ML20215J143
| ML20215J143 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 02/28/1987 |
| From: | Roberts E EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20215J129 | List: |
| References | |
| CON-FIN-A-6483, RTR-REGGD-01.097, RTR-REGGD-1.097 EGG-NTA-7079, EGG-NTA-7079-01, EGG-NTA-7079-1, NUDOCS 8705070242 | |
| Download: ML20215J143 (26) | |
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EGG-NTA-7079 February 1987 INFORMAL REPORT s
idaho CONFORMANCE TO REGULATORY GUIDE 1.97, THREE MILE ISLAND NUCLEAR STATION, UNIT N0. 1 National Engineering Laboratory Managed J. W. Stoffel by the U.S.
Department ofEnergy l
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Prepared for the
,,, ,,, _ ,, U. S. NUCLEAR REGULATORY COMMISSION DOE Contract No. DE-AC07-761D01570 8705070242 870219 PDR ADOCK 05000289 SP PDR
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DISCLAIMER This book was prepared as an account of work sponsored by an egency of the United States Government. Neither the United States Government nor any agency thereof, not any of their employees, makes any warranty, express or imphed, or assumes any legal liebslity or responsibility for the acCVacy Completeness, or usefulness of any informa'. son, apparatus, product or process disclosed, or represents that its use would not infringe privately owned nghts. References herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessanly constitute or imply its endorsement, recommendation, or favoring by the Uruted States Government or any agency thereof. The views and opinions of authors expressed herein do not necessanly state or reflect these of the United States Government or any agency thereof.
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l EGG-NTA-7079 !
- TECHNICAL EVALUATION REPORT l CONFORMANCE TO REGULATORY GUIDE 1.97 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1 Docket No. 50-289 ,
l J. W. Stoffel l
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Published February 1987 i l
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f Idaho National Engineering Laboratory
, EG&G Idaho, Inc.
1 Prepared for the
! U.S. Nuclear Regulatory Commission I
Washington, D.C. 20555 i
Under DOE Contract No. DE-AC07-76ID01570 4 FIN No. A6483 )
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ABSTRACT This EG&G Idaho, Inc., report reviews the submittal for Regulatory Guide 1.97, Revision 3, for Unit No. 1 of the Three Mile Island Nuclear Station and identifies areas of nonconformance to the regulatory guide.
Exceptions to Regulatory Guide 1.97 are evaluated and those areas where
- sufficient basis for acceptability is not provided are identified.
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- Docket No. 50-289 i TAC No. 51361 11
t 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 Comission, Office of Nuclear Reactor- Regulation,
. Division of PWR Licensing-A, by EG&G Idaho, Inc., NRR and I&E Support Branch.
The U.S. Nuclear Regulatory Comission funded the work under authorization B&R 20-19-10-11-3.
Docket No. 50-289 TAC No. 51361 1
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CONTENTS ABSTRACT .............................................................. 11 FOREWORD .............................................................. 111
- 1. INTRODUCTION ..................................................... 1
- 2. REVIEW REQUIREMENTS .............................................. 2
- 3. EVALUATION ....................................................... 4 3.1 Adherence to Regulatory Guide 1.97 ......................... 4 3.2 Type A Variables ........................................... 4 3.3 Exceptions to Regulatory Guide 1.97 ........................ 5
- 4. CONCLUSIONS ...................................................... 17
- 5. REFERENCES ....................................................... 18 e
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CONFORMANCE TO REGULATORY GUIDE 1.97 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1
- 1. INTRODUCTION On December 17, 1982, Generic Letter No. 82-33 (Reference 1) was c issued 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 regarding Regulatory Guide 1.97, Revision 2 (Reference 2), relating to the requirements for emergency response capability. These requirements have been published as Supplement No. 1 to NUREG-0737, "TMI Action Plan Requirements" (Reference 3).
