions of two-phase flow through the relief valves are not present at Fort St. Vrain.
The PCRV relief valves are Target Rock Model 690-000 valves utilizing three stages of control. We are informed by the manufacturer that these valves are very similar to the safety valves used on BWR's.
1281 019
Page 2 The PCRV relief valves are not required to operate under any normal plant transient or accident condition except as previously described.
Under these conditions, operation of the valves is not expected to occur during the life of the plant.
The PCRV safety valve tank, the piping leading to this tank from the penetration, and the piping and components internal to the tank conform to the requirements of Specification 11-T-1 (seismic design).
The seismic analysis perfomed on this piping and components resulted in no loss of function for a maximum horizontal and vertical force of 7.21g and 3.40g respectfully.
In addition, the PCRV safety valve tank is a secondary containment housing the piping from the PC'"/, M'-tion 'lves, rupture discs and safey relief valves.
In the event of leaat in thest. devices, the leakage will be contained within the secondary containment.
The PCRV safety valve discharge piping has also been seismically analyzed for the combination of themal, weight and pressure stresses, and an operating base earthqucke (OBE).
The analysis also considered a discharge involving weight and pressure stresses and a design basi, earthquake. All of the pipe hangers were analyzed for the worst case condition which is a combination of thermal, weight and pressure stresses, and a DBE earthquake.
For all of these conditions, the seismic analysis resulted in no loss of function to the piping and supports.
As previously discussed, the duty of the relief valves is minimal because it passes only helium, steam, or a mixture of the two.
No liquid water is passed through the valves.
Operation of this type of valve as a steam relief valve'has been proven at numerous installations over a number of years.
Thus, as has been discussed above, the PCRV safety valves, piping and supports have been adequately qualified to perform their intended function under expected operating conditions for design basis transients and accidents.
Therefore, further qualification testing is not necessary of the safety val /e and piping installation at Fort St. Vrain.
1281 020
Section 2.1.3a -- Direct Indication of Power-0perated Relief Valve and Safety Valve Position for PWRs and BWRs NRC Position:
" Reactor systen relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve position detection device or a reliable indication of flow in the discharge pipe."
PSC Reply:
The reactor containment system of the FSV reactor does not utilize power-operated ufety or relief valves.
Unlike PWR's or BWR's, operation of the safety valves utilized as ultimate protection for the PCRV is not expected during system operational transients.
Norma [ operating pressure in the primary system of the FSV reactor is about 700 psi at 100% reactor power. The pressure protection for the PCRV is designed to relieve at about 845 psig, the design working pressure of the PCRV. The coolant used is a compressable gas, and pressure changes experienced during system transients are relatively small.
No high capacity make-up systems are installed, nor would the installed primary coolant make-up systems need to be used during a small loss of cooLnt accident insofar as the reactor is capable of being adequately cooled with atmospheric pressure in the primary system at decay heat levels.
Reactor trip signals are generated as the result of low primary system pressure to minimize the heat which must be removed during a low primary coolant pressure condition.
The PCRV relief valves are preceded in the system piping by leak-tight metallic rupture disks. The interspace between the rupture disks and the relief valves is continually monitored for pressure.
An increase in pressure to 5 psig indicates a leak in the rupture disk and indicates the need for repair or replacement.
Figure 2.1.3a is included for clarification.
Refer to FSV FSAR Section 6.8.2 for additional infonnation.
The only identified mechanism for increasing reactor containment pressure to the relieving pressure of the rupture disk / relief valves would be an unchecked moisture leak into the PCRV. This condition would require the failure of the moisture monitors to detect and isolate the leaking steam generator and trip the reactor, the failure of the reactor to trip on high primary system pressure, and the failure of the operator to take corrective action.
(ReferenceFSAR Section6.8.1.)
In the event the rupture disk / safety valves do function, the interspace pressure alarm will sound in the control room, along with a high radiation alam on the relief valve vent pipe to the abnosphere. These alarms will indicate to the operator that the valve has opened. The primary coolant pressure indicators in the control room will also tell the Operator an "over-pressure" conditicn exists.
Reset presure of the relief valves is approximately 680 psig.
Reactor coolant pressure is indicated in the control room by three pressure indicators.
In the event a relief valve does not close, reactor pressure will continue to drop.
Operator response to this indication would be to manually isolate the relief valve.
1281 021
Page 2 In the event the relief valve was not isolated and the reactor coolant system depressurizes to abnospheric pressure, the reactor can still be adequately cooled at previously stated.
(Reference FSAR Section 14.4.3).
PSC therefore considers that primary coolant pressure indication is the most criticci parameter during any loss of coolant condition on a gas-cooled reactor and that.5SV PCRV relief valve flow / position instrumentation, while indirect, adequately indicates the full range of possible relief valve conditions to operating personnel.
1281 022 e
e a
6 4
A TSFNCRI CtkTSIDE RsACToa GUILDING d-EXHAUST F Lita IntET SET AT 15 PSIG n '
s, SET AT Z.5t1R/HR RAH NOTE 5 SET AT 150 PSIG NOTE PCRV VENT Fit.1ER PS l
SET @ 612 PSIG SET @ 196 PSIG g e
'G NOTE 6 PRIMARY CCOLANT 5
NOTE NOTE NOTE PS SYSTEM f
THERMAL l
SLEEVE SET (h 812 PS t G e
NOTE 7 SET O 832 PSIC OPT Li o L.C. a xL.C.
"nS!"R,2,NG m
SET O 720 LtNC 3/4.* SCH* 80
.lJ PStG SgT @ 5 PStG HEltUM STORACi
.M SY$ TEM l!. PS PS U
i T 10!4 Ll SET @ 5 PSIG N
-y'F*C*
1%
NOTES:
y U
6
- l. REDU'.0 ANT AL ARMS WHEN VALVE NOT FULLY CPEN.
- 2. CUPT 'RE JISC ASSEf t3LY.
^
P NS[0 (LO.
C1 S. DE TE CT S AFE T Y V ALVE BELLOWS LE AKAGE.
N
- 6. HAN0 WHEELS LOCkt0 OPEN.
- 7. VALVE STEM PACKING CLAND VENT LINES t4 FIG.2.l.3 a --Process flow dicgrant PCRV safety valve installation 4
Section 2.1.3b -- Instrumentation for Detection of Inadequate Core Cooling in PWRs and BURS NRC Position:
1.
" Licensees shall develop procedures to be used by the operator to recognize inadequate core cooling with currently available instru-mentation. The licensee shall provide a description of the existing instrumentation for the operators to use to recognize these conditions.
A detailed description of the analyses needed to fonn the basis for operator training and procedure development shall be provided pursuant to another short-tenn requirement, " Analysis of Off-Nonnal Conditions, Including Natural Circulation" (see Section 2.1.9 of this appendix).
In addition, each PWR shall install a primary coolant saturation meter to provide on-line indication of coolant saturation condition.
Operator instruction as to use of this meter shall include considera-tion that is not to be used exclusive of other related plant parameters.
2.
Licensees shall provide a description of any additional instrumenta-tion of controls (primary or backup) proposed for the plant to supple-ment those devices cited in the preceding section giving an unambiguous, c _sy-to-interpret indication of inadequate core cooling. A descrip-t1on of the functional design requirements for the system shall also be included.
A description of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided."
PSC Reply:
Sufficient infonnation is presently available at Fort St. Vrain to detect inadequate core cooling based upon the follcaing:
1.
The primary coolant utilized at FSV is single-phase helium gas, eliminating the possibility of core voiding due to a change in the coolant phase. Thus, primary coolant saturation meters or level indicators are not applicable to FSV.
2.
Helium circulator shaft speed indication is available in the control room for each of the four helium circulators.
3.
Pressure differential instrumentation is provided on each of the four helium circulators to indicate average helium coolant flow through the core in the control room.
4.
Core region outlet thennocouples are installed in the outlet of each of the 37 fuel regions at FSV to provide region temperature indica-tion and alana in the control room.
Inadequate primary system flow would re: ult in an increase in region temperatures which would be indicated by thennoccuple response.
