ML20214P468
| ML20214P468 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 09/30/1986 |
| From: | Stachew J IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20214P471 | List: |
| References | |
| CON-FIN-D-6022, CON-FIN-D-6813, TASK-2.B.1, TASK-2.B.3, TASK-2.F.1, TASK-2.F.2, TASK-3.D.3.4, TASK-TM EGG-NTA-7356, GL-83-36, GL-83-37, TAC-54535, NUDOCS 8609220324 | |
| Download: ML20214P468 (25) | |
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EGG-NTA-7356 September 1986 INFORMAL REPORT e
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I f EGG-NTA-7356 i TECHNICAL EVALUATION REPORT FOR j CONFORMANCE TO NRR GENERIC LETTERS 83-36 AND 83-37 T- BY FORT ST. VRAIN NUCLEAR GENERATING STATION Docket No. 50-267 Review Performed by J. C. STACHEW Published September 1986 Idaho National Engineering Laboratory I i
- Prepared for thec.
U. S. Nuclear Regulatory Commission Washington, D.C. 20555
. Under DOE Contract DE-AC07-76ID01570 FIN No. 6813
1
. i CONTENTS ABSTRACT .............................................................. 111 ! FOREWORD .............................................................. 111
- 1. INTRODUCTION ..................................................... 1
- 2. , DISCUSSION AND EVALUATION ........................................ 2 2.1 Reactor Coolant System Vents (II.B.1) ...................... 2 2.2 Post-Accident Sampling (II.B.3) ............................ 3 2.3 Long Term Auxiliary feedwater System Evaluation (II.E.1.1) ...................................... 4 l
l 2.4 Noble Gas Effluent Monitors (II.F.1.1) ..................... 7 2.5 Sampling and Analysis of Plant Effluents (II.F.1.2) ........ 9 2.6 Containment High-Range Radiation Monitor (II.F.1.3) ........ 10 2.7 Containment Pressure Monitor (II.F.1.4) .................... 11 l 2.8 Containment Water Level Monitor (II.F.1.5) ................. 12 2.9 Containment Hydrogen Monitor (II.F.1.6) .................... 13 2.10 Instrumentation for Determination of Inadequate Core Cooling (II.F.2) ...................................... 14 2.11 Control Room Habitability Requirements (1I1.0.3.4) ......... 14
- 3.
SUMMARY
.......................................................... 18
- 4. REFERENCES ....................................................... 19
- e. .
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I f ABSTRACT This EG&G Idaho, Inc., report evaluates various submittals prov.ided by Public Service Company of Colorado for the Fort St. Vrain Nuclear Generating Station. The submittals are in response to Generic Letters No. 83-36 and 83-37, "NUREG-0737 Technical Specifications (TS)." Applicable sections of the Technical Specifications for the plants are evaluated to determine compliance to the guidelines established in the i generic letters. FOREWORD This report is supplied as part of the " Technical Assistance for Operating Reactors Licensing Actions," being conducted for the U.S. Nuclear Regulatory Commission, Washington D.C., by EG&G Idaho, Inc., NRR and I&E Support. The U.S. Nuclear Regulatory Commission funded the work under authorization B&R 20-19-10-11 1, FIN No. 06022. s.- Docket No. 50-267 TAC No. 54535
. .' F TECHNICAL EVALUATION REPORT FORT ST. VRAIN NUCLEAR GENERATING STATION
- 1. INTRODUCTION On November 1, 1983, letters were sent by the Director, Division of Licensing, to all boiling water reactor and all pressurized water reactor licensees. These Generic Letters (83-36 and 83-37) provided staff guidance on the content of the Technical Specifications associated with certain items in NUREG-0737.3 Public Service Company of Colorado submitted responses to these Generic Letters in the following The correspondences: P-84046, P-84101, P-84242, and P-85448.
following report provides the technical evaluation of the PSC submittals and makes recommendations for resolving the remaining issues. t
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- 2. DISCUSSION AND EVALUATION The licensee was requested to provide Technical Specifications for l several different systems. Each of these proposals is discussdd and evaluated in an individual subsection below:
2.1 Reactor Coolant System Vents (II.B.1) f , The Generic Letter contained the following statement.
"At least one reactor coolant system vent path (consisting of at least two valves in series which are powered from emergency buses) shall be operable and closed at all times (except for cold shutdown and refueling) at each of the following locations:
- a. Reactor Vessel Head
. b. Pressurizer Steam Space
- c. Reactor Coolant System High Point" ll l 2
~= __ _
J A typical Technical Specification for reactor coolant system vents was provided. For the plants using a power operated relief valve (PORV) as a reactor coolant system vent, the block valve was not required to be closed as long as the PORV was operable. [ valuation: This item, II.B.1, was declared to be not applicable (NA) to Fort St. Vrain in earlier Nuclear Regulatory Commission (NRC) correspondence. Therefore, no Technical Specifications are required for this ites, f j 2.2 Post-Accident Samplina (II.B.3) The Generic Letter contained the following statement.