GPU Nuclear Corporation, the licensee for the Three Mile Island Nuclear Station, provided a response to Section 6.2 of the ger.eric letter on October 1, 1984 (Reference 4). Additional information was submitted on June 5, 1986 (Reference 5). These responses provide a comparison of the licensee's instrumentation to the recommendations of Revision 3 of Regulatory Guide 1.97 (Reference 6).
This report provides an evaluation of that material.
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- 2. REVIEW REQUIREMENTS Section 6.2 of NUREG-0737. Supplement No. 1, sets forth the documentation to be submitted in a report to the NRC describing how the licensee complies with Regulatory Guide 1.97 as applied to emergency response fact 11 ties. The submittal 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 1
- 7. Location of display
- 8. Schedule of installation or upgrade The submittal should identify any deviations from the recommendations of Regulatory Guide 1.97 and provide supporting justification or alternatives for the deviations identified.
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 subject.
, At these meetings, it was noted that the NRC review would only address !
exceptions taken to Regulatory Guide 1.97. Where licensees or applicants explicitly state that instrument systems conform to the regulatory guide, i
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1 it was noted that no further staff review would be necessary.. Therefore, j I
this report only addresses exceptions to Regulatory _ Guide 1.97. The
.following evaluation is an audit of the licensee'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 NRC Generic Letter 82-33 on October 1, 1984. Additional information was submitted on June 5, 1986. The responses describe the licensee's position on post-accident monitoring instrumentation. This evaluation is based on that ,
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 recommendations of Regulatory Guide 1.97, Revision 3. The review lists the regulatory guide variables, showing either full compliance, noncompliance with justification, or noncompliance with a commitment and schedule to upgrade. The licensee states that all upgrade modifications are scheduled for completion by the second refueling outage after restart, designated refueling outage 7R. Therefore, we conclude that the licensee has provided an explicit commitment on conformance to Regulatory
< Guide 1.97. Exceptions to and deviations from the regulatory guide are noted in Section 3.3.
3.2 Tvoe A Variables Regulatory Guide 1.97 does not specifically identify Type A variables, i.e., those variables that provide the information required to permit the control room operator to take specific manually controlled safety actions.
The licensee classifies the following instrumentation as Type A.
- 1. Reactor coolant system (RCS) cold leg water temperature i
- 2. RCS pressure
- 3. Core exit temperature 4
- 4. . Degrees of subcooling
- 5. Containment hydrogen concentration
- 6. Low pressure injection / decay heat removal system flow
- 7. Flow in high pressure injection system
- 8. Refueling water storage tank level
- 9. Steam generator level j 10. Steam generator pressure
- 11. Auxiliary or emergency feedwater flow
- 12. Condensate storage tank water level I
This instrumentation meets the Category 1 recommendations consistent with the requirements for Type A variables, except as noted in Section 3.3.
3.3 Exceptions to Reaulatory Guide 1.97 The licensee identified deviations and exceptions from Regulatory Guide 1.97. These are discussed in the following paragraphs.
! 3.3.1 Reactor Coolant System (RCS) Soluble Boron Concentration Regulatory Guide 1.97 recommends on-line instrumentation with a range of 0 to 6000 ppm. The licensee has not provided this on-line
. instrumentation, but can obtain the information by utilizing the post-accident sampling system and on-site laboratory analysis.
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The licensee deviates from Regulatory Guide 1.97 with respect to post-accident sampling capability. This deviation goes beyond the scope of this review and is being addressed by the NRC as part of their review of NUREG-0737, Item 11.8.3.
3.3.2 RCS Cold Leo Water Temperature
- Regulatory Guide 1.97 recommends instrumentation with a range of 50 to 700*F for this variable. The licensee has supplied instrumentation with a I range of 50 to 650*F. The licensee considers the existing range adequate based on the maximum steam generator pressure of 1200 psig and a
corresponding saturation temperature of 600*F. Therefore, the cold leg-i temperature would always be at or below this value.
i Based on the licensee's statement that the instrumentation will remain on scale for any anticipated event, we find the range of this instrumentation acceptable.