5.
The ratio of core thennal power to core helium flow is measured, indicated and recorded in the control room.
1281 021 i
Page 2 t
6.
Three pressure transducers provide reactor coolant pressure input to control room indication and the plant protective system. One indica-tion is of primary coolant differential pressure across the core, which would be used by the operator to access adequacy of core cooling.
Procedures such as the " Safe Shutdown Under l'ighly Degraded Conditions" procedure presently exist at FSV to define operator action to be taken under slowly degrading plant conditions. Under extreme accident conditions, such as che design basis accidents, emergency procedures are utilized.
For accident conditions that are less severe than design basis accidents, operator corrective actions are based upon assessments of plant conditions utilizing i.he fonnal emergency procedures as " guides" rather than strictly following a fonaal proce-dure that is responsive to a specific hypothesized incident.
This operator action philosophy is appropriate as the FSV gas-cooled reactor has inherent safety characteristics (such as the ability of the core to sustain a total loss of coolant flow for a period of 30 minutes with no fuel or primary system component damage) that allow time.fer operator analysis of the plant abnomal condition pric to instituting corrective action.
PSC believrs that operator assessment of corrective measures to be taken under accident ^onditions, rather than operator actions based upon following established pro.edures, is consistent with the plant design philosophy and is in the best interest of the health and safety of the public.
1281 025
Section 2.1.4 -- Containment Isolation Provision for PilRs and BWRs NRC Position:
1.
"All containment isolation syste.. designs shall comply with the recommendations of SRP 6.2.4; i.e., that there be diversity in the parameters sensed for the initiation of containment isolation.
2.
All plants shall give careful reconsideration to the definition of essential and non-essential systems, shall identify each system detennined to be essential, shall identify each system detennined to be non-essential, shall describe the basis for selection of each essential system, shall modify their containment isolation designs accordingly, and shall report the results of the re-evaluation to the NRC.
3.
All non-essential systems shall be automatically. isolated by the containment isolation signal.
4.
The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves.
Reopening of contain.aent isolation valves shall require deliberate operator action."
PSC Rep _l :
l Diverse containment isolation is not directly applicable to the FSV plant because as designed and licens W, the FSV plant does not utilize a reactor containment similar to PilR and BilR reactor 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 penetra-tions and primary closures constituting the primary containment.
Secondary closures on the PCRV penetrations and the PCRV concrete structure constitute the secondary containment.
There is normally no radioactive primary coclant contained in piping external to the PCRV except for very small bore primary coolant sample lines used to draw samples for radiolog: cal and chemical analyses.
To further confine and process any accidental radioactive releases, the PCRV and reactor plant associated systems are located in a " reactor building".
The reactor building is a vented tertiary confinement containing a continuously operating ventilation system, including high efficiency particulate air filters (HEPAs) and charcoal adsorbers that process any accidental radioactivc releases in the ventilation system's exhaust stack.
Isolation valves are provided in all piping that passes through PCRV penetrations in compliance with Design Criterion 53 in Appendix C of the FSV FSAR, which states " penetrations that require closure for the containment function shall be protected by redundant valving and associated apparatus.
Automatic operaticn of these isolation valves is initiated by the detection of radiation in the external system, in the reactor building, or by detection of higher-than-expected flow rates as appropriate to the individual system design.
1281 026
Page 2 No transfer of potentially contaminated fluids is made automatically from one system to another or from the PCRV containment to systems outside the PCRV. All such transfers require specific operator action.
Transfers of potentially contaminated fluids from the reactor building for disposal outside the reactor building are made automatically.
- However, all contaminated fluid discharge lines are continuously monitored for radio-activity above acceptable levels, and discharges are automatically isolated and the fluid contained for processing upon receiving a high activity alann.
Containment isolation as referred to in the NRC position stated above is not applicable to the " containment" utilized at Fort
't. Vrain.
PSC has reviewed the containment philosophy implemented at Fort St. Vrain and has concluded that no i adifications to systems are required.
1281 02'
Section 2.1.5a -- Dedicated Penetrations for External Recombiners or Post-Accident Parge Systems NRC Position:
" Plants using external recombiners or purge systems for post-accident combustible gas control of the containment atmosphere should provide contain-ment isolation systems for external recombiner or purge systems that are dedicated to %t service only, that meet the redundancy and single failure requirements sf General Design Criteria 54 and 56 of Appendix A tc 10 CFR Part 50, and that are sized to sat ify the flow requirements of the recombiner or purge system."
Section 2.1.5c -- gpability to Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant NRC Position (liinority View):
1.
"All licensees of light water reactor plants shall have the capability to obtain and install recombiners in their plants within a few days following e. accident if containment access is impaired and if such a system is needed for long-term post-accident combustible gas control.
2.
The procedures and bases upon which the recombiners would be used on all clants should
'.n. the subject of a review by the licensees in considering shielding requirements and personnel exposure limitations as demonstrated to be necessary in the case of T!11-2."
PSC Reply:
The need for hydrogen recombiners and post-accident combustible gas con-trol of the containment atmosphere of the Fort St. Vrain reactor is precluded since the basic design of the FSV gas-cooled reactor is substantially different fran PUR's and BUR's.
Tee Fort St. Vrain gas-cooled reactor incorporates a ceramic core (graphite) cooled by an inert gas, helium.
Under nonaal and most abnonnal or accident conditions, no chemical reaction between reactor ccre or primary system materials and the inert p 'imary coolant helium gas is expected, and therefora no hydrogen or other combustible gases are generated.
In the event moisture is accidently injected into the primary coolant as tha result of a steau generator tube leak, reference FSAR Section 14.5.3, or due to an upset in the helium circulator bearing water system, the moisture and core graphite will under o a chemical reaction at nonnal reactor operating temperatures resulting in the production of hydrogen and carbon monoxide.
Because of this rs stion, provisions are made to continuously monitor the primary coolant im moisture and to automatically shut down the reactor in the event the dew poin; temperature of the primary coolant reachs 67 F (c 500 ppm H O at 700 psia).
Restrictions on reactor operating te'nperatures as a function 2
or primary systec moisture concentration are incorporated into the plant Tcanical Specifications and operating procedures.
1281 020
Page 2 Because hydrogen and carbon monoxide are normally expected to be present in the primary coolant, provisions have been incorporated into the plant design to coatinuously remove and process these gases.
It is therefore Public Service Company of Colorado's position that no additional provisions or processing equipment is required at the Fort St.
Vrain facility to handle the combustible gases generated in the primary system as the result of the chemical reaction of the core graphite and moisture that may cn occasion enter this system.
Primary coolant containing trace amounts of the combustible gases hydrogen and carbon monoxide would not accumulate inside the reactor confinement building even if the primaiy coolant were to be accidentally released fran the PCRV.
In the highly unlikely event a PCRV rupture disk / relief valve assembly should be called upon to function, the primary coolant passing through the assembly would be vented to the atmosphere, not into an enclosed containment structure.
In the event primary coolant should leak into the reactor confinement building, the leak would be readily detected by the installed area radiation monitors and building ventilation system exhaust stuck radiation monitors.
Operator action to identify and correct the leak would be initiated.
There would be no concern for an accumulation of combustible gases within the reactor confinement building insofar as the primary coolant would contain these gases in only trace amounts, and the continuing operation of the reactor building ventilation system would further reduce the conbustible gas concentra-tion and expel it to the atmosphere.
It is therefore Public Service Company of Colorado's position that no additional provisions are required for monitoring and controlling the reactor building atmosphere to detect and control the concentration of combustible gases during normal and/or post-accident conditions.
1281 027
Section 2.1.6a -- Integrity of Systems Outside Containnent Likely to Contain Radioactive Materials Engineered Safety Systems and Auxiliary Systems NRC Positio_n:
Applicants and licensees shall immediately implement a program to reduce leakage from systems outside containnent that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels. This program shall include the following:
1.
Immediate Leak Reduction a.
Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment.
b.
11easure actual leakage rates with system in operation and report them to the NRC.
2.