" Licensees should ensure that their plant has the capability to obtain and analyze reactor coolant and containment atmosphere samples under accident conditions. An administrative program should be established, implemented and maintained to ensure this capability. The program should include:
4
- a. Training of personnel
- b. Procedures of sampling and ang1,ysis, and - ,
- c. Provisions for maintenance of sampling and analysis l
equipment. I "It is acceptable to the staff, if the licensee elects to reference this program in the administrative controls section of the Technical Specifications and include a detailed description of the program in the plant operations 3 l
manuals. A copy of the program should be easily available to the operating staff during accident and transient conditions." A typical Technical Specification for post-accident sampling was provided which further required the capability to sample and analyze radioactive iodines and particulates in plant gaseous effluents. Evaluation: NRC letter, dated August 28, 1985,9 stated that the review of this item should be included in the Technical Specification Upgrade Program (TSUP)astheUpgr$deProgramdraftmorecloselyfollowedtheguidance contained in the Generic Letters than the proposed application of July 31, 1984. Post-accident sampling is covered in the TSUP draft,
, dated November 30, 1985, page 6-25. Specification 6.8.4.c adequately addresses iodines and particulates in the gaseous effluents and containment sampling as well as reactor coolant sampling. Since Fort St. Vrain has no Containment in the sense of the Generic Letter, sampling of Reactor Building atmosphere is proposed in lieu of containment sampling. This is judged to be an acceptable alternative.
As a result of the review of the cited material, the licensee's response is judged to meet the requirements of the Generic Letter for l I Item II.B.3. 2.3 Lono T_erm Auxiliary Feeddater System Evaluation (II.E.1.1) The Generic Letter contains the following statement.
"The objective of this item is to improve the reliability and performance of the auxiliary feedwater (AFW) system. Technical Specifications depend on the results of the licensee's evaluation and staff review 4
l 1
of each plant. The limiting conditions of operation (LCO) and surveillance requirements for the AFW system should be similar to safety-related systems. Typical generic - Technical Specifications are provided in Enclosure 3. These
- specifications are for a plant which has three auxiliary feedwater pumps. Plant-specific Technical Specifications could be established by using the generic Technical i Specifications for the AFW system."
haluation: ! The licensee's most recent response to this item states that the Fort St. Vrain comparable system to the PWR auxiliary feedwater system is the Prestressed Concrete Reactor Vessel (PCRV) liner cooling system, addressed by Technical Specification LCO 4.2.13. Additional proposed and existing Technical Specifications were discussed in the licensee's f submittal, which relate to " concerns expressed in G-84080, March 8, 1984, relative to the operability of safe shutdown cooling equipment." Although application of criteria developed for PWRs is sometimes considerably modified when applied to FSV, sufficient justification does not appear to exist, in this instance, for designating the PCRV liner cooling system as the FSV system comparable to the PWR auxiliary feedwater system. First, the NRC has previously accepted the emergency feedwater and emergency condensate systems as satisfying the equipment requirement of Item II.E.1.1 Also, Public Service Company's (PSC's) earlier response to the NRC's request f or Technica1gSpecifications, was that they propose to use existing LCO 4.3.4, which again are specifications on the emergency feedwater and emergency condensate systems. Second, the intent of Item II.E.1.1 is to ensure operability of the cooling mode l normally relied upcn to remove heat from the primary coolant system when the main feedwater system is not available. This intent of Item II.E.1.1 is set forth in NUREG-0737 (page II.E.1.1-1, and -2 ); Standard Review Plan, Section 10.4.9, and the associated branch technical position
; ASB 10-1. The cooling mode normally relied upon at FSV on loss of main feedwater is the emergency feedwater and emergency condensate systems