3.3.3 RCS Hot Leo Water Temperature l
i Regulatory Guide 1.97 recommends instrumentation with a range of 0 to 700*F for this variable. The licensee has supplied instrumentation with a
- range of 120 to 920*F. The licensee states that at temperatures less than l
300*F, the plant will be in the decay heat removal mode, in cold shutdown, and this temperature is not then required. The decay heat removal system
! has additional temperature instrumentation to monitor the RCS in this temperature range. Category 1 core exit thermocouples also provide
] information below 120*F.
Based on the alternate instrumentation and the justification provided i by the licensee, we conclude that the instrumentation supplied for this
- variable is adequate and, therefore, acceptable. ,
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3.3.4 RCS Pressure Regulatory Guide 1.97 recommends instrumentation for this variable with a range from 0 to 3000 psig. In Reference 4, the licensee stated that instrumentation with a range from 0 to 2500 psig is provided for this variable. The licensee stated that no additional operator action would be taken or performed with an extended range from 2500 to 3000 psig, and that
. the code safety valves on the pressurizer are set to relieve pressure at 2500 psig.
In Reference 5, the licensee states that they will be in compliance with the regulatory guide requirement by refueling outage 7R. We find this commitment acceptable.
3.3.5 Radiation Level in Circulatina Primary Coolant The licensee indicates that radiation level measurements to indicate fuel cladding failure are provided by the following instruments:
- 1. Letdown line radiation monitors (during normal operation)
- 2. Post-accident sampling system.
The post-accident sampling system is available with the reactor isolated, and is being reviewed by the NRC as part of their review of NUREG-0737, Item II.B.3.
Based on the alternate instrumentation provided by the licensee, we conclude that the instrumentation supplied for this variable is adequate and, therefore, acceptable.
3.3.6 RHR Heat Exchanaer Outlet Temperature Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable with a range of 40*F to 350*F. As such, environmentally qualified 7
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instrumentation is required in accordance with 10 CFR 50.49. In j Reference 4, the licensee states that the existing range of 0 to 300*f is sufficient to cover all post-accident conditions since decay heat removal operation is initiated when the RCS temperature is below 300*F.
Based on the licensee's justification, we find this range adequate to
- l monitor this variable during all accident and post-accident conditions.
l however, the licensee did not provide justification for the environmental qualification deviation in Reference 4.
i In Reference 5, the licensee states that this instrumentation has been 1 incorporated on the TMI-1 environmental qualification master list. We find this acceptable.
4 3.3.7 Accumulator Tank level and Pressure Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The licensee has provided Category 3 instrumentation that, except for environmental qualification, is Category 2. The licensee justifies this deviation in Reference 4 by stating that these instruments
- provide the operator information pertaining to tank status during normal j operation, and that since the core flooding system is totally passive, no j monitoring of these parameters is required for any manual actions to
- mitigate the consequences of an accident. Reference 5 restated the l licensee's position that Category 3 instrumentation is adequate to monitor
- this variable.
1 The existing instrumentation is not acceptable, An environmentally 1
qualified instrument is necessary to monitor the status of these tanks.
- The licensee should designate either level or pressure as the key variable ,
to determine accumulator discharge and provide instrumentation for that variable that meets the requirements of Regulatory Guide 1.97 and ,
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1 3.3.8 Accumulator Tank Isolation Valve Position Regulatory Guide 1.97 recomends Category 2 instrumentation for this i variable. The licensee states that these are motor operated valves. They are open for reactor operation. The circuit breakers for these valves are open and de-energized when the reactor is critical. Therefore, the licensee recommends that this variable be reclassified as Category 3.
Based on the licensee's justification and the fact that these valves j are open and do not change position during or following an accident, we
! consider Category 3 instrumentation adequate for this variable.
I 3.3.9 Boric Acid Charaina Flow i
4 I The licensee does not have instrumentation for this variable. The licensee states that the charging system is not part of the emergency core cooling system (ECCS). High pressure injection and low pressure injection l
l are the flow paths of the ECCS that are monitored. Therefore, we find that this variable is not applicable at the Three Mile Island Station.