Continuing Leak Reduction Establish and implement a program of preventive maintenance to reduce leakage to as-low-as-practical levels.
This program shall include periodic integrated leak tests at a frequency not to exceed refueling cycle intervals.
PSC Reply,:
The entire primary coolant system for the Fort St. Vrain HTGR is contained within the prestressed concrete reactor vessel (PCRV). The only system that processes primary coolant is the helium purification system.
All helium purifi-cation syste.n equipment items containing significant activity, except the hydrogen removal and regeneration equipment, are enclosed in PCRV top-head penetrations and wells.
Piping fro.a equipment enclosed in PCRV wells to areas outside the PCRV have remote manual isolation valves outside the wells. There is no need for personnel access to the PCRV wells following an accident.
The regeneration and hydrogen removal equipment located outside the PCRV contain activity that corld be released through leakage or failure, but such releases would be swept away and filtered by the reactor building ventilation system. The consequences of such leakage are less significant than other postulated releases, all of which are within 10 CFR 20 limitations.
In the event of an accident involving penmnant loss of forced circulation cooling, the primary coolant loop is depressurized by transferring helium to storage via the helium purification system and the primary coolant circuit is isolated.
Dae to the inherent characteristics of an HTGR, there is anple time to complete this depressurization before the primary coolant gas temperature or the gasborne fission product activity increase to levels affecting helium purification system performance.
1281 030
Page 2 Primary coolant sampling lines are the only system external to the contain-ment that would handle liquids or gases containing large radioactive inventories after an accident. The sampling lines, only one of which is in operation at any given time, are constructed of welded 1/4" 0D x 1/16" ID stainless steel tubing and are provided with autoratic isolation valves actuated by area radia-tion monitors.
Moreover, the lines are located within the Reactor Building which provides a third confinement boundary preventing direct release of radioactive materials to the environment.
The radioactive gas waste system collects, filters and monitors waste gases generated in the reactor plant and limits their discharge to rates con-fonaing to 10 CFR 20 requirements.
The consequences of total release of the maximum radioactive inventory present in the gas waste system have been evaluated in the FSAR (Section 14.6.2) and shown to be within 10 CFR 20 limitations.
The radioactive liquid waste system permits storage and identification of liquid wastes so as to permit disposal of the wastes in accordance with 10 CFR 20 limitations.
Leakage from this system would be containeJ within the reactor building.
Each of the above systems is an active, operating system during normal plant operation.
As such, any leakage would be detected by area radiation monitoring equipment and corrected. Additional pariodic testing and preventive maintenance is not considered necessary for the Fort St. Vrain HTGR.
1281 03:
I i
1 d Environmental Qualification Section 2.1.6.b
- Design Review of Plant Shielding ans Which May Be Used In Post-t
_of Equ p ent for Spaces /Sys em i
_ Accident Operations _
flRC Positiun:
dioactivity equivalent to "llith the assu:nption of a post-accident release of ra the equivalent of 50%
thc.c described in Regulatory Guides 1.3 and 1.4 (i.e., inventory are co of the core radiciodine and 100% of the core noble gas diation and shielding in the primary coolant), each licensee shall perfom a ra ult of an accident, design review.f the spaces around systems that nay, as a resThe d l room, radwaste contain highly radioactive materials.
location of vital areas and equipment, such as the controtrol centers, control stations, emergency power supplies, motor conduring po limited or safety equipnent areas, in which personnel occupancy nay be unduly may be unduly degraded by the radiation fields of these systems.
l areas and protection Each licensee shall provide for adequate access ent or temporary shielding, The design review shall determine w or post-accident procedural controls. types of corrective a quested by Section
_PSC Reply:
i PSC will perform the radiation and shielding design r 80 for submission of the results of the review to the flRC.
2.1.6b.
1281 C3?
I i
t
Section 2.1.7.a -- Automatic Initiation of the Auxiliary Feedwater System for PWRs.
NRC Position:
" Consistent with satisfying the requiremants of General Design Critorion 20 of Appendix A to 10 CFR Part 50 with respect to the timely initiation of the auxiliary feedwater system, the following requirements shall be implemented in the short tenn:
1.
The design shall provide for the automatic initiation of the auxiliary feedwater system.
2.
The automatic initiation signals and circuits shall be designed so that a single failure will not result in the loss of auxiliary feedwater system function.
3.
Testability of the initiating signals and circuits shall be a feature of the design.
4.
The initiucing signals and circuits shall be powered fro:n the emergency buses.
5.
Manual capability to initiate the auxiliary feedwater system from the control room shall be retained and shall be implemented so that a single failure in the manual circuits will not result in the loss of system function.
6.
The a-c motor-driven pumps and valves in the auxiliary feedwater system shall be included in the automatic actuation (simultaneous and/or sequential) of the ir ads to the emergency buses.
7.
The automatic inituting signals and circuits shall be designed so that their failure will not result in the loss of manual capability to initiate the AFWS from the control room.
In the long term, the automatic initiation signals and circuits shall be upgraded in accordance with safety-grade requiremenh.
PSC Reply:
The most important difference between the HTGR and a PWR in regard to the need for automatic initiation of auxiliary feedwater is in the length of time that the reactor core can tolerate a loss of forced circulation cooling without damage.
For a PWR, this time is very short, however, for the Fort St. Vrain HTGR if forced circulation is restored within five hours, there is no significant fuel damage (i.e. fuel particle coating failures), and there is no affect on the health and safety of the public. Therefore, manual actuation of the emergency cooling water systems is acceptable.
In addition, the FSV "TGR provides several back-up sources of cooling water for each of the two ondary coolant loops:
1281 03
Page 2 a)
The emergency feedwater header autonatically supplies watar directly to the steam generators on loss of pressure in the main dwater line.
This header also supplies feedwater to the helium uf rculator water turbine drives, the circulator bearing water surge tanks, the circulator bearings and the main steam desuperheaters.
b)
In the event of total failure of all three boiler feedwater pumps, an energency condensate line is provided to supply water to the stean generators and heliun circulator water turbine drives. One or more of the fcur condensate pu nps (2-powered fron essential buses, P.-powered fro:.1 non-essential buses) nonaally supply water to this line.
Ilse of this source of cooling water is initiated by re: note manual actuation of the isolation valves following steaa generator depres su riza tion.
c)
Two auxiliary boiler feed.nter pumps (powered from essential buses) can also supply water to the emergency condensate line via renote manual actuation of isolation valves.
d)
In the event that all of the above pumps are inoperable, two fire water pu ips ((w diesel engine driven, one notor driven from essential buses) can be connected to the emergency condentate line via a remote manual actuated isolation valve or to the energency feedwater header via a removable spaol piece.
Each of these backup cooling methods is capable,f renoving stored and residual heat following a reactor scraa froa full power. Also, with the exception of the firewater system, all of the above systecis are operating systems, not standby systenis, thus their availability is, to a large extent, assured at all tines.
The existing FS" backup secondary cooling systenis neet the operability, testability and initiation time requirements appropriate for an HTGR.
1281 03'
Section 2.1.7b -- Auxiliary Feedwater Flow Indication to Steam Generators For P11Rs NRC Position:
" Consistent with srtisfying the requirements set forth in GDC 13 to provide the capability in the control room to ascertain the actual performance of the AFilS when it is called to perforn its intended function, the following require-ments shall be implemented:
1.
Safety-grade indication of auxiliary feedwater flow to each steata generator shall be provided in the control roo:n.
2.
The auxiliary feedwater flow instrument channels shall be powered from the emergency buses consistent with satisfying the emergency power diversity requirements of the auxiliary feedwater system set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9."
PSC Reply:
Instrumentation is provided in the Fort St. Vrain Control Roc:a to indicute stean generator cooling water flow, regardless of its source, under all norral and abnor.nal modes of operation.
This instru.nentation includes both safety-grade and nonsafety-grade flou detectors, transmitters, controllers, monitors, indicators and recorders, including those listed in Table 2.1.7b-1.