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I described in FSV FSAR, Sections 10.3.6, " Loss of Main Feedwater Line or Condensate Line," and 10.3.7, " Simultaneous Loss of All Three Boiler Feed Pumps.' The PCRV liner cooling system, on the other hand, is part of the safe shutdown cooling system relied upon for a permanent loss of forced circulation (LOFC), which is the Design Basis Accident No. I described in FSAR, Section 14.10. The licensee should, therefore, retain LCO 4.3.4 as the. appropriate Technical Specifications to satisfy the Generic Letter, but these Specifications should be augmented as indicated in the NRC Letter, G-84080, as described below. The licensee, in their response, should address Technical Specifications appropriate to circulator operation on water turbine drive. These Technical Specifications may already exist and only require identification, or, to be determined by the licensee, may require additional new specifications. Loss of main feedwater at FSV would usually result in loss of circulator steam drive. Therefore, emergency feedwater or emergency condensate will only be effective in supplying cooling if the circulator water turbine drive is ensured. This added requirement for FSV has no counterpart in the PWR loss of main feedwater scenarios. In the PWR, primary coolant circulation is by electric motor driven primary i coolant pumps. As such, loss of main feedwater (and steam) does not usually effect the operability of these electric motor driven pumps. In the subject licensee's letter, reference is made to " concerns expressed in i G-84080, March 8, 1984, relative to the operability of safe shutdown cooling equipment." The NRC letter, G-84080, however, only stated that "in addition, appropriate surveillance testing requirements should be included in the FSV TS which demonstrate header ' operability including proper circulator operation when powered from each header." This previous NRC statement apparently was to clarify the additional need for circulator water turbine drive and the emergency feedwater and emergency condensate header operability. Involvement of safe shutdown cooling (reactor plant cooling system /PCRV liner cooling system) is not necessary. As pointed out above, safe shutdown cooling is required by the permanent loss of forced circulation accident scenario. Forced circulation emergency cooling using the emergency feedwater and condensate systems and circulator water turbine drive, is, on the other hand, the system relied upon for loss of main 1 6
feedwater. While the safe shutdown cooling system (FSAR Section 9.7.5) and forced circulation emergency cooling (FSAR Section 6.3) have some common components, only the latter are of concern as the comparables to the PWR auxiliary feedwater system. A majority of the Technical Specifications referenced in the subject itcensee's letter, are associated with the safe shutdown cooling /PCRV liner cooling systems. The licensee should extract only those specifications from their March 30, 1984, letter that deal with forced circulation emergency cooling, such as those on the helium. circulators, bearing water system, and bearing water accumulators. These latter specifications together with an augmented LCO 4.3.4, (see G-84080 requirements for added operability and surveillance testing for this LCO), and whatever other specifications on the forced circulation emergency cooling judged necessary by the licensee, would then constitute the required licensee response. As a result of the review of the cited material, the licensee is not in compliance with the requirement of Item II.E.1.1;:therefore, this issue is considered to be open.. 2.4 Noble Gas Effluent Monitors (II.F.1.1) The Generic Letter contained the following statement.
" Noble gas effluent monitors provide information, during and
( following an accident, which are considered helpful to the operator in assessing the plant con.dition. It is desired l ; that these monitors be operable at all times during plant operation, but they are not required for safe shutdown of the plant. In case of failure of the monitor, appropriate actions should be taken to restore its operational l , capability in a reasonable period of time. Considering the importance of the availability of the equipment and possible delays involved in administrative controls, 7 days is considered to be the apprcpriate time period to restore the operability of the monitor. An alternate method for 7
monitoring the effluent should be initiated as soon as practical, but no later than 72 hours after the identification of the failure of the monitor. If.the monitor is not restored to operable conditions within the 7 days after the failure a special report should be submitted to the NRC within 14 days following the event, outlining the l cause of inoperability, actions taken and the planned schedule for restoring the system to operable status." l A typical Technical Specification for noble gas effluent monitors was also provided which specified monitor locations and measurement ranges. Evaluation: The licensee responded in the February 9, 1984, submittal
- that existing FSV specifications ELCO 8.1.1g), 3), 4), and 8) and AC 7.4 adequately meet the intent of Generic Letter 83-37. The November 30, 1985, draft submittal, page 3/4 3-78, also references Specification 8.1.1 for requirements on gaseous effluent monitoring. NRC letter, dated January 9, 1986, accepted the design of monitors RT-7324-1 and -2 as meeting the instrument requirements for the noble gas effluent monitors.
