ML20112B747
| ML20112B747 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 12/31/1984 |
| From: | Udy A EG&G, INC. |
| To: | NRC |
| Shared Package | |
| ML20100E423 | List: |
| References | |
| CON-FIN-A-6483, RTR-NUREG-0737, RTR-NUREG-737, RTR-REGGD-01.097, RTR-REGGD-1.097 TAC-51082, NUDOCS 8501100558 | |
| Download: ML20112B747 (17) | |
Text
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Enclosure CONFORMANCE TO REGULATORY GulDE 1.97 COOPER NUCLEAR STATION t
A. C. Udy I
Published December 1984 EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Comission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-761001570 FIN No. A6483 p- - -
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ABSTRACT This EG&G Idaho, Inc., report provides a review of the submittal for Regulatory Guide 1.97, Revision 2, for the Cooper Nuclear Station. Any exception to the guidelines of Regulatory Guide 1.97 are evaluated and those areas where sufficient basis for acceptability is not provided are also identified.
,. s FOREWORD This report is supplied as part of the " Program for Evaluating .
Licensee / Applicant Conformance to RG 1.97 " being conducted for the U.S.
Nuclear Regulatory Commission Office of Nuclear Reactor Regulation, Division of Systems Integration, by EG&G Idaho, Inc., NRC Licensing Support Section.
The U.S. Nuclear Regulatory Commission funded the work under authorization 87r 20-19-10-11-3.
Docket No. 50-298
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'l CONTENTS 1
A85 TRACT................................................... 11 FOREW0RD................................................... 11
- 1. INTR 000CTION.......................................... 1
- 2. REVIEW REQUIREMENTS................................... 1
- 3. EVALUATION............................................ 1 3.1 Adherence to Regulatory Guide 1.97............... 3 3.2 Type A Variables................................. 3 3.3 Exceptions to Regulatory Guide 1.97.............. 4
- 4. CONCLUSIONS........................................... 13
- 5. REFERENCES............................................ 14 i.
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CONFORMANCE TO REGULATORY GUIDE 1.97 i COOPER NUCLEAR STATION )
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- 1. INTRODUCTION On December 17, 1982 Generic Letter No. 82-33 (Reference 1) was issued j by D. G. Eisenhut. Director of the Division of licensing, Nuclear Reactor Regulation, to all licensees of operating reactors, applicants for operating licenses and holders of construction permits. This letter included additional clarification regard,ing Regulatory Guide 1.97 Revision 2 (Reference 2),
relating to the requirements for emergency response capability. These requirements have been published as Supplement 1 to NUREG-0737, "TMI Action Plan Requirements" (Reference 3). .
Nebraska Public Power District, the licensee for the Cooper Nuclear Station, provided a response to the generic letter on April 15, 1983 (Reference 4). The response to Section 6.2 of the generic letter was submitted on March 1,1984 (Reference 5), and revised on April 16, 1984 (Reference 6).
This report provides en evaluation of this material.
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! 2. REVIEW REQUIREMENTS Section 6.2 of NUREG-0737, Supplement 1, sets forth the documentation to be submitted in a report to the NRC describing how the licensee meets the guidance of Regulatory Guide 1.97 as applied to emergency response facilities. The submittsi should include documentation that provides the following information for each variable shown in the applicable table of Regulatory Guide 1.97.
- 1. Instrument range
- 2. Environmental qualification
- 3. Seismic qualification
- 4. Quality assurance
- 5. Redundance and sensor location ,
- 6. Power supply
- 7. Location cf display
- 8. Schedule of installation or upgrade.
Further, the submittal should identify deviations from the guidance in the regulatory guide and provide supporting justification or alternatives.
Subsequent to the issuance of the generic letter, the NRC held regional meetings in February and March 1983, to answer licensee and applicant questions and concerns regarding the NRC policy on this matter. At these meetings, it was noted that the NRC review would only address exceptions taken to the guidance of Regulatory Guide 1.97. Further, where licensees or applicants explicitly state that instrument systems conform to the provisions of the guide it was noted that no further staff review would be necessary.