2 3.3.10 Pressurizer Level i
Regulatory Guide 1.97 recomends Category 1 instrumentation for this variable. The licensee considers pressurizer level instrumentation to be
- Category 2 and has not met the environmental qualification requirement for the temperature compensation elements.
1 The justification provided by the licensee for the Category I deviation is that pressurizer level is only used as an indicator to the 4
operator that throttling of the high pressure injection flow is allowed. !
. Therefore, the licensee's position is that the pressurizer level is an
! indication of system operating status but is not a key variable. In Reference 5, the licensee submitted a table showing the effects of the' loss
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of a temperature compensation element on the indicated pressurizer level.
This table indicates that if an element is shorted at a pressurizer temperature greater than 100*F the pressurizer could be water solid, and indicate on scale. If an element fa'ls open, at low pressurizer levels, level will indicate off-scale low. If an element fails open at high pressurizer levels, indication will be off-scale high before the ,
pressurizer is actually water solid. ,
It appears that the licensee is addressing their justification for this deviation on the need for pressurizer level instrumentation for core cooling only. However, the purpose of this instrumentation, as stated in Regulatory Guide 1.97, is to ensure proper operation of the pressurizer.
Pressurizer level is a key variable used to ensure proper operation of the pressurizer. The licensee has not provided sufficient justification for deviating from the regulatory guide requirements for this variable. The l licensee should commit to installing Category 1 Temperature Compensation elements for this variable.
3.3.11 Pressurizer Heater Status Regulatory Guide 1.97 recommends instrumentation to monitor the current drawn by the pressurizer heaters. The licensee's instrumentation consists of on/off indication of the pressurizer heaters. The licensee considers this to be sufficient indication when used in conjunction with RCS pressure.
Section II.E.3.1 of NUREG-0737 requires a number of the pressurizer heaters to have the capability of being powered by the emergency power sources. Instrumentation is to be provided to prevent overloading a diesel generator.
In Reference 5, the Itcensee has maintained the position that an on-off mode of indication is adequate to monitor this variable. The licensee states that the most direct and effective measure of heater performance is the response of reactor coolant pressure. The 10 l
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licensee further states that the diesel current can be monitored with the
- diesel ammeters which enables the operator to determine (based on the known power consumption of the heaters) whether he can load the heaters without
{ overloading the diesels.
- We find the justification provided by the licensee unacceptable. The i, light indicating the pressurizer heater circuit breaker is closed does not
! , indicate that the heaters are in fact energized or what amount of heaters are working. A means of monitoring pressurizer heater current in the control room should be provided.
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i 3.1.12 Ouench Tank Temperature ,
Regulatory Guide 1.97 recommends instrumentation for this variable j with a range from 50 to 750*F. The installed instrumentation has a range l of 0 to 275'F. The licensee states that the tank is isolated with a j reactor trip and that the existing temperature range is adequate to detect leakage into the tank.
l In Reference 5, the licensee states that a relief valve set at 40 psig
! (saturation temperature 287'F) and a rupture disc set at 55 psig
! (saturation temperature 308'F) are installed on the tank. This means that j for a short period of time the temperature of the tank could be above the i existing range. The licensee has committed to provide the capability to
! monitor the complete postulated temperature range by refueling outage 7R.
j We find this connitment acceptable.
l l 3.3.13 Safetv/ Relief Valve Positions or Main Steam Flow i 4
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- Regulatory Guide 1.97 recommends Category 2 instrumentation for this l . variable. The licensee has provided Category 3 instrumentation. The licensee states that Category 3 instrumentation is acceptable for this i
l* variable because they consider the key variables to determine the steam j generator (SG) safety / relief valve position or main steam flow to be SG 1evel and SG pressure. Valve position indication is provided as backup instrumentation.