-d Specification 8.1.1.g), 3) does not identify the mode applicability.
Although Specification 8.1.lg), 8) states that best efforts shall be exerted to return one or more failed instruments to operable status within thirty days, it does not comply with the 7 days of the Generic Letter or
-the 14 days for a Special Report. Also, the Generic Letter recommended sevenlocationsfornoblegaseffluenYmonitoringwhileRT-7324-1and-2 together only monitor one location (this is for information only as many of the locations in the Generic Letter are not applicable to FSV and also because the NRC letter accepted monitoring of only one location).
Also, the daily channel check and quarterly functional test, Specification ESR 8.1.1, do not comply with the Generic Letter requirements of at least. Once per 12 hours and at least once per 31 days, respectively.- No specific information was submitted to justify these areas of noncompliance with the Generic Letter. Also, the existing Technical Specifications ELCO 8.1.lg), 8
t and ESR 8.1.1 are not in standard format and as such are not as suited for making this change to, as would be Tables 3.3.2-1 and 4.3.2-1 in the November 30, 1985, draft. 1 1 As a result of the review of the cited material, the licensee is not l in compliance with the requirements on Item II.F.1.1; therefore, this issue is considered to be open. 2.5 Samplina and Analysis of Plant Effluents (II.F.1.2) The Generic Letter contains the following statement.
"Each operating nuclear power reactor should have the capability to collect and analyze or measure representative 1 samples of radioactive iodines and particuiates in plant gaseous effluents during and following an accident. An administrative program should be established, implemented l and maintained to ensure this capability. The program should include: >
- a. Training of personnel i
- b. Procedures for sampling and analysis, and l 'c. Provisions for maintenance of sampling and analysis
!. equipment. " 6.
"It is acceptable to the staff, if the licensee elects to reference this program in the adminis.trative controls
> section of the Technical Specifications and include a
' detailed description of the program in readily available procedures to the operating staff during accident and transient conditions."
i, 9
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Evaluation: Item II.F.1.2 was declared to require no further action in earlier NRC correspondence. Therefore, no further Technical Soecificati'ons are required for tF' item. 2.6 Containment High-Range Radiation Monitor (II.F.1.') 3 The Generic Letter contained the following statement.
"A minimum of two in-containment radiation level monitors with a maximum range of 10 rad /hr (10 R/hr for photon only) should be operable at all times except for cold shutdown and refueling outages. In case of failure of the monitor, appropriate actions should be taken to restore its operational capability as soon as possible. If the monitor is not restored to operable condition within seven days after the failure, a special report should be submitted to the NRC within 14 days following the event, outlining the cagse of inoperability, actions taken and the planned schedule for restoring the equipment to operable status.
" Typical serveillance requirements.are shown in Enclosure 3. The setpoint for the high radiation level alarm should be determined such that spuriolis alarms will be precluded. Note that the acce'p table calibration techniques for these monitors are discuss.ed in NUREG-0737."
Evaluation: The licensee responded in the July 31, 1984, submittal. However, NRC letter, dated August 28, 1985,' stated that the review of this item should be included in the Technical Specification Upgrade Program as the Upgrade Program draft more closely followed the guidance contained in the Generic Letter than the proposed application of July 31, 1984. l 10 \_ _
. H Table 3.3.2-1 and the basis, P. 3/4 3-77, of the TSUP draft identifies only one monitor RT-93250-14 with an alarm of <3.0 R/hr and an upper range limit of 10 R/hr. This is not in compliance with the requirements of two operable monitors each with an upper range limit of 10 R/hr. Also, the action statement does not require a special report to be submitted to the NRC if the monitor is not restored to operable status within seven days after fa.ilure as required by the Generic Letter and as committed to by PSC in their letter P-84101, dated March 30, 1984. However, PSC does require reactor shutdown within 36 hours if the monitor is not restored to operable status within 7 days. The monitor is located in the Reactor Building since Fort St. Vrain does not have a Containment in the sense used in the Generic Letter.
NRC letter, dated May 14, 1984, stated that PSC's single high-range radiation monitor was acceptable since the maximum credible dose in the Reactor Building was less than 2R/hr, and with sucn a low dose, the monitor is easily replaceable, making a redundant system unnecessary. As a result of the review cited above and the previous NRC acceptance of the instrumentation, the-licensee's proposed Technical Specifications in the TSUP is judged to meet the requirements of the Generic Letter. 2.7 Containment Pressure Monitor (II.F.1.4) l The Generic Letter contained the following statement.