In addition to indicating and recording steaa generator cooling water flow in the Controi Room, the listed instrumentation includes a lca flow alarm (set at 22% of full flou) in the Control Roon, remote indication of steam generator cooling water flow at the Remote Shutdown Panel in the 480V Switchgear Roo.a and local indication of emerger:y feedwater. low.
Safety-grade instrumentation powered by essential busses is provided for all system control and reactor safety functions.
Although some of the indi-cating and recording instruuentation is not classified safety grade, it is oowered by an instrument bus that is fed by transformers connected to the 480V essential busses and is therefore considered highly reliable.
1281 c35
TABLE 2.1.7b-1 FEE 0 HATER FLOW IriSTRUMEllTATI0fl I.
Loop 1 Feeduater Flow Type 14 umber
, Location
- Safety Grade Flow Element FE 2205 Local Yes Flow Transmitter, High Rg.
FT 2205 Local Yes Flow lionitors Ff1 2205-1,2,7 135(AEER)
Yes Flow Controller FC 2205 105(CR)
Yes Flow Monitors FM 2205-4,5,6 135(AEER) flo Flow Indicator FI 2205 149 (480V) fio Flow Indicator FI 2205-1 105 (CR) flo Flow Indicator FI 2205-2 105 (CR) lio Flow Recorder FR 2205 105 (CR) llo Flow Switch Lou FSL 2205-1 170(AEER),
fio Flow Alarm Lou FAL 2205 105(CR) 11 0 Valve Position Indicator ZI 2205 105(CR)
Yes Flow Transmitter, Lou Range FT 2207 155 (TB,IiEZ)
Yes Flow I:on r
Fit 2207 105(CR) llo Flow Controller FC 2207 135A (AEER) llo Flow Transmitter (to PPS)
FT 2209 155(TB,MEZ)
Yes Flow itonitor FM 2209-1 139 (AEER)
Yes Flow Transmitter (to PPS)
FT 2211 155(TB,MEZ)
Yes Flow i:onitor Fil 2211-1 I40 (AEER)
Yes Flou Trensmitter (to PPS)
FT 2213 155 (TB, MEZ)
Yes Flow lionitor F1 2213-1 I43 (AEER)
Yes II.
Loop 1 Energency Condensate to S/G Reheater _Section Typ_e 11unbe_r_
Locatica*
Safety Grade Flow Element FE 2293 Local Yes Flow Transmitter FT 2239 1128(RB,EL4759')
Yes Flou ibnitor FM 2239 135B (AEER)
Yes Flow Controller FC 2239 105 (CR)
Yes Fl u Recorder FR 2239 105(CR)
Yes III.
Loop 2 Feeduater Flow Typ_e fiumber Location
- Safety Grade Flou Element FE 2206 Local Yes Flow Transmitter, High Rg.
FT 2206 154 (TB,itEZ)
Yes Flow Monitors Fil 2206-1,2,7 136A(AEER)
Yes Flow Controller FC 2206 105(CR)
Yes Flou ;;onitors Fil 2206-4,5,6, I36A (AEER) flo Flow Indicator FI 2206 I49 (480V)
No Flow Indicator FI 2206-1 105(CR) lio Flow Indicator FI 2206-2 105 (CR) llo Flow Recorder FR 2206 105(CR) fio Flow Switch Lou FSL 2206-1 170 (AEER) fic Flow Alana Lou FAL 2206 105(CR)
No Valve Position Indicator Z1 2206 105(CR)
Yes 1281 036
Table 2.1.7b-1 (Continued)
III.
Loop 2 Feedwater Flow (Continued)
Type
!! umber Location
- Safety Grade Flow Transmitter, ' ow Rg.
FT 2208 154 (TB,itEZ)
Yes Flow lionitor Fil 2208 136A (AEER) lio Flow Controller FC 2208 105 (CR) lio FlowTransmitter(toPPS)
FT 2210 Local Yes F1ow lionitor Fil 2210-1 139 (AEER)
Yes Flow Transmitter (to PPS' FT 2212 154(TB,MEZ)
Yes Flow !!onitor Fl1 2212-1 I40(AEER)
Yes Flow Transmitter (to PPS)
FT 2214 154(TB,!!EZ)
Yes Flow lionitor FM 2214-1 143 (AEER)
Yes IV.
Loop 2 Emergency Condensate to S/G Reheater Sections M e, fluaber Location
Yes Flow lionitor Fil 2240 136B (AEER)
Yes Flow Controller FC 2240 105 (CR)
Yes Flow Recorder FR 2240 105(CR)
Yes V.
Emergency Feedwater Type
!! umber Location
Yes F1ow lionitor Fli 2297 I45 (AEER) 11 0 Flow Indicatirg Switches High FISil 2297,8,9 Local Yes Flow Indicator FI 2297 102(CR) lio IV.
Steap Generator !!odule Feedwater Flow Type ljumber Location
- Safety Grade Flow Elealents FE 2222 12 Local Yes Flow Transmitters FT 2222 12 1125, 1131, 1134, I140 (RB,EL4799')
Yes Flow lionitors Fit 2222 12 1358, 136B (AEER) llo Flow Recorder,11ultipaint FR 2222 113 (CR) 11 0 Location by equipment rack nunber and physical location.
Physical locations given as:
CR -- Control Room AEER -- Auxiliary Electrical Equipment Roo:a 480V -- 430V Switchgear Roon 1281 03' TB, !!EZ -- Turbine Building, liezzanine Level TB, Grade -- Turbine Building, Grade Level RB, EL4759' -- Reactor Building, Elevation 4759'
Section 2.1.8a -- Improved Post-Accident Sampling Capability
!!RC Position:
"A design and operational review of the reactor coolant and containment atmosphere sampling sydems shall be per omed to deternine that capab'lity of r
personnel to promptly obt. 'n (less th
..am) a sample under accident condi-tions without incurring a radiatim eg; sun-
) any individual in exce;s of 3 and 18 3/4 Rems to the whole bsJy or extre.aitias, respectively.
Accident conditions sheuld assume a Re.: :atory Guide 1.3 or 1.4 rt ase of fission products.
If ti; review indict."
that personne' could not promptly and safety obtain the sanple", 'cMitional
- sign features or shielding should be provided to meet the criteria.
A design and operatien review of the radiological spectrum analysis faci-lities shall be perfonned to detenaine the capability to promptly (less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) quantify certain radioisotopes that are indicators of the degree of core danage.
Such radionuclides are noble gasus (which indicate cladding failure),
iodines and cesiums (which indicate high fuel temperatures), and non-voetile isotopes (which indicate fuel melting).
The initial reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release.
The review sSould also consider the ef fects of direct radiation fran piping and co'nponents in the auxiliary building and possible contamination and direct radiation fro.a air-borne effluents.
If the review indicates that the cnalyses required cannot be perfonaed in a pro: apt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to neet the criteria.
In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions.
Procedures shall be provided to perfona boron and chloride chenical analyses assuming a highly radioactive initial sanple (Regulatory Guide 1.3 or 1.4 source tena).
Both analyses shall be capable cf being coapleted pranptly; i.e., the boron sample analysis within an hour and the chloride sauple analysis within a shift."
PSC Reply:
PSC will perfona the design review of post-accident sanpling capability requested by Section 2.1.8a.
PSC has established a tentative date of January 1,1980 for submission of the results of the review to the !!RC.
Because of the short time span available until January 1,1980, PSC can make no coriitment as to the date for iuplementation of revised procedures, the description of any proposed modifications or inplementation of any proposed modification.
Following the design review to he canpletc<i by January 1,1980, cocaitnent dates for procedure revisions, description of any proposed nodification cad iuplementation of any proposed modification will be submitted.
1281 03')
Section 2.1.8b -- Increased Range of Radiation ilonitors NRC Position:
"The requirements associated with this recommendation should be considered as advanced implementation of certain requirements to be included in a revision to Regulatory Guide 1.97, " Instrumentation to follow the Course of an Accident,"
which has already been initiated, and in other Regulatory Guides, which will be promulgated in the near-tern.
1.