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4' The licensee considers the valve position indication to be a backup for the Category 1 steam generator level and pressure instrumentation. As f'
the regulatory guide allows backup instrumentation to be Category 3, we
. find this deviation acceptable.
l 3.3.14 Containment Soray Flow
- Regulatory Guide 1.97 recommends Category 2 instrumentation for this .
variable. The licensee has provided instrumentation that, except for environmental qualification, is Category 2. The licensee did not submit 1 justification for the environmental qualification deviation in Reference 4.
l In Reference 5, the licensee states that this instrumentation has been incorporated in the TMI-1 environmental qualification master list. We find l this acceptable.
1 3.3.15 Containment Atmosphere Temocrature
! Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable with a range from 40 to 400*F. The licensee has supplied
. Category 3 instrumentation with a range of 0 to 300*f. Their justification 1
{ for this deviation is that the primary variable required to show accident ;
j mitigation and containment integrity is reactor building pressure, a l Category I variable. The Itcensee considers the containment atmosphere
! temperature to be a Category 3 variable.
l j The key variable used by the licensee for reactor building monitoring j is the reactor building pressure, which is monitored by Category 1 i instrumentation; the reactor building atmosphere temperature is a backup
! variable for reactor building accident monitoring, and as such, is measured l by Category 3 instrumentation. .
j We find that the licensee's application of Category 3 backup '
instrumentation is in accordance with the regulatory guide, i
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The licensee states that the presently installed 0 to 300*F.
containment temperature indicators provide sufficient range to monitor the entire spectrum of containment temperature transients as analyzed in the FSAR.
i Based on this justification, we find that the existing range is adequate to monitor this variable during all accident and post-accident
- . conditions.
3.3.16 Containment Sump Water Temperature Regulatory Guide 1.97 recommends Category 2 instrumentation for this-variable. The Itcensee has supplied instrumentation that,'except for environmental qualification, is Category 2. In Reference 4, the licensee-states that the minimum available net positive suction head for the decay i heat removal pump is independent of sump temperature and no automatic or manual actions are initiated based on this temperature. No additional justification was provided by Reference S.
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j The temperature of the sump water is useful to the operator in
- determining the amount of containment heat removed during recirculation.
Therefore, an environmentally qualified means of determining the containment sump water temperature should be provided by the licensee.
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- t 3.3.17 Letdown Flow-Out i e Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The licensee does not consider this variable to be a post-accident Category 2 instrument, and has supplied Category 3
, instrumentation. The licensee states that this variable is not required in
. the mitigation of an accident and that the letdown system is isolated by any accident requiring containment isolation. l
, 1 As this variable is not utilized in conjunction with a safety system, I
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t we find that the instrumentation provided for this variable is acceptable, i
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3.3.18 Component Coolina Water Temperature to Engineered Safety Feature (ESF) System Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The licensee is supplying instrumentation that, except for environmental qualification, is Category 2. The licensee states that the ,
decay heat removal heat exchanger outlet temperature provides an adequate measure of the decay heat removal closed cooling water system heat removal .
capability.
In Reference 5, the licensee states that this instrumentation is not located in a harsh environment, therefore, qualification to the requirements of 10 CFR 50.49 is not required. Thus, we find this instrumentation acceptable.
3.3.19 Component Coolina Water Flow to ESF System Regulatory Guide 1.97 recommends Category 2 flow instrumentation for this variable. The licensee does not have instrumentation for this variable. The licensee justified this exception in Reference 4 by stating that, since all decay heat and nuclear services closed cycle cooling systems component cooling water valves are manually operated and art:
nornelly open, pump status and system temperature is sufficient indication for system operation.
In Reference 5, the licensee gave information on the availability of pump discharge pressure indication and low flow alarms in the control room in addition to the pump status and temperature indication. We conclude that the intent of Regulatory Guide 1.97 is met with the instrumentation provided. Therefore, we find this deviation acceptable.