- e. .
" Containment pressure should be continuously indicated in l
l the control room of each operating reactor during Power I l Operation, Startup and Hot Standby modes of operation. Two Channels should be operable at all times when the reactor is operating in any of the above mentioned modes. Technical l Specifications for these monitors should be included with other accident monitoring instrumentation in the present Technical Specifications. Limiting conditions for operating (including the required Actions) for the containment 11
pressure monitor should be similar to other accident monitoring instrumentation included in the present Technical Specifications." ( Lvaluation: Item II.F.1.4 was declared to require no further action in earlier NRC corr'espondence. Therefore, no further Technical Specifications are required for this item. 2.8 Containment Water Level Monitor (II.F.1.5) The Generic Letter contained the following statement.
"A continuous indication of containment water level should be provided in the control room of each reactor during Pcwer Operation, Startup and Hot Standby modes of operation. At least one channel for narrow range and two channels for wide range instruments should be operable at all times when the reactor is operating in any of the above modes.
" Narrow range instruments should cover the range from the bottom to the top of the containment sump. Wide range instruments should cover the range from the bottom of the containment to the elevation equivalent to a 600,000 gallon (or less if justified) capacity.
1 i: .
" Technical Specifications for containment water level l monitors should be included with other accident monitoring instrumentation in the present Technical Specifications.
LCOs (including the required Actions) for wide range l monitors should be similar to other accident monitoring instrumentation included in the present Technical l Specifications. LOCs for narrow range monitor should l include the requirement that the inoperable channel will be 1 12
restored to operable status within 30 days or the plant should be brought to at least a hot standby condition within the next six hours. If both monitors are inoperable, at least one monitor should be restored to operable status within 72 hours or the plant should be brought to at least hot standby condition within the next six hours." Evaluation: Item II.F.1.5 was . declared to be not applicable to Fort St. Vrain in earlier NRC correspondence.8 Therefore, no Technical Specifications are required for this item. 2.9 Containment Hydrogen Monitor (II.F.1.6) The Generic Letter contained the following statement.
"Two independent containment hydrogen monitors should be operable at all times when the reactor is operating in Power Operation or Startup modes. LCO for these monitors should include the requirement that with one hydrogen monitor inoperable, the monitor should be restored to operable status within 30 days or the plant should be brought to at least a hot standby condition within the next six hours. If both monitors are inoperable, at least one monitor should be restored to operable status within 72 hours or the plant should be brought to at least hot standby condition within the next six hours."
Evaluation: l l Item II.F.1.6 was declared to be NA to Fort St. Vrain in earlier NRC l correspondence.8 Therefore, no Technical Specifications are required for this item. l l 13
1 2.10 Instrumentation for Determinaticn of Inadeauate Core Cooling (II.F 2) Item II.F.2 was declared ,12 to be superceded by the implementation of Regulatory Guide 1.97 in response to Generic Letter 82-28, and , appropriate Technical Specifications will be required and will be tracked separately with that issue (Multi-Plant Action A-17). 2.11 Control Room Habitability Reautrements (III.D.3.4) The Generic Letter made the following statement.
" Licensees should assure that control room operators will be adequately protected against the effects of the accidental release of toxic and/or radioactive gases and that the nuclear power plant can be safely operated or shutdown under design basis accident conditions. If the results of the analyses of postulated accidental release of toxic gases (at or near the plant) indicate any need for installing the ,
toxic gas detection system, it should be included in the Technical Specifications. Typical acceptable LC0 and surveillance requirements for such a detection system (e.g. chlorine detection system) are provided in Enclosure 3. All detection systems should be included in the Technical Specifications. n.
"In addition to the above requirements, other aspects of the control room habitability requirements should be included in the Technical Specifications for the control room emergency air cleanup system. Two independent control room emergency air cleanup systems should be operable continuously during all modes of plant operation and capable of meeting design requirements. Sample Technical Specifications are provided in Enclosure 3."