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Therefore, this report only addresses exceptions to the guidance of Regulatory Guide 1.97. The following evaluation is an audit of the ?icensee's submittals based on the review policy described in the NRC regional meetings.
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- 3. EVALUATION The licensee provided a response to Item 6.2 of the generic letter on March 1, 1984. This was revised on April 16, 1984. The response describes the licensee's position on post-accident monitoring instrumentation. This i
evaluation is based on the material.
3.1 Adherence to Regulatory Guide 1.97 The licensee has provided a review of their post-accident monitoring instrumentation that compares the instrumentation characteristics against the recomendations of Regulatory Guide 1.97, Revision 2. ,
Reference 5 provides the licensee's status and schedule for implementation of Regulatory Guide 1.97 requirements. Reference 6 provides justification for those deviations from the regulatory guide requirements that are identified by the licensee.
Therefore, it is concluded that the licensee has provided an explicit commitment on conformance to the guidance of Regulatory Guide 1.97, except for those deviations that were justified by the licensee as noted in Section 3.3.
3.2 Type A Variables Regulatory Guide 1.97 does not specifically identify Type A variables,
. i.e., those variables that provide information required to permit the control
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r operator to take specific manually-controlled safety actions. The licensee classified the following instrumentation channels as Type A variables.
- 1. Reactor pressure vessel (RPV) level
- 2. RPV water pressure 4
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- 3. Drywell pressure l
- 4. Suppression pool water level
- 5. Suppression pool water temperature.
All of the above variables are also included as Type B or D Variables.
3.3 Exceptions to Regulatory Guide 1.97 The licensee identified the following deviations from the recommendations ;
3.3.1 Neutron Flux
' Regulatory Guide 1.97 re6ommends Category 1 instrumentation for this variable with a rar.ge of 10-6 to 125 percent of full power. The licensee has supplied instrumentation for this variable with a range of 10-5 to 125 percent of full power, saying it will be implemented as Category 3. Thus the lower limit of the recommended range and the category of the instrumentation supplied deviate from what is recommended.
The licensee notes that accident scenarios resulting in an increase in reactivity could only be caused by inadvertent removal of boron that was added by the standby liquid control system or by other effects such as a change in
[ temperature or fission product poisoning. Since these reactivity additions would likely have a slow rate of change, the licensee concludes that power readings in the range of 10-5 percent of full power would give the operator ~
sufficient time (21.6 minutes) to identify the problem and take corrective action before the power reached 0.5 percent of full power. The licensee has not shown that this time is within the time permitted by Standard Review Plan 15.4.6 (i.e., 15 minutes froa alarm to loss of shutdown margin). We find that the justification provided by the licensee for a deviation in the range recommendations for neutron flu ~x is incomplete, and therefore the deviation is not acceptable.
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- l The licensee has proposed some upgrading of the neutron flux instrumentation; however, they have not identified which of four proposed options they will implement.
In the process of our review of neutrnn flux instrumentation, we note that the mechanical drives of the detectors have not satisfied the environmental qualification requirement cf Regu'atory Guide 1.97. This deviation is similar to most BWRs. A Category 1 system that meets all the
- criteria of Regulatory Guide 1.97 is an industry development item. Based on our review, we conclude that the existing instrumentation is acceptable for interim operation. The licensee should follow industry development of this equipment, evaluate , newly developed equipment, and install Category 1 instrumentation when it becomes available.
3.3.2 Coolant Level in Reactor .
The licensee indicates that there will be one redundant system of compensated water level channels installed during the 1986 refueling outage.
These are said to have a range of 0 to 60 in. Another single channel, with a range from 0 to 69.5 in. will be complete by March 31, 1985. Assuming the instrument zero is the same, less than six feet of reactor vessel level will be measured.