Noble gas effluent monitors shall be insi.alled with an extended range designed to function during accident conditions as well as during nonaal operating conditions; multiple nonitors are considered to be necessary to cover the ranges of interest, Notjle gas effluent nonitors with an upper range capacity of a.
10 pCi/cc (Xc-133) are considered to be practical and should be installed in all operating plants.
b.
Noble gas effluent nonitoring shall be provided for the total range of concentragion extending fro:a normal condition (ALARA) to a maximum of 10 pCi/cc (Xe-133).
liultiple monitors are considered to be necessary to cover the rcnges of interest. The range capacity of individual monitors should overlap by a factor of ten.
2.
Since iodine gaseous effluent nonitors for the accident condition are not considered to be practical at this time, capability for effluent nonitoring of radiciodines for the accident condition shall be provided with sampling conducted by adsorption on charcoal or other media, followed by onsite laboratory analysis.
3.
Ingcontainnent radiation level monitors with a maximum range of 10 rad /hr shall be installed.
A ainiau:a of two such monitors that are physically separated shall be provided. lionitors shall be designed and qualified to function in en accident environment."
PSC Reply:
As indicated in the PSC reply to Section 2.1.8c, a noble gas effluent nonitor providing a continuous display, recording and alann in the gontrol roo:a exists at FSV. The instrument range, however, is limited to 1 x 10 cp:a.
The FSV radio chemistry laboratory contains gamma spectroscope equipment which can be used for the analysis of ef fluent sa:1ples.
PSC will further evaluate noble gas effluent and in-containnent radiatica level menitoring and provide the NRC with a response by January 1,1980.
1281 037
Section 2.1.8c -- Improved In-Plant Iodine Instrumentation f1RC Position:
"Each licensee shall provide equipment and associated training and proce-dures for accurately determining the airborne iodine concentration in areas within the fa-ility where plant personnel nay be present during an accident."
PSC Reply:
Equipment exists at the FSV facility to discriminate noble gases fro:n iodine gases in the reactor building ventilation exhaust stack.
The vent stack airborne iudine rancentration is continuously displayed, alarmed and recorded in the control room.
Two control roo:a alana functions are provided; the first being a trouble alarm on the iodine detector to indicate loss of background signal, loss of power, or an increased level of detected radiation above back-ground but below the instrument setpoints, and the second being ghe high radia-tion alarm. The instrument range, however, is limited to 1 x 10 cpm.
For accident situations, instrugentation is provided that is capable of measuring iodine levels up to 1 x 10 mr/hr, but this instrumentation does not discriminate between iodine and noble gas.
PSC will review the equipment and associated training and procedures for detennining airborne iodine concentration at FSV during accident conditions and will provide the results of the review to the T1RC. The results of this review are tentatively scheduled for submission to the T4RC by January 1,1980.
1281 040
[
,Section 2.1.9 -- Analysis of Design and Off-Nonnal Transients and Accidents NRC Position:
" Analyses, procedures and training addressing the following are required:
1.
Small break loss-of-coolant accidents; 2.
Inadequate core cooling; and 3.
Transients and accidents.
Some analysis requit ments for small breaks have already been specified by the Bulletins and Orders Task Force.
These should be completed.
In addition, protest calculations of sone of the Loss of Fluid Test (LOFT) small break tests (scheduled to start in September,1979) shall be performed as means to verify tne analyses performed in support of the small break emergency procedures and in support of an eventual long-term verification of compliance with Apper. dix K of 10 CFR Part 50.
In the analysis of inadequate core cooling, the following conditions shall be analyzed using realistic (best-estimate) methods:
1.
Low reactor coolant system inventory (two examples will be required -
LOCA witn forced flow, LOCA without forced flow).
2.
Loss of natural circula~.on (due to loss of heat sink).
These calculations shall include the period of time during which inadequate core cooling is approached as well as the period of time during which inadequate core cooling exists.
The calculations shall be carried out in real tine far enough that all important phenomena and instrument indications are included.
Each case should then be repeated taking credit for correct operator action.
These additional cases will provide the basis for developing appropriate emergency procedures.
Taose calculationc should also provide the analytical basis for the design of any additional instrunentation needed to provide operators with an unambiguous indication of vessel water level and core cooling adequacy (see Section 2.1.3.b in this appendix).
The analyses of transients and accidents shall include the design basis events specified ja Section 15 of each FSAR. The analyses shall include a single active failure for each systen called upon to function for a particular event. Consequential failures shall a'.;o be considered.
Failures of the operators to perfora required control nanipulations shall be given consideration for pennutations of the analysas.
Operator actions th t could cause the complete loss of function of a safety system shall also be considered.
At present, these analyses need not address passive failures or multiple systet failures in the short tera.
In the recent analysis of small break LOCA's, complete loss of auxiliary feedvater was considere<. The complete loss of J
auxiliary feeduater aay be added to the failures being considered in the analysis of transients and accidents it if is concluded that nore is needed in operator training beyond the short-tera actions to upgrade auxiliary feedwater system reliabili ty.
Similarly, in the long tern, nultiple failures and passive failures uay be considered depending in part on staff review of the results of the short-term analyses.
1281 04 I
Page 2 The transient and accident analyses shall include event tree analyses, which are supplemented by computer calculations for those cases in which the system response to operator actions is unclear or these calculations could be used to provide important quantitative infonnation not available from an event tree.
For example, failure to initiate high-pressura injection could lead to core uncovery for sone transients, and a computer calculation could provide
- nformation on the amount of time available for corrective action.
Reactor simulators may provide some information in defining the event trees and would be useful in studying the information available to the operators.
The transient and accident analyses are to be perfonned for the purpose of identifying approp-riate and inappropriate operator actions relating to important safety considera-tions such as natural circulation, prevention of core uncovery, and prevention of more serious accidents.
The info'rmation derived from the preceding analyses shall be included in the plant energency procedures and operator training.
It is expected that analyses perfonned by the ilSSS vendors will be put in the fonn of emergency procedure guidelines and that the changes in the procedures uill be implemented by each licensee or applicant.
In addition to the analyses perforned by the reactor vendors, analyses of selected transicnts should be perfomed by the i4RC Office of Research, using the best available computer codes, to provide the basis for comparisons with the analytical methods being used by the reactor vendors.
These cunparisons together with comparison to data, including LOFT small break test data, will constitute the short-tem vcctification effort to assure the adequacy of the analytical methods being used to generate cuergency procedures."
pSC Response:
During the licensing review of the Fort St. Vrain reactor, careful evalua-tion was made of those sents that could result in loss of primary system coolant and the resultant ability to provide continued cooling of the reactor core.
The results of these analyses is conts.ned in FSAR Section XIV, Parts 14.4, 14.7, 14.8 and 14.11.
Also included is the analyses of the facility's ability to cope with and to provide core coaling in the event "Pennanent Loss of Forced Circulation" is experienced.
In sumary, the evaluation of the NRC Staff at the time of the licensing review identified the loss of primary coolant through the rupture of a 2" diameter line in the heliua purification system regeneration piping.
In the case of the Fort St. Vrain reactor, this is considered a "small break loss-of-coolant accident".
The only other line failure that could result in the direct loss of primary coolant would be the rupture of a primary coolant sanple line.
This line has an inner diameter of approxima tely 1/8" and its short-tena affect on reactor coolant inventory would be neglible.
A massive failure in the primary systcu with sudden release of its contents has been considered and analyzed.
The results of this analysis is contained in FSAR Section XIV, Part 14.11, Design Basis Accident No.2 " Rapid Depressurization/
Dlowdown".
1281 cm m
Page 3 In summary, the analysis indicated that the reactor core could be ade-quately cooled with the primary system at " atmospheric" pressure to remove decay and residual heat.
From the standpoint 4 inadequate core cooling, at the time of the licens-ing review, an analysis e core cooling using the PCRV Liner Cooling System to remove decay and residual heat was made.
The results of this analysis is contained in FSAR Section XIV, Part 14.10, Design Basis Accident No.1, " Perma-nent Loss of Forced Circulation (LOFC)".
This analysis also assumed the loss of one of the two PCRV Cooling Water Loops.