14
Evaluation: NRC letter, dated August 28, 1985,' stated that the review of this item should be included in the Technical Specification Upgrade Program as the Upgrade Program draft more closely followed the guidance contained in the Generic Letter than the proposed application of July 31, 1984.6 The chlorine detection and alarm system is specified on P. 3/4 3-93 of the November 30, 1985, draft. Only one system is specified rather than the required two independent systems. NRC letter, dated September 8, 1983,I accepted the single chlorine detection and alarm system as meeting the criteria identified in Item III.D.3.4 of NUREG 0737. Therefore, this aspect of the chlorine detection and alarm system is assumed settled. Action statements, 3.3.2.6.a.1 and 3.3.2.6.a.2, require returning an inoperable system to operable status within 24 hours or for the chlorine bottle discharge valves to be closed, and a patrol made every two hours. This is not in compliance with the Generic Letter requirement to initiate the emergency ventilation system within one hour of having both chlorine detection systems inoperable (in this case only one). Also, SR 4.3.2.6.a specifies a channel check once per 24 hours versus the required once per 12 hours. The licensee in Attachment 2 to their November 27, 1985, Letter,7 states that operating the emergency ventilation system with a chlorine leak would just circulate chlorine rather than fresh air. So the emergency ventilation action was deleted and a patrol action substituted. Also, 12 hours was changed to 24 hours since ! that is the frequency generally used for chan,nel checks at Fort St. Vrain. Fort St. Vrain's general channel check frequency is inadequate as an argument here since FSV has no unique features relative to chlorine detection. FSAR Section 7.4.1, states that the preferred mode of operation during an on-site toxic gas release is the minimum makeup mode using emergency makeup flow. Also, Regulatory Guide 1.95,14 requires 4 l that immediately after control room isolation, the emergency recirculating charcoal filter or equivalent equipment designed to remove chlorine be l started up and operated. The FSV emergency ventilation system which can ! take suction from the turbine building or outside air and has i l 15
r filters F-7502, F-7503, and F -7504 appears to have been designed for and is capable of chlorine cleanup as stated in the licensee's letter of December 20, 1980. Therefore, in spite of the licensee's discussion in Attachment 2, the action statement and the surveillance frequency on channel checks are not in compliance with the guidelines. The control ro6m emergency ventilation system is specified on page 3/4 7-56 of the November 30, 1985, draft. Only one system is specified rather than the required two independent systems. Again, NRC letter, dated September 8, 1983, accepted the single control room emergency ventilation system. This aspect of the emergency ventilation system is assumed settled. Action Statement 3.7.9.a allows 7 days to restore one inoperable fan or requires shutdown in the next 24 hours. Action Statement 3.7.9.b allows 24 hours to restore one inoperable fan when control room pressure is less than 0.05 inch water gauge or requires shutdown in the next 24 hours. One fan is adequate to maintain 0.05 inch water gauge pressure (Basis, page 3/4 7-59). Sufficient air is available in the control room even if it is isolated to sustain 25 people for at least 12 hours with only 0.1% oxygen depletion (see FSAR, Section 11.2.2.6). For these reasons and as no direct comparable is made in the Generic Letter (the Action Statement there is for two independent systems), these action statements are judged to be acceptable. Action Statement 3.7.9.c would allow operation for 7 days with shutdown in the next 24 hours with the control room emergency makeup ventilation filter inoperable. Again, no direct comparable is made in-the Generic Letter. Loss of the filter defeats the purpose 'of the emergency ventilation, and is l equivalent to having no ventilation. Therefore, FSV is not in compliance l with the intent of the guidance in 'he Generic Letter which requires shutdown within 6 hours when only one of the two systems continues to be inoperable beyond 7 days. In the shutdown and refueling mode of FSV, the Action Statement allowing 7 days for restoration is, again, not in compliance with the guidance of, in this case, immediately suspending core - alterations and operations which may result in positive reactivity changes upon total loss of the control room emergency ventilation / cleanup. FSV does not have a 12-hour frequency check of the control room air temperature l l 16 I - _
7 or 31-day frequency of a 10-hour flow check and is, therefore, not in compliance with the guidelines. Testing of the filter is not in compliance j with the guidelines in the following ways: (1) specifies a one sided test flow versus a range (450 ACFM versus 450 ACFM110%); (2) lacks a penetration test after partial or3 complete charcoal adsorber replacement; (3) lacks a heater dissipation test; (4) lacks a test of automatic switchover into the recirculation mode; and (5) specifies penetration of less than 5% at 30 degrees C, 95% RH rather than 3% at 30 degrees C, 95% RH (see ANSI N509-1980). Also, the Licensee stated in the letter, dated March 30, 1984 (Ref. 5), that existing LCO 4.10.1 requires control room heating, ventilation, and air conditioning (HVAC) isolation damper operability and that the control room fans and dampers are tested according to existing TS SR 5.10.1. However, in the TSUP, Section 3.7.6.3, the existing requirement for reactor shutdown after 72 hours with inoperable dampers has been deleted without explanation. SR 5.10.1 tests the response of the control room fans and dampers to a simulated signal from the Halon Fire Suppression System.' This surveillance has no applicability to Item III.D.3.4 which is concerned with response to toxic gases (chlorine) and radiation, not fire suppression. As a result of the review of the cited material, the Licensee's response is judged to be not in compliance with the requirements of the Generic Letter, Item III.D.3.4. Y l
; e 17
- 3.