Our examination of the Final Safety Analysis Report for the Cooper Nuclear Station (FSAR, Reference 7), shows that 2/3 of the core height is at
[ 305 in., the normal water level is at 557 in. and the upper instrumentation nozzles (N12 A & B) are at 586 in. Thus, we conclude that the range supplied does not comply with the recommended range. The licensee has not given any justification in support of this deviation. We conclude that the licensee should supply redundant Category 1 instrumentation for this variable with a
, total range as recomended by Regulatory Guide 1.97, i.e., from the bottom of the core support plate to the lesser of the top of the vessel or the centerline of the main steamline.
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3.3.3 Drywell Sump Level, Drywell Drains Sump Level Regulatory Guide 1.97 recommends Category 1 instrumentation for these variables. The licensee proposes to install Category 3 instrumentation for these variables during the 1986 refueling outage. No safety related system is actuated either automatically or manually as a result of the sump level. The drywell sump systems are automatically isolated at the primary containment penetration should an accident signal occur.
For small leaks, this Category 3 instrumentation will continue to function as the drywell temperature and pressure will not have changed significantly. TheAtfore,thesumplevelscanbeusedasaleadingindicator of reactor coolant system leakage. For larger leaks, the sumps will fill promptly, negating this information because the sumps isolate due to the increase in drywell pressure caused by the accident. The sumps can be assumed full with Category 3 instruments once containment isolation occurs at 2 psig.
In either case, we find the Category 3 instruments provided for this variable acceptable.
3.3.4 Radiation Level in Circulating Primary Coolant The license'e states that radiation level measurements to indicate fuel
_ cladding failure are provided by the following instruments:
- 1. Post-accident sampling system
- 2. Condenser off-gas re.diation monitors
- 3. Main steamline radiation monitors 7
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- 4. Primary containment radiation monitors
- 5. Containment hydrogen concentration monitors Based on the justification provided by the licensee, we conclude that the instrumentation supplied for this variable is adequate, and therefore, acceptable.
3.3.5 Containmont and Drywell Hydrogen Concentration Regulatory Guide 1.97 recoseends redundant Category 1 instrumentation for this variable with a range of 0 to 30 percent. The licensee does have instrumentation for this variable. There is instrumentation with three separate capabilities, however, there is no redundance for a given range.
Ranges available are O to 5 percent O to 10 percent and 0 to 20 percent. The ,
power sources are not identified.
We conclude that the licensee should identify the specific deviations taken from Regulatory Guide .l'97 and justify those deviations.
3.3.6 Containment and Drywell Oxygen ~ Concentration t
Regulatory Guide 1.97 recommends redundant Category 1 instrumentation for this variable with a range of 0 to 10 percent. The licensee has this information for the inverted containment, with three separate ranges.
However, there is no redundance for a given range. The ranges available are O to 5 percent, O to 10 percent and 0 to 25 percent. The instrumentation with the range of 0 to 25 percent does not have appropriate seismic qualification; the other instruments do. Thus, there is qualified instrumentation with redundancy up to 5 percent. Also, there is redundancy up to 10 percent but one of the instruments is not qualified.
Based on the above, we find that the existing instrumentation is acceptable.
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3.3.7 Radiation Exposure Rate
- l Regulatory Guide 1.97, Revision 2, recommends Category 2 instrumentation for this variable with a range from 10-1 to 104 R/hr. The licensee has provided Category 3 radiation exposure rate monitors (rather than Category 2) that have ranges that are lower than recommended by Regulatory Guide 1.97.
These are stated as being influenced by piped radioactive fluids. The
- licensee concludes that this makes it impractical to detect primary j containment breach by use of these monitors, and that Category 3 instrumentation is suitable fer this application.
The licensee states that the plant noble gas effluent monitors are adequate to monitor the effluent from the secondary containment. The licensee determines the habitability of secondary containment by a combination of atmosphere sampling and portable radiation survey instruments, not fixed .
location radiation exposure rate meters.