In sumnary, the results of this analysis indicated that the reactor core decay and residual heat could be removed, not without experiencing fuel particle damage, by the liner cooling system and that the health and safety of the plant operating staff and the public would not be jeopardized.
In conclusion, Public Service Company of Colorado considers the evaluations, analyses and conclusions documented in the Fort St. Vrain FSAR,Section XIV to meet the intent of the analysis requested by the Staff for small break loss-of-coolant accidents, inadequate core cooling, and transients and analysis.
The results of these analyses and the required operator action to cope with these eccidents is incnrporated in the facilities Operating and Emergency Fracedures and are included as a part of Operator Trainira.
1281 CC
/
' to flRC Letter Vassallo to Fuller Dated September 13, 1979 --
Instrumentation to f4onitor Containment Conditions During the Course of an Accident NRC Position:
" Consistent with satisfving the requirements set forth in General Design Criterion 13 to provide the s spibility in the control room to ascertain contain-ment conditions during the course of an accident, the following requirements shall be implemented:
1.
A continuous indication of containment pressure shall be provided in the control room.
fleasurement and indication capability shall include three times the design pressure of the containment for concrete, four times the design pressure for steel, and minus five psig for all containments.
2.
A continuous indication of hydrogen concentration in the containment atmosphere shall be provided in the control room, ficasurement shall be provided in the control room.
fleasurement capability shall be provided over the range of 0 to 10% hydrogen concentration under both positive and negative ambient pressure.
3.
A continuous indication of containment water level shall be provided in the control room for all plants.
A nar cd range instrument shall be provided for P' irs and cover the range from the bottom to the top of the containment sump. Also for PilRs, a wide range instrument shall be provided and cover the range frem the bottom of the contain-ment to tha elevation equivalent to a 500,000 gallon <ipacity.
For BilRs, a wide range instrument shall be provided and :ovar the range from the bottom to 5 feet above the normal water level of the suppres-sion pool.
The containment pressure, hydrogen concentration and wide ronge containment water level measurements shall meet the design and qualification provisions of Regulatory Guide 1.97, including qualification, redundancy, and testability.
The narrow range containment water level measurement instrumentation shall be qualified to meet the requirements of Regulatory Guide 1.89 and shall be capable of being pe,-iodically tested."
PSC Reply:
1.
Containment Pressure lionitoring The primary containment at Fort St. Vrain has three pressure monitors which indicate, record and alarm high and low pressure in the control room.
The pressure ranga of these instruments is 0-1000 psi.
There is no need for pressure instrumentation above this range due to PCRV pressure protection by two parallel relief valves set at stepped relief pressures of 793 psig and 312 psig.
As an additional nonitor, a recent modification installed two additional pressure transmitters and a digital control roon indicator.
One transmitter has a 0-100 psia range and the other has a 0-1000 psig range.
1281 04'
Page 2 The reactor confinement building has six over-pressure monitors that alam in the control room. The pressure range of these instru-ments is 0-30"H 0 positive.
The reactor builidng is normally main-9 tained at -1/4"f10 via a pressure differential indicator / controller 2
with a -2"H 0 to 0"H 0 range.
If the reactor building reaches a 7
2 positive pressure of 3"H 0, the reactor buildino louvers will open 9
and vent the building.
The reactor building ov' -pressure indicators would sense, indicate and alar.n this condition.
Based on the above, no PCRV primary coolant or reactor confine-ment building pressure indication problems will occur at Fort St.
Vrain.
xisting containment pressure instrumentation is adequate.
2.
Containment Hydrogen Monitoring All reactor coolant (helium) at Fort St. Vrain, including hydro-gen gas and other entrained gases, remains in the PCRV not only for cooling the reactor core but for processing and purification.
Therefore, hydrogen gas inside the primary containment at Fort St.
Vrain will not be detrimental to the cooling capability or safe shutdown capability of the reactor.
In the unlikely event that hydrogen gas enters the reactor confinecent building at Fort St. Vrain, the gas would be vented off by the reactor building ventilation system through exhaust filters and would not be retained in the building.
In addition, discharge of the primary system relief valves is through the PCRV relief filter and not to the reactor confinement building.
Based on tho above, continuous control room indication of primary coolant and reactor confinement building hydrogen levels at Fort St.
Vrain is not considered necessary.
3.
Containment Water Level Monitoring The nuclear reactor at Fort St. Vrain is cooled by gas and not water.
Shutdown of the reactor is accomplished by control rod inser-tion.
Emergency shutdown in the event of rod fa' lure is accomplished by pressurized shutdown hoppers that drop boron balls into the reactor.
Water is not used for shetdown or energency core spray of the reactor in a llTGR.
Venting of reactor cooling quench water and/or primary reactor coolant water to the containment sunp is not applicable for Fort St. Vrain.
Therefore, continuous indication in the control room of containment sump water level is not necessary at Fort St. Vrain.
1281 c45
. to NRC Letter Va: sallo to Fuller Dated September 13, 1979 --
_ Installation of Remotely Operated High Point Vents in the Reactor Coolant System NRC Position:
"Each applicant and licensee shall install reactor coolant system and reactor vessel head high point vents remotely operated fro:n the control room.
Since these vents fona a part of the reactor coolant pressure boundary, the design of the vents shall confonn to the requirement of Appendix A to 10 CFR Part 50 General Design Criteria.
In particular, these vents shall be safety grade, and shall satisfy the single failure criterion and the requirements of IEEE-279 in order to ensure a low probability of inadvertent actuation.
Each applicant and licensee shall provide the following information con-cerning the design and operation of these high point vents:
1.
A description of the contruction, location, size, and power supply for the vents along with results of analyses of loss-of-coolant accidents initiated by a break in the vent pipe.
The results of the analyses should be demonstrated to be acceptable in accordance with the acceptance criteria of 10 CFR 50.46.
2.
Analyses denonstrating that the direct venting of noncondensable gases with perhaps high hydrogen concentrations does not result in violation of combustible gas concentration limits m containment as described in 10 CFR Part 50.44, Regulatory Guide 1./ (Rev.1), and Standard Review Plan Section 6.2.5.
3.
Procedural guidelines for the operators' use of the vents.
The infonnation available to the operator for initiating or tenninating vent usage shall be discussed."
PSC Reply:
The nuclear reactor at Fort St. Vrain is cooled by heliun gas.
Providing reactor high point vents on a HTGR is not necessary.
Therefore, the recommended installation of reactor coolant system and reactor vessel head high point vents remotely operated fro.a the control room is not applicable to Fort St. Vrain.
1281 046
Section 2.2.la -- Shift Supervis_or's Responsibilities NRC Position:
1.
"The highest level of corporate management of each licensee shall issue and periodically reissue a management directive that emphasizes the primary management responsibility of the shift supervisor far safe operation of the plant under all conditions on his shift and that clearly establishes his comme.nd duties.
2.
Plant procedures shall be reviewed to assure that the duties, respon-sibilities, and authority of the shift supervisor and control room operators are properly defined to effect the establishment of a definite line of command and clear delineation of the command decision authority of the shift supervisor in the control room relative to other plant management personnel.
Particular emphasis shall be placed on the following:
a.
The responsibility and authority of the shift supervisor shall be to naintain the broadest perspective of operational conditions affecting the safety of the plant as a matter of highest priority at all tines when on duty in the control room.
The idea shall be reinforced that the shift supervisor should not become totally involved in any single operation in tines of emergency when multiple operations are required in the control room.
b.
The shift supervisor, until properly relieved, shall remain in the control room at all times during accident situations to direct the activities of control room operators.
Persons author-ized to relieve the shift supervisor shall be specified.
c.
If the shift supervisor is temporarily absent from the control room during routine operations, a lead control room operator shall be designated to assume the control room command function.
These temporary duties, responsibilities, and authority shall be clearly specified.
3.
Training programs for shift supervisors shall emphasize and reinforce the responsibility for safe operation and the management function the shif t supervisor is to provide for assuring safety.
4.
The administrative duties of the shift supervisor shall be reviewed by the senior officer of each utility responsible for plant operations.