SUMMARY
The following subsections describe those issues that are, considered to have been satisfactorily addressed by the Licensee: Post-Accident Sampling (II.B.3)
. Sampling and Analysis of Plant Effluents (II.F.1.2)
Containment High-Range Radiation Monitors (II.F.1.3) ) Containment Pressure Monitor (II.F.1.4) ? l The Licensee is not in compliance with the Generic Letter guidance for the following items: Long-Term Auxiliary Feedwater System Evaluation (II.E.1.1) Noble Gas Effluent Monitor (II.F.1.1) Control Room Habitability Requirements (III.D.3.4) In previous correspondence with the Licensee, the following items have been designated as not applicable to Fort St. Vrain: Reactor Coolant System Vents (II.B.1) Containment Water Level Monitor (II.F.1.5) Containment Hydrogen Monitor (II.F.1.6) Instrumentation for determining of. inadequate core cooling, Item II.F.2, will be tracked separate 1 relative to Regulatory Guide 1.97. l t i 18
(
. 4. REFERENCES
- 1. D. G. Eisenhut, NRC letter to all Boiling Water Reactor Licensees, "NUREG-0737 Technical Specifications (Generic Letter 83-36),"
i November 1, 1983.
- 2. D. G. Eisenhut, NRC letter to all Pressurized Power Reactor Licensees, "NUREG/0737 Technical Specifications (Generic Letter 83-37),"
November 1, 1983.
- 3. NUREG-0737, Clarification of TMI Action Plan Reauirements, published by the Division of Licensing, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, November 1980.
- 4. D. W. Warembourgh letter to D. G. Eisenhut, " Generic Letters 83-37 and 83-37," February 9, 1984, Public Service Company of Colorado.
- 5. D. W. Warembourgh letter to D. G. Eisenhut, "NUREG-0737 Technical Specification Changes per Generic Letters 83-36 and 83-37," March 30, 1984, Public Service Company of Colorado.
- 6. O. R. Lee letter to E. H. Johnson, " Proposed Technical Specification Changes for NUREG-0737 Items," July 31, 1984, Public Service Company of Colorado.
- 7. O. R. Lee letter to H. N. Berkos, " Upgrade Technical Specifications,"
Novesser 27, 1985, Public Service Compa'n'y of Colorado.
- 8. R. A. Clark letter to D. Warembourgh, " Resolution of NUREG-0737 Requirement's as They Apply to Fort St. Vrain," March 24, 1982, U.S.
~
Nuclear Regulatory Commission, Enclosure 1." I 1 19
l
- 9. P. C. Wagner letter to M. Holmes, August 28, 1985, U.S. Nuclear Regulatory Commission.
- 10. K. Heitner letter to R. F. Walker, " Fort St. Vrain Post-Accident Reactor Building Effluent Activity Monitor," January 9, 1986, U.S.
Nuclear Regulatory Commission. 11.. E. H. Johnson letter to 0. R. Lee, May 14, 1984, U.S. Nuclear Regulatory Commission.
- 12. E. H. Johnson letter to 0. R. Lee, March 8, 1984, U.S. Nuclear Regulatory Commission.
- 13. G. L. Madsen letter to 0. R. Lee, September 8, 1983, U.S. Nuclear Regulatory Commission.
- 14. Regulatory Guide 1.95, Protection of Nuclear Power Plant Control Room Operators Aaainst An Accidental Chlorine Release, Revision 1, January 1977, U.S. Nuclear Regulatory Commission.
s
- 15. D. Warembourgh letter to D. G. Eisenhut, " Fort St. Vrain Unit No. 1 TMI Action Plan Requirements NUREG 0737," December 20, 1980, Public Service Company of Colorado.
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