Regulatory Guide 1.97, Revision 3 (Reference 8), changes this variable to Category 3. Therefore, the only deviation of the Cooper stat!on for this variable is the range supplied for a given location. The licensee has not shown any analysis of radiation levels expected for the monitor location.
4 The licensee should show that the existing radiation exposure rate monitors have ranges that encompass the expected radiation levels in their
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locations.
3.3.8 Suppression Chamber Spray Flow Drywell Spray Flow Regulatory Guide 1.97 recomends instrumentation for these variables with a range of 0 to 110 percent of design flow. Sections 4.3.5 and 8.5 of the FSAR show that the existence of spray flow to the suppression pool and drywell can be established. This is done by use of the residual heat removal system flow in conjunction with valve lineup. These parameters are indicated in the control room. The licensee indicates that the effectiveness of these flows is i
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indicated by pressure and temperature changes in the drywell and the suppression chamber. We find that this instrumentation is adequate for this variable.
3.3.9 Drywell Atmosphere Temperature Regulatory Guide 1.97 reconnends instrumentation for this variable with a range from 40 to 440*F. The licensee has provided instrumentation for this variable with a range of 50 to 350*F. The deviation is not measuring frca 40 to 50*F and from 350 to 440*F. This deviation has not been justified by the licensee.
i The licensee sh'ould either justify the deviation from the recommended range or re-span the instrumentation to coincide with the range recommended by Regulatory Guide 1.97, 3.3.10 Standby Liquid Control System Flow Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The licensee, not presently having instrumentation for this variable, has committed to install Category 3 instrumentation for this '
variable during the 1986 refueling outage. They acknowledge that the " final ATWS (anticipated transient without scram) rule may require re-evaluation."
In justifying this deviation the following is noted. The standby liquid control system storage tank level gives indication that flow is occurring.
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Yhe reactivity change in the reactor as measured by neutron flux is an l
[ indication of flow. The pump motor indicating lights show system operation.
The squib valve continuity indicating lights are an indication of flow.
l We find that the proposed Category 3 indication of flow, in combination with other diverse indication means, is acceptable for this variable.
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3.3.11 Standby Liquid Control System Storage Tank Level ,
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Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The licensee has instrumentation that except for environmental qualification, is Category 2. They acknowledge that the " final ATWS rule may require re-evaluation."
The licensee justifies using this instrumentation based on the mild environment in which the standby liquid control system must operate. We find this acceptable.
3.3.12 Reactor Building or Secondary Containment Area Radiation Regulatory Guide 1.97 recomends Category 2 instrumentation for this variable with a range of 10-1 to 104 R/hr for the Mark I containment. The '
licensee has some instruments with a range of 10-5 to 10-1 R/hr, and a channel with a range 1 of 10-1 to 103 R/hr. All these instruments are Category 3 rather than the recomended Category 2.
The licensee reports that the use of local radiation exposure rate monitors to detect breach or leakage through primary containment penetrations results in ambiguous indications. This is due to the radioactivity in the primary containment, the radioactivity in the fluids flowing in the emergency core coolant system piping and the amount and location of fluid and electrical
] penetrations. The licensee concludes that the use of the plant noble gas effluent monitors is the proper way to accomplish the purpose of this variable. Therefore, the licensee concludes that the existing Category 3 instrumentation for this variable is adequate.
Based on the above, we concur that the existing Category 3
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instrumentation and ranges are adequate.
3.3.13 Plant and Environs Radiation Regulatory Guide 1.97 recomends instrumentation for this variable with ranges of 10-3 to 104R/hr-photons and 10-3 to 10 4rads /hr beta and low energy 11
photons. The licensee has instrumentation for this variable with ranges of 10-3 to 103 R/hr and 10-3 to 200 rads /hr. The licensee states that the
" existing equipment range is adequate," but does not state why it is adequate.
The licensee should provide justification for the deviation in range.