Administrative functions that detract fro.a or are subordinate to the uanagement responsibility for assuring the safe operation of the plant shall be delegated to other operations personnel not on duty in the control room."
1281 ce
Page 2 PSC Reply:
By January 1,1980, the Vice President Production will issue a management directive that emphasizes the primary responsibilities of the Shift Supervisor for safe operation of the plant and that shall clearly establish the Shift Supervisors authority and responsibilities.
This management directive will be subject to annual revieu and will be revised or updated as necessary os a result of the annual review.
Existing Administrative Procedures and Policies shall be reviewed and revised as necessary to address control room operational activities during accident conditions, lines of authority and succession, temporary relief and temporary absences.
These procedures will be issued prior to January 1,1980.
1281 C40
Section 2.2.lb -- Shif t Technical Advisor NRC Position:
"Each licensee shall provide an on-shift technical advisor to the shift supervisor.
The shift technical advisor may serve more than one unit at a multi-unit site if qualified to perform the advisor function for the various units.
The shift technical advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline and have received specific training in the response and analysis of the plant for transients and accidents.
The shif t technical advisor shall also receive training in plant design and layout, including the capabilities of instrumentation and controls in the control room.
The licensee.shall assign nornal duties to the shift technical advisors that pertain to the engineering aspects of assuring safe operations of the plant, including the review and evaluation of operating experience."
PSC Reply:
With respect to the Task Force position for Shift Technical Advisors, PSC proposes to meet the January,1980 requirements as follows:
1.
Three engineers will be assigned the duties of site Technical Advisors and will be placed on, call to respond to accident conditions at the plant.
2.
The use of three engineers will ensure necessary response to the plant site, will permit on-call coverage on a rotating basis, and will ensure adequate coverage for vacations, sickness, and routine absences from the site.
3.
The Engineers will be assigned to the Technical Services Department under the supervision of the Technical Services Supervisor who reports directly to the llanager of Nuclear Production.
This organizational assignment will provide necessary independence froia plant operations and will penait a reporting authority to a high level of aanagement (See Figure 2.2.lb for overall organizational diagram).
4.
Within the Technical Services Department, the engineers will be given the responsibilities of the Shift Technical Advisor with additional responsibilites for operational and plant maintenance engineering functions which will keep them abreast of daily plant conditions, proposed modifications, modifications in progress, operational problems, procedural and licensing changes, degrading trends, equipment and syste.n problems, and overall plant status.
5.
Although the position of Shif t Technical Advisor does not require an operating license, the engineers will be required to attend initial licensing classes and will be subjected to training and internal testing to ensure their knowledge and cc:aprehension of plant operations.
The engineers will be subject to continuing training such as the operator requalification program to ensure they maintain a high level
.of proficiency concerning plant operations, emergency procedures, etc.
1281 04?
Page 2 With the exception of having these personnel on shift and the recommended 10-minute response time, we believe the above proposed methods meet the require-ments for the position of Shift Technical Advisor.
PSC does not consider it necessary to provide shif t coverage or to meet the 10-minute response time which are criteria developed based on water reactor technology.
At Fort St. Vrain accident conditions develop very slowly in comparison ~ with water reactors, thereby providing substantially. greater response times.
Fort St. Vrain has in the past experienced loss of feedwater and temporary interruption of forced reactor coaling and has recovered with no damage to fuel and/or other primary system components.
Since the reactor utilizes helium as a coolant and a fully cera:aic fuel and refractory-type core, it is not possible for a Tril-2 accident scenario to develop at Fort St. Vrain.
During the FSV licensing review, recovery of forced circulation cooling of the reactor following a 30-minute interruption in cooling from 100% reactor power was considered and analyzed and was found to result in no damage to the fuel or other primary system components.
An evaluation and analysis of a "pennanent loss of reactor forced circu-lation cooling" was also made during the licensing review and is documented in the FSV FSAR.
In sunmary, the Fort St. Vrain reactor can experience a permanent loss of forced circulation cooling without impairing the health and safety of the Plant Operating Staff or the public.
Under loss of forced circulation cooling conditions (see FSAR DBA-1, Section 14.10), we have up to five (5) hours to restore forced cooling.
Even with a permanent loss of forced cooling, the maximum off-site dose over a six (6) month duration are orders of magnitude less than 10 CFR 100 limits.
Under the maximum depressurization accident (DBA-2, see FSAR Section 14.11), a total release of the primary coolant results in off-site doses that are about one-half of 10 CFR 100 limits.
It is not possible to experience disassociation of the reactor coolant into explosive mixtures, and no change will occur in the graphite core structure even at high temperatures.
Given these inherent safety characteristics and long reactor system response tines, the role of the Shif t Technical Advisor is minimized and most certainly the need for shift coverage and immediate response is not warranted.
PSC considers a response time of two (2) hours for the Technical Advisor to be more than adequate to respond to conditions that may develop at Fort St. Vrain.
Our proposal for utilizing on-call Technical Advisors would provide for a response of less than two hours under the nost adverse conditions.
This response, along with the response of the remainder of our technical staff, considering the nature and characteristics of Fort St. Vrain, will ensure more than adequate assessment of accident situations that may develop.
He would anticipate having the three (3) Technical Advisors on-call January 1, 1980.
Recognizing that engineers with HTGR background may be difficult to obtain, these positions uay have to be filled on a temporary basis utilizing contract personnel until a permanent staff can be developed and trained.
1281 050
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)I Figure 2.2.lb
Section 2.2.lc -- Shift and Relief Turnover Procedures NRC Position:
"The licensee shall review and revise as necessary the plant procedure for shift and relief turnover to assure the following:
1.
A checklist shall be provided for the oncoming and offgoing control room operators and the onco:ning shif t supervisor to complete and sign.
The following items, as a minicum, shall be included in the checklist:
a.
Assurance that critical plant parameters are within allowable limits (parameters and allowable limits shall be listed on the checklist).
b.
Assurance of the availability and proper alignment of all systems c sential to the prevention and mitigation of operational trans-ients and accidents by a check of the control console (what to check and criteria for acceptable status shall be included on the checklist);
c.
Identification of systems and components that are in a degraded mode of operatinn permitted by the Technical Specifications.
For such systems and conponents, the length of time in the degraded mode shall be compared with the Technical Specifications action statement (this shall be recorded as a separate entry on thechecklist).
2.
Checklists or logs shall be provided for completion by the offgoing and onco;aing auxiliary operators and technicians.
Such checklists or logs shall include any equipment under raaintenance of test that by the.nselves could degrade a system critical to the prevention and mitigation of operational transients and accidents or initiate an operational transients (what to check and criteria for acceptable status shall be included on the checklist); and 3.
A system shall be established to evalucte the effectiveness of the shift and relief turnover procedure (for example, periodic indepen-dent verification of system alignments)."
PSC Reply:
Existing Adninistrative Procedures for Fort St. Vrain provide for shift turnover.
These administrative procedures will be reviewed and revised as necessary in light of the Lessons Learned Task Force guidelines.
These procc-dures nay require extensive review and evaluation in developing and preparing new checklists and logs.
It is anticipated that such reviews and evaluations can be completed to affect an inple:aentation date of March 1,1900.
1281 052
Section 2.2.2a -- Control Room Access NRC Position:
"The licensee shall make provisions for limiting access to the control room to those individuals responsible for the direct operation of the nuclear power plant (e.g., operations supervisor, shift superviosr, and control room operators), to technical advisors who may be requested or required to support the operation, and to predesignated NRC personnel.
Provisions shall include the following:
1.
Develop and implement an administrative procedure that establishes the authority and responsibility of the person in charge of the control room to limit access.
2.
Deveicp and implement procedures that establish a clear line of authority and responsibility in the control room in the event of an emergency.
The line of succession for the person in charge of the control rocm shall be established and limited to persons possessing a current senior reactor operator's license.
The plan shall clearly define the lines of connunication and authority for plant management personnel not in direct command of operations, including those who report to stations outside of the control room."