3.3.20 Radioactive Gas Holdup Tank Pressure Regulatory Guide 1.97 recommends control room instrumentation for this variable with a range of 0 to 150 percent of design pressure. The itcensee has local indication only. The licensee states that the design pressure 14
for these tanks.is 150'psig. 'When the pressure reaches 82 psig, it initiates a local high pressure alarm. Also, the pressure can be indicated.
on a local indicator on demand. At 85 psig, the relief. valve opens and discharges to the auxiliary building, where it will be detected and indicated by the auxiliary building radiation monitor. Also, when the relief valve opens, it will annunciate in the common problem panel in the control room. ,
In Reference 4, the licensee did not state what the range of the local indicator is nor state that the instrumentation is accessible post-accident. In Reference 5, the licensee states that the range of the ,
local instrument is 0-100 psig, which is adequate to monitor the tank pressure. The licensee also stated that the local indication is available on the radioactive waste control panel which is accessible after an accident. We find this acceptable.
3.3.21 Status of Standby Power and Other Enerav Sources Important to Safety Regulatory Guide 1.97 recommends Category c instrumentation for this variable. The licensee has provided instrument air instrumentation that, except for environmental qualification, is Category 2.
In Reference 5, the licensee states that this instrumentation is not located in a harsh environment, therefore, qualification to the requirements of 10 CFR 50.49 is not required. We find this acceptable.
3.3.22 Vent from Steam Generator Safety Relief Valves or Atmospheric Dumo Valves
. The instrumentation provided for this variable has a range of 3.96 x 10' to 980 pC1/cc. Regulatory Guide 1.97 recommends 10' to 103 pC1/cc. The existing range does not envelop the upper end of the 15
d recommended range. The existing range deviates from the reconenended range by 20 pC1/cc, but is adequate to provide the necessary accident and post-accident information. Therefore, this is an acceptable deviation from Regulatory Guide 1.97.
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- 4. CONCLUSIONS i
Based on our review, we find that the licensee either conforms to or is justified in deviating from Regulatory Guide 1.97, with the following 4 exceptions:
!. deviation as part of their review of NUREG-0737, Item II.B.3 (Section 3.3.1).
- 2. Accumulator tank level and pressure--the licensee should provide a level or pressure instrument for this variable that is environmentally qualifted in accordance with Regulatory Guide 1.97 and 10 CFR 50.49 (Section 3.3.7).
- 3. Pressurizer level--the licensee should commit to installing
- redundant Category 1 instrumentation for this variable.
(Section 3.3.10).
- 4. Pressurizer heater status--the licensee should provide the i
recommended instrumentation (Section 3.3.11).
- 5. Containment sump water temperature--the Itcensee should identify an environmentally qualified means of monitoring this variable (Section 3.3.16).
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- 5. REFERENCES
- 1. NRC letter, D. G. Eisenhut to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Permits, " Supplement No. 1 to NUREG-0737--Requirements for Emergency Response Capability (Generic Letter No. 82-33)," December 17, 1982.
- 2. Inst'rumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions Durina and followina an Accident, Regulatory Guide 1.97, Revision 2, NRC, Office of Standards
- Development, December 1980.
- 3. Clarification of TMI Action Plan Reautrements. Reautrements for Emergency Response Capability, NUREG-0737 Supplement No. 1, NRC, Office of Nuclear Reactor Regulation, January 1983.
- 4. GPU Nuclear Corporation letter, H. D. Hukill to Office of Nuclear Reactor Regulation, NRC, October 1, 1984, Serial No. 5211-84-2252
- 5. GPU Nuclear Corporation letter, H. D. Hukill to Office of Nuclear Reactor Regulation, NRC, " Emergency Response Capability-Conformance to Regulatory Guide 1.97," June 5, 1986, 5211-86-2097.
- 6. Instrumentation for Licht-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions Durina and Fo110wina an Accident, Regulatory Guide 1.97, Revision 3, NRC, Office of Nuclear Regulatcry Research, May 1983.
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This EG&G Idaho, Inc. report reviews the submittals for Three Mile Island, Unit No.1, and identifies areas of nonconformance to Regulatory Guide 1.97.
l Exceptions to these guidelines are evaluated and those areas where sufficient basis for acceptability is not provided are identified.
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