3.3.14 Accident Sampling (Primary Coolant. Containment Air and Sump)
Regulatory Guide 1.97 recomends sampling and onsite analysis capability for the reactor coolant system, containment sump. ECCS pump room sumps and other similar auxiliary building sump liquids and containment air. The s licensee's post-accident sampling and analysis as recomended by the regulatory guide, except for the following deviations.
! 1. Chloride content--the minimum observable concentration is 10 ppb; the' maximum range is 10 ppm rather than 20 ppm.
- 2. Oxygen content--the minimum observable concentration is 10 ppb; the maximum range is 1 ppm rather than 20 ppe.
- 3. Dissolved hydrogen or total gas--the licensee calculates this, however, they do not state how it is done, nor do they equate it to the recommended analysis range of 0 to 2000 cc (STP)/kg.
- 4. The licensee does not sample the sumps recommended by the regulatory guide.
- 5. The licensee has not provided information (required by Section 6.2 of Reference 3) to show compliance to the regulatory guide for hydrogen and oxygen content of the containment air.
The licensee takes exception to the guidance of Regulatory Guide 1.97 with respect to post-accident sampling capability. This exception goes beyond the scope of the review and is 15eing addressed by the NRC as part of their review of NUREG-0737. Item II.B.3.
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- 4. CONCLUSIONS Based on our review, we find that the licensee either conforms to, or is justified in deviating from, the guidance of Regulatory Guide 1.97 with the following exceptions:
- 1. Neutron flux--the licensee's present instrumentation is acceptable on an interim basis until Category 1 instrumentation is developed and installed (Section 3.3.1).
- 2. Coolant level in rector--the licensee should supply instrumentation for this v,ariable with the range as reconmended in the regulatory guide (Sec' tion 3.3.2).
'3. Containment and drywill hydrogen concentration--the licensee should
- identify the specific deviations taken and provide justification for thesedeviations(Section3.3.5).
- 4. Radiation exposure rate--the licensee should show that the ranges supplisd for this variable encompass the radiation level at the instrument location (Section 3.3.7).
- 5. Drywell atmosphere temperature--the licensee should justify the deviations from the reconnended range or supply the recommended
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range (Section 3.3.9).
- 6. Plant and environs radiation--the licensee should provide the basis for accepting the existing instrumentation ranges (Section 3.3.13).
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- 5. REFERENCES
- 1. NRC letter D. G. Eisenhut to All Licensess of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Permits,
" Supplement No. I to NUREG-0737--Requirements for Emergency Response Capability (Generic Letter No. 82-33)," December 17, 1982.
- 2. Instrumentation for Light-Water-Cooled Nuclear Power Plants to Asses 2 Plant and Environs Conditions During and Following an Accident, Regulatory Guide 1.97, Revision 2. U.S. Nuclear Regulatory Commission (NRC),OfficeofStandardsDevelopment, December 1980.
- 3. Clarification of TMI Action Plan Requirements. Requirements for Emergency Response Capability, NUREG-0737 Supplement No. 1, NRC, Office of Nuclear Reactor Regulation, January 1983.
- 4. Nebraska Public Power District letter, J. M. Pilant to D. G. Eisenhut, NRC, " Response to NUREG-0737, Supplement 1," April 15, 1983 LQA8300129.
- 5. Nebraska Public Power District letter, J. M. P11 ant to D. G. Eisenhut, NRC, "NUREG-0737. Supplement 1 - Regulatory Guida 1.97 " March 1, 1984, .
.NLS8400073. --
I 6. Nebraska Public Power District letter, J. M. Pilant to D. G. eisenhut, I
NRC, "NUREG-0737, Supplement 1 - Reonlatory Guide 1.97 " April 16, 1984.
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- 7. ' Safety Analysis Report, Cooper Nuclear Station," Nebraska Public power
- District, 1971.
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- 8. Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Environs Conditions During and Following an Accident, Regulatory Guide 1.97, Revision 3, NRC, Office of Nuclear Regulatory Research,
- May 1983.
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