PSC Rep _ly:
An Administrr.tive Procedure will be written to formalize existing policies which allow the Shif t Supervisor, the Superintendent of Operations or plant management to rastrict access to the control room during both normal and emergency operaticrd.
Existing Emergency Procedures establish the emergency director as the Shift Supervisor and provide for the succession of the person in charge of plant operations to poness a current Senior Operators License.
FSV Emergency Procedures are currently under review and will ce revised to explicitly incor-porate control room access control and will clearly establish the lines of authority and responsibility in energency situations.
Procedures will be revised by January 1,1980.
1281 053
. Section 2.2.2b - _0nsite Technical Support Center NRC Position:
"Each operating nuclear power plant shall maintain an onsite technical support center separate from and in close proximity to the control roon that has the capability to display and transmit plant status to those individuals who are knowledgeable of and responsible for engineering and management support of reactor operations in the event of an accident.
The center shall be habitable to the same degree as the control room for postulated accident conditions.
The licensee shall revise his emergency plans as necessary to incorporate the role and location of the technical support center.
t pertain to the as-built conditions and layout of structures, Records systems and components shall be stored and filed at the site and accessible to the technical support center under emergency conditioas.
Examples of such records include systen descriptions, general crangement drawings, piping and instrument diagrams, piping system iso'letrics, electrical schematics, wire and cable lists, and single line electrical diagrams.
It is not the intent that all records described in ANSI fMS.2.9-1974 be stored and filed at the site and accessible to the technical support center under emergency conditions; however, as stated in that st6ndard, stcrage systems shall provide for accurate retrieval of all pertinent infonaation without undue delay."
PSC Re g :
1.
Interim On-Site Technical Support Center The existing Emergency Procedures establish the equivalent of the interia On-Site Technical Support Center, although by title it is called the On-Si te Command Pos t.
The interin On-Site Command Post at Fort St. Vrain is located adjacent to the control roo:a and, by the use of controlled access, pennits ready access by specifically authorized personnel to the control roon to monitor instruments and plant status during an eaergency.
The On-Site Conmand Post will house approximately eight (8) technical support personnel and will be equipped with counercial telephones to penait prisry comunications with the Control Roon as well as other c'lergency centers.
In addition, the center will have back-up ra'Jio commu-nications (battery operated).
(See Figure 2.2.2b for overall emergency center plans and conaunications for various emergency centers.) Emergency Procedures will b3 revised to designate those personnel assigned to the Technical Support Center and to define the responsibilities of those personnel during accident condi tions.
If for sNe reason the prinary area designated for the On-Site Comaand Post cannot be utilized, alternate locations will be designated.
The interin Technical Support Center (On-Site Conaand Post) will be equipped with essential drevings and procedures (i.e., P&I diagrams, FSAR, one-line electrical schematics, Technical Specifications and Emergency Procedures).
The location of the Center provides reasonable access to all drawings and procedures, providas reasonable control roca access, and providas commanication for of f-site support.
1281 05' P
Page 2 Fro;n the information gained at our Regional fleeting in Las Vegas, the above will fulfill the requirements of the interim Technical Support Center required by January 1,1980.
2.
Final on-Site Technical Support Center We are presently reviewing the requirements for the final On-Site Technical Support Center (required by January 1,1981), and we will be addressing these requirements in future correspondence.
1281 055
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Section 2.2.2c -- Onsite Operation Support Center NRC Position:
"An area to be designated as the onsite operational support center shall be established.
It shall ba separate from the control room and shall be the place to which the operations support personnel uill report in an energency situation.
Communications with the control room shall be provided.
The emer-gency plan shall be revised to reflect the existence of the center and to establish the methods and lines of communication and management."
PSC Reply:
Upon identification of an operational situation existing in the facility that may require immediate response fro:n operational and support personnel, the plant eraergency siren is sounded and, in accordance with present emergency procedures, all plant personnel, operational and support, go immediately to their " emergency stations".
In general, these " emergency stations" are those locations from which they normally work.
In-plant communications exist between the " emergency stations" and the Control Room so that in the event help is required, it is immediately available through the in-plant communications sys te.a.
All plant supervisory personnel report to the Technical Support Center to take direction fro.n the Shift Supervisors in responding to the event.
The supervisors, re charged with the responsibility of establishing and maintaining communications with the " emergency stations".
In the event plant ambient radiological conditions would require plant evacuation, the equivalent of an on-site operational Support Center has been established per the Fort St. Vrain Emergency Procedures, although by title it is called the Personnel Control Center.
The Emergency Procedures also aake provisions for alternate on-site Personnel Control Centers to accommodate direction cf releases.
Three alternate off-site locations have also been designated in the event the site should have to be evacuated.
Conmunications with the control roo:a as wel 1. as other.miergency centers can be maintaincJ by co:auercial telephone.
The Personnel Control Center Emer -
gency Kit is also equipped with a portable radio for back-up communications.
He believe we are in full compliance utih the requirements for an On-Site Operational Support Center.utQ1 pr7 ua,
RESP 0!iSE TO Ef1 CLOSURE 7 The following response is keyed to the paragraph numbers of Enclosure 7 to flRC letter Vassallo to Fuller dated September 13, 1979: 1. Upgrade Emergency Plans PSC has been working with the State of Colorado for some time now to upgrade the Radiological Emergency Response Plan. To date, the plan has been updated and reviewed by the IRAP and RAP Committees with the exception of the co:nmunications annex. The State is developing the communications annex and is procuring necessary communications equipment with a target date for conpletion in early December,1979. The Emergency Plan is presently scheduled for review by the !!RC review team in April,1980. PSC has, however, objected to this late review (see PSC letter P-79205 attached) on the basis that guidelines which have been published are not entirely applicable to Fort St. Vrain and that compliance to guidelines issued for water reactors are not warranted for Fort St. Vrain. To date, we have received no response to the attached letter. With reference to the uniform action guide levels, PSC has just received f!UREG-0610, and we are not in agreement with the proposed action guide levels. PSC intends to comment on flVREG-0610 by the requested comment date of December 1,1979. Given the position set forth by letter P-79205 attahed, the December 1,1979 comment date for NUREG-0610 and the scheduled review of April,1980 for Fort St. Vrain Emergency Plans, it is not possible to comply with the implementation schedule set forth by Enclosure 8. It is anticipated that compliance could be achieved by mid-1980 only if the problem areas identified by letter P-79205 are resolved in a timely fashion. 2. Radiation Ionitors This subject is addressed in the responses to i1RC positions P. 1.8a, 2.1.8b and 2.1.8c in this letter. The respanse of our radiation monitoring system was also addressed some tine ago by PSC letter P-79130 which has been attached for your information. 3. Emergency 0,oeration_s Center As indicated in P-79205, PSC has made plans and arrangements to utilize a facility in the City of Fort Lupton (approximately 10 miles from the site) for the Emergency Operations Center. At the regional NRC meetings, the licensees in attendence were inforned that the Energency Operations Canter should be within two (2) niles of the site. PSC is of the opinion that the Fort Lupton facility is more than adequate to meet the requinea,ents of the Emergency Operations Center, but we need a tiuely resolution to our position as expressed in P-79205 before proceeding further. 1281 050
Page 2 Our plans for the Technical Support Center have been outlined in the response to tiRC position 2.2.2b in this letter. 4. Environmental Monitoring Again, PSC has set forth its position on environmental monitoring in PSC letter P-79205. Until specific guidelines are established for Fort St. Vrain, we cannot proceed further. 5. State / Local Plans As indicated under Item 1 above, PSC has updated the Radiological Emergency Response Plan, and we are working with the State of Colorado to. ensure appropriate emergency actions. As indicated in letter P-79205, PSC is not in agreement with the extension of the EPZ to ten (10) miles although our present plans include provisions for evacuation of the affected segment out to eight (8) niles. Until we can resolve the position set forth by letter P-79205 PSC cannot proceed further. 6. Test Exercises The present draft of the Radiological Emergency Response Plan includes provisions for review of plans and test exercises that neet or exceed the flRC guidelines. 1281 057 